AUGEAN SECTION 73 EAST NORTHANTS RESOURCE …€¦ · environmental statement section 73...

AUGEAN ENVIRONMENTAL STATEMENT

SECTION 73 APPLICATIONS

EAST NORTHANTS RESOURCE MANAGEMENT FACILITY

AU/KCE/SPS/1612/01/ES

April 2012 AU_KCEp11821

APPENDIX ESG

APPLICATION FOR DISPOSAL OF LLW INCLUDING HV-VLLW UNDER THE RADIOACTIVE SUBSTANCES ACT 1993 FOR THE EAST NORTHANTS RESOURCE

MANAGEMENT FACILITY. SUPPORTING INFORMATION. JULY 2009 (ON CD)

Application for disposal of LLW including HV-VLLW

Under the Radioactive Substances Act 1993, for the East Northants Resource

Management Facility

Supporting Information

July 2009

Application for disposal of LLW including HV-VLLW under RSA 1993, for the East Northants Resource Management Facility: Supporting Information

July 2009 2

Preface This authorisation application was prepared by Augean plc with support from:

Decommissioning and waste specialists from the United Kingdom Atomic Energy Authority (UKAEA) Harwell site (recently renamed RSRL).

Technical assessments were provided by Galson Sciences Ltd using a framework developed by the Scotland & Northern Ireland Forum for Environmental Research.

Supplementary technical studies were provided by UKAEA Ltd. Technical Services Group.

Background materials concerning radioactivity for those unfamiliar with the subject were obtained from the International Atomic Energy Authority (IAEA).

Occupational radiation protection advice was provided by the Health Protection Agency (HPA).

Capability statements for the professional team are given in Annex J.


July 2009 3

Contents Summary Supporting Information for the Application 1.0 Introduction 2.0 Authorisation 2.1 Background 2.2 What is LLW? 2.3 Strategic Need 3.0 Policy and Regulatory Background 3.1 Radioactive Substances Regulation 3.2 The Radioactive Substances Act 3.3 Risk 3.4 UK Government Policy 3.5 Basic Safety Standards 3.6 Environmental Permitting Regulations 3.7 Conservation Regulations 3.8 Ionising Radiations Regulations 3.9 Nuclear Industry LLW Strategy 3.10 Other 4.0 Site Background Information 4.1 Site Description and Local Environment 4.2 Business Plans and Site Development Plans 4.3 Existing Permits 5.0 Radioactive Waste Disposal Proposal 5.1 Principles and Dose Criteria 5.2 Sources of Waste 5.3 Road Transport 5.4 Receipt and Assay 5.5 Accumulation and Quarantine 5.6 Disposal, Waste Emplacement, Compaction, Cover and Handling 5.7 Worker Radiation Protection 5.8 Environmental Radioactivity Monitoring 6.0 Waste Disposal History 7.0 Proposals for Liquid and Gaseous Discharges 8.0 Radioactive Waste Disposal Consequence Assessment and Radiological Capacity 8.1 Pre-Closure – expected to occur

Direct Radiation Exposure from Waste Handling and Emplacement 8.2 Pre-Closure – expected to occur

Exposure from Gas Generation from the Landfill 8.3 Pre-Closure – not expected to occur

Dropped Load of Waste 8.4 Pre-Closure – not expected to occur


July 2009 4

Wound Exposure 8.5 Pre-Closure – not expected to occur

Exposure from Fire 8.6 Pre Closure and Aftercare Period – expected to occur

Exposure from Leachate Processing Offsite – Sewage Works 8.7 Pre Closure and Aftercare Period – not certain to occur

Exposure from Leachate - Spillage 8.8 Pre Closure and Aftercare Period – not certain to occur

Exposure from Aerosols 8.9 Post-Closure – expected to occur

Exposure by Using Groundwater at Nearest Abstraction Point 8.10 Post-Closure – expected to occur

Exposure from Gas Generation from the Landfill 8.11 Post-Closure – expected to occur

Exposure to Wildlife from all sources 8.12 Post-Closure – expected to occur

External dose from emplaced wastes 8.13 Post –Closure not expected to occur

Exposure by Using Groundwater from a Borehole Constructed at the Boundary of the Landfill

8.14 Post –Closure not expected to occur Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill

8.15 Results of the Assessment 8.16 Landfill Radiological Capacity

9.0 Radioactive Waste Disposal Proposed Authorisation Conditions and Waste Acceptance

Criteria 9.1 Potential Conditions Arising - Standard RSA Authorisation Template 9.2 Potential Conditions Arising - Existing Landfill Permit & the Landfill Regulations 9.3 Conditions Arising from the Site Specific Risk Assessment and Industry Practice 10.0 BPEO Assessment for LLW Disposal of Waste from Nuclear Sites 10.1 BPEO 11.0 BPM and ALARA Assessment for the Proposed Radioactive Waste Disposal 11.1 ALARA 11.2 BPM 12.0 Landfill Engineering and BAT Features of the Existing Landfill 13.0 Waste Hierarchy and Waste Minimisation at Source 14.0 Summary of the Existing Environmental Statement for the Site and Impacts of the

Proposal 15.0 Outline of Management and Operating Arrangements 16.0 Stakeholder Consultation 17.0 The Applications Forms 17.1 Waste Disposal 18.0 Conclusion


July 2009 5

References Figures Figure 1 Site Location Figure 2 Site Layout Glossary Annexes

A Radiation, People and the Environment (IAEA, 2004)

B Suitability Assessment – Galson Sciences

C ENRMF, IRRs 1999, Radiation Risk Assessment for Low Level Waste Disposal, HPA

D Dose Rate calculations in support of Low Level waste disposal authorisation, TSG(09)0487

E SNIFFER Methodology Information

F Copy of Application Form

G Example Capacity Calculation Layout

H Calculation of dose rate at landfill, TSG(09)0488

I Baseline Groundwater and Leachate Sample Results

J Capability Statements


July 2009 6

Summary Introduction S1 This document provides supporting information for an application for

authorisation under the Radioactive Substances Act 1993, for disposal of solid Low Level Radioactive waste (LLW) of up to 200 Bq/g, including High Volume Very Low Level Waste (HV-VLLW), at the East Northants Resource Management Facility (ENRMF), operated by Augean plc.

S2 The waste has a very low radioactivity content which this application

demonstrates would present a very low risk if disposed. S3 This document provides information to the Environment Agency, as regulator, in

order that they can consider the application for authorisation. This document is also a public document.

S4 This document contains specialist terms which are required to communicate the

information to the regulator and it is recognised that this may make the document less accessible to a wider audience. The main document contains a comprehensive Glossary of technical terms used. A more detailed booklet on radiation, people and the environment (published by the International Atomic Energy Agency) is referenced in Annex A to the main document. These information sources may be of further use to readers new to the subject matter.

S5 This application for an authorisation contains proposed arrangements and

conditions which are subject to regulatory approval and changes. If the application is granted, the conditions that apply will be those established by the authorisation and by detailed supporting operational documentation prepared to address the authorisation.


July 2009 7

Background S6 The use of landfill is an established approach to the disposal of LLW with low

specific activity and is supported by Government policy (ref 3). Disposal of LLW to landfill is authorised under the Radioactive Substances Act 1993 (ref 4) using permits/authorisations issued by the Environment Agency in England. The permitting arrangements are currently under review to incorporate the approach within the Environmental Permitting Regulations 2010 (ref 19).

S7 Disposal routes in the UK for LLW are limited and often the only option available

is disposal to the LLW repository near to the village of Drigg in Cumbria. The LLW repository does not have capacity for the volumes of the full range of LLW (up to 4000 Bq/g alpha and 12,000 Bq/g beta/gamma) that will be generated from broad decommissioning of the nuclear industry. The disposal of LLW at the lower end of the range of specific activity is not thought to be a sustainable use of the repository, which has been designed and engineered to a standard suitable for materials with a radioactive content at the higher end of the range for LLW. The strategic need for alternative fit for purpose disposal routes is established and detailed within the UK nuclear industry LLW strategy (consultation) (ref 20) and for the non-nuclear industry in UK government policy (ref 3).

S8 The proposed LLW contains very small amounts of radioactivity; less than or

equal to 200 Bq/g. The waste can be handled safely by humans in direct contact with the material in a manner similar to other low hazard wastes. The material is a radioactive waste in accordance with legal definitions but, in the case of this application, it contains radioactivity of less than the bottom 5% of the range of low level radioactive wastes. The waste does not need special security measures.

S9 The LLW that will be disposed of arises from the decommissioning and clean-up

of nuclear industry sites and from non-nuclear industry sources, such as hospitals.

S10 Typically the waste is rubble, soils, crushed concrete, bricks and metals that arise

from demolition of buildings that were previously used for nuclear research or power generation. A large programme of work to decommission the nuclear legacies created since the 1940’s is currently underway in the UK that will generate significant volumes of LLW. The UK Nuclear Industry LLW strategy and supporting inventories (ref 20) provide detailed information on the potential types and nature of the wastes.

S11 During decommissioning, the highest radioactive hazards are removed prior to

demolition of structures. What remains after decommissioning is a mixture of construction materials/soils either that can be proven clean or which sometimes contain trace levels of radioactivity. Efforts are made to separate out radioactivity, to sort wastes, to recycle materials and to reuse materials. The wastes that remain with trace levels of radioactivity after this process are typical of the wastes proposed for disposal at ENRMF.


July 2009 8

Radioactivity & Risk

S12 This summary contains a short Annex which briefly explains radioactivity. Humans are exposed to ionising radiation every day. This exposure comes from background radiation.

S13 Humans can have additional exposure from other sources; for example having an x-ray or flying in a commercial aeroplane results in additional exposure to ionising radiation. It is necessary to limit exposure because there is a possibility of adverse health effects.

S14 In the UK there is a consensus that for exposure to radioactivity the risk of a fatality of 1 in 1,000,000 (one in a million) per year can be regarded by society as a level of risk to a member of the public beyond which further reduction may not be justified.

S15 A risk of one in a million is very low. For comparison the average annual risk of

death for the following is approximately:

Smoking 10 cigarettes per day 1 in 200 Natural causes for someone aged 40 1 in 700 Accidents in the home 1 in 10,000 Lighting strike 1 in 10,000,000 (1 in 10 million)

S16 The waste disposal process proposed has been designed such that the risk to

the public in the long term is broadly less than one in a million per year. This is consistent with the risk guidance level set by regulatory guidance (ref 18) and is better than the risk constraint established by the HPA guidance (ref 14) of 1 in 100,000 per year. This risk guidance level is achieved by limiting the radioactive content and amount of the waste that the landfill can receive. Hence long term public safety is an inherent feature of the proposal and does not depend on future human actions.

S17 For unlikely intrusion events, including for example, if the landfill is excavated by

future society and the land reused for residential properties, a dose guidance level has been used which is the lower end of the range indicated by regulatory guidance (refs 18, 14).

S18 Over time radioactivity decreases because radioactive decay is a process that

eventually leads to the original wastes becoming non-radioactive. For some types (radionuclides), this takes so long that it can be ignored, but for others the effect is relatively quick and after only a few years or decades the waste becomes essentially non-radioactive.


July 2009 9

The East Northants Resource Management Facility S19 The landfill site lies approximately 2.5km north of the village of King’s Cliffe in the

East Northamptonshire District of the County of Northamptonshire (Figure 1). The closest village to the site is Duddington, approximately 2.2km to the North West. The setting is generally rural with a majority of the land surrounding the landfill site comprising open farmland or woodland.

S20 Landfilling operations at East Northants Resource Management Facility

commenced in 2002. The site has been previously known as the Kingscliffe landfill site and as the Slipe Clay Pit.

S21 The facility is the subject of a Permit and operates as a hazardous waste landfill

with a number of ancillary waste activities and treatment processes on the site. The landfill site is engineered to the highest standards consistent with hazardous landfills. It is lined with a composite barrier of high density polyethylene and 1.5m thickness of clay. The disposal rate of the engineered landfill cells with hazardous and inert (the inert material is used for cover and construction of access tracks) waste is permitted at a maximum of 249,999 tonnes/year (Figure 2).

S22 It is currently envisaged that landfill operations will continue until approximately

2013, dependant on the actual importation rate. The site will be progressively restored and once complete will undergo a defined scheme of capping and final restoration. The afteruse of the site will be principally grassland and wildflower meadows for ecological and agricultural purposes.

S23 The proposal for LLW disposal at the site will not change the annual tonnage, the

total capacity of the site or the physical features that contributed to the original landfill permitting decision.


July 2009 10

The LLW Disposal Proposal S24 The application document provides an outline of the proposed arrangements for

the LLW disposal process. After granting of the authorisation this outline would be developed into detailed operating arrangements in accordance with the authorisation conditions.

S25 The process has the following key features:

The wastes will be transported to the site in accordance with relevant transport regulations that apply to radioactive wastes. The regulations are established to control the risks from, for example, transport accidents that result in waste spillage. The waste would typically be contained in double sealed bulk bags or 200 litre metal drums.

Loose or exposed LLW waste will not be transported to the landfill site or handled at the site.

Wastes arriving at the landfill under the authorisation will be pre-notified both for transport purposes and for acceptability against the waste acceptance criteria. Prior to physical receipt of the waste a package of information concerning the characteristics of the waste will be submitted by the sender for acceptance by the landfill. Augean’s Technical Assessment team will check the characterisation information to ensure that the waste is adequately described and that the waste meets the waste acceptance criteria. These checks may involve quality assurance analysis that is independent of the sender.

Wastes arriving at the landfill will be subject to physical inspection to check the integrity of the waste packages and to check the external radiation dose.

If a waste consignment fails to be acceptable upon receipt at the site entrance and can safely be returned to the sender, it will be refused entry to the site.

The stringent pre-acceptance measures will ensure that only acceptable wastes are received at the landfill. In the very unlikely case that a waste consignment fails to be acceptable upon receipt and may not be safe to return to the sender (for example, if a package has been damaged) the landfill site operator will quarantine the waste in a safe area set aside for that purpose. The response plan for such cases would utilise the resources of the consignor and would involve the regulatory authorities.

Acceptable wastes will be disposed to the landfill void after receipt. The waste will be moved to the landfill working face along roads made of suitable hardcore materials.

The waste packages will be lifted using mechanical equipment and placed into the landfill at the base of the waste face. Waste packages will not be tumble tipped.


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Immediately after completion of a phase of waste emplacement, the waste will be covered with at least a 300mm thickness of suitable cover on all exposed surfaces and sufficient to ensure that the dose rate at a height of 1 metre is less than 2 microSv/hr.

A record will be kept of the waste disposal location.

S26 Workplace and environmental monitoring will be carried out including, groundwater monitoring, leachate monitoring, surface water monitoring and monitoring of the working areas.


July 2009 12

Key Facts Summary Operational

Legal Dose Limit for Workers 20 mSv/yr++

Legal Dose Limit for the Public 1 mSv/yr

Dose Criteria for Workers for this application <1 mSv/yr

Dose Constraint for the Public from a single source 0.3 mSv/yr

Design Dose Criterion for the Public for this application <0.02 mSv/yr

Maximum doserate 1 metre above covered emplaced waste 2 microSv/hr

Minimum daily cover thickness over emplaced waste 300 mm

Maximum doserate at perimeter of quarantine area 2 microSv/hr

Maximum doserate at 1 metre from package 10 microSv/hr

Current and future maximum annual landfill input for waste 249,999 tonnes/yr Under Environmental Permit

For this specific application LLW (including HV-VLLW) is:

o Solid

o Less than or equal to 200 Bq/g total specific activity

o Has a greater specific activity than an applicable exemption or exclusion level**

o May contain waste defined in policy as HV-VLLW

o May contain waste less than an exemption or exclusion level where separation is impracticable

o May contain waste that were it not radioactive would be classified as Inert, Non-Hazardous or Hazardous

Post Closure and Aftercare Period

Long term risk guidance level for public exposure 1 in a million/yr

Long term dose guidance and design level for public exposure 0.02 mSv/yr

Dose guidance level for exposure from intrusion events 3 mSv/yr – 20 mSv/yr (The range reflects the gradation between long term and transient events) ++mSv and microSv are measures of radiation dose (see Annex to the summary) ** Exemption and exclusion levels are the lower limits of what is legally defined to be radioactive waste


July 2009 13

Summary of the Environmental Impacts of the Proposal S27 The risk from radioactivity has been assessed using a conservative predictive

model. The model has been used to calculate the capacity of the landfill for radioactivity that would result in a risk of less than one in a million per year to the public in the long term and the capacity that meets the dose criteria set for the unlikely circumstance of future human intrusion into the waste.

S28 In practice the risk will be even further reduced through good operating practices

and future management arrangements, but these are not assumed by the model. S29 An assessment has been carried out of the exposure of the landfill workers which

shows that exposures can be maintained below a dose criterion of 1 mSv/yr (the dose limit for workers is 20 mSv/yr). This has been confirmed by advice received from the Radiological Protection Advisor for the site, the HPA (ref 16).

S30 The following list summarises the overall impact of the LLW proposal on the

existing environmental impact statement for the landfill site.

Impact to Groundwater An insignificant risk from pollution

Impact to Surface Water An insignificant risk from pollution

Impact to Landscape No change in the landform

Traffic Impact No additional traffic, very low risks from traffic incident

Impact from Noise No additional noise

Impact to Ecology No additional landtake, insignificant risk to animals.

Impact to Air Quality An insignificant risk from release to atmosphere

Impact to Human Health Insignificant risk in the long and short term

Archaeology No impacts identified

Proposed Authorisation Conditions S31 The application contains a series of proposed authorisation conditions which are

subject to agreement with the Environment Agency. A proposal is made to reflect certain conditions from the existing landfill permit/risk assessment into the LLW authorisation in order to ensure consistency with current limits and standards.


July 2009 14

S32 The actual capacity of the landfill for LLW depends on the mixture of different radionuclides disposed. This is due to the fact that some radionuclides present more risk under certain scenarios than others.

S33 The exact mixture of radionuclides that will be sent to the landfill is not known

prior to the process commencing because in many cases the wastes have not yet been generated by the senders through completion of their decommissioning works. The mixture of nuclides in any particular consignment would always be known and approved by Augean prior to receipt.

S34 The proposal is that the capacity of the landfill is subject to a total capacity limit combined with a series of other conditions. The total capacity limit would apply from the date of issue until closure of the landfill or until the capacity is reached. The landfill would receive no more LLW wastes under the permit once the capacity limit is reached. The capacity limit cannot be expressed as a single number because it depends on the mixture received up to any point in time, so the proposal is for a continuously revised capacity limit based on individual nuclides (including appropriate daughter chains). The total capacity limit would be established using an authorised spreadsheet model agreed with the regulator. The spreadsheet model would represent the most restrictive case from the risk assessment and would produce as an output the remaining capacity of the landfill on an individual nuclide basis given the exact wastes received to that point in time. Prior to accepting any further waste the model would be used by the landfill operator to determine that the consignment would not lead to a breach of the total capacity limit.

S35 This disposal is not in addition to the existing 249,999 tonne per year landfill

disposal rate but is part of it and hence no additional traffic results.

Example Waste Stream S36 The application contains information on an example waste stream from the

Harwell site. It is proposed that the authorisation at ENRMF facilitates the reception of LLW with a specific activity of less than or equal to 200 Bq/g from any consignor in the UK where the consignor demonstrates to their regulator that disposal to ENRMF is the best practicable environmental option. The Harwell site is typical of the decommissioning sites and information from the site has been included for purposes of illustration only.


July 2009 15

Questions & Answers S37 Why has the East Northants Resource Management Facility been proposed for

LLW disposal?

The East Northants Resource Management Facility is a modern landfill site constructed to the high quality standards required for hazardous waste disposal and is hence technically suitable for LLW disposal.

There are few such well engineered sites in the UK. The UK Nuclear Industry Strategy (ref 20) notes that whilst transport and proximity are important considerations, when considered on a national level the issue is not a strong differentiator between options because the additional impact to transport infrastructure or carbon emissions is low. The proposal in this case would not result in a net increase in traffic to the site because the annual tonnage capacity limit is unchanged.

S38 Why is it proposed to use a hazardous waste site for LLW?

Hazardous waste sites are constructed using high standards of environmental protection engineering and are subject to rigorous regulation throughout their operating period and post-closure. The site has also an established pre-acceptance, technical assessment, consignment and acceptance regime unlike non-hazardous landfills. Regulations require that LLW is managed using the Best Practicable Means and this is achieved by using a hazardous waste site.

S39 What is the worst case impact that could happen as a result of the disposal of

LLW?

The worst case events are considered in detail in the authorisation application.

The worst case during the waste disposal phase is that a waste package is dropped and the contents spilled. The consequences of such an unlikely occurrence are minor.

The worst case after closure of the landfill site is an occurrence in which the waste is excavated without knowledge of the contents and then the waste is used, for example, as the surface material for new housing development. This is a very unlikely occurrence and the consequences of it happening are that members of the public would be exposed to low levels of radioactivity, below the regulatory dose criteria.

The radioactive content of the LLW (<=200 Bq/g) is limited by the authorisation to be of such low concentration that it is inherently incapable of giving rise to significant exposure to members of the public under credible scenarios.


July 2009 16

S40 What are the impacts of transporting the LLW to the site?

The safety of the transport of radioactive materials (including LLW with very low radioactivity content) is governed by UK dangerous goods transport regulations. These require the wastes to be contained in packages appropriate for the level of radioactivity bearing in mind what would happen in a transport accident.

If the waste were involved in an accident during transport to the site an established response arrangement involving the emergency services augmented by suitably qualified and experienced advisors and monitoring specialists would be enacted. If waste were spilled it would be a simple matter of recovering the spilt materials and sweeping the road. The levels of radioactivity involved would not require extensive arrangements during the recovery operation and the risk to members of the public from exposure to radioactivity would be very low.

The number of vehicle movements and the environmental impact of transport are not increased by this authorisation application because the total capacity of the landfill is unchanged by the proposal.

S41 Have other alternatives to landfill disposal been considered?

Every site in the nuclear industry (which includes the power stations and research sites) wishing to consign LLW to the landfill will first have to demonstrate to the regulatory authorities that landfill of their waste is the Best Practicable Environmental Option (BPEO). This is a requirement for the consigning sites to be granted a transfer authorisation under the Radioactive Substances Act which they will require to send wastes. They will also have to demonstrate that they have complied with the waste hierarchy and have therefore exhausted options to Reduce, Recycle and Reuse the materials.

Consideration of the BPEO requires all alternatives to be considered. In some cases, for example, LLW can be treated, incinerated or recovered through smelting. Landfill disposal will be considered as a last resort after these other approaches have been considered.

It is likely the majority of LLW will arise from historically contaminated land and buildings for which the other waste management options are generally less applicable.

S42 What controls would be in place to ensure the LLW waste can be disposed of

safely?

The waste will be subject to pre-acceptance tests to ensure it is acceptable.

The waste will be checked upon arrival at the site. Radioactivity can be easily and immediately measured using simple instruments to ensure waste acceptability.


July 2009 17

The waste will be handled in enclosed containers.

The waste will be disposed immediately after receipt and covered with material layers.

The landfill is experienced with handling hazardous wastes, has well developed procedures/arrangements and has a good safety culture.

Specific procedures and a radiation protection plan will be established for the LLW operations.

Monitoring will be carried out to ensure protection of the workers and the public.

The operations are subject to specific authorisation, regulation and inspection by the EA and works will be carried out in accordance with the authorisation conditions.

S43 Who regulates the disposal of LLW?

The principal regulator in England is the Environment Agency and the disposal is authorised under the Radioactive Substances Act. The occupational safety of workers and the public is regulated by the Heath & Safety Executive principally under the Ionising Radiations Regulations. The road transport of radioactive materials is regulated by the Department for Transport.

Conclusion

S44 There is a strategic need for landfill waste disposal routes for materials containing very low levels of radioactivity that arise from decommissioning the nuclear industry sites and from other sources. Several such routes are likely to be required in the UK and the proposal to use the East Northants Resource Management Facility would provide a significant capability.

S45 The amount and concentration of radioactivity in the waste is very low and presents a very low risk. A risk assessment has been carried out and the capacity of the landfill to receive the waste has been estimated using a conservative method.

S46 The proposal for disposal of LLW is subject to agreement of, and the issue of an authorisation by the Environment Agency.


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Annex to the Summary: What is Radioactivity?

Introduction

Humans are exposed to ionising radiation every day. This exposure comes from background radiation.

Humans can have additional exposure from other sources; for example having an x-ray or flying in a commercial aeroplane results in additional exposure to ionising radiation. It is necessary to limit exposure because there is a possibility of adverse health effects.

Annex A to the application document gives full background information on radiation.

What is Ionising Radiation?

All matter is made up of atoms consisting of a nucleus surrounded by negatively charged electrons, similar to the sun surrounded by the planets. The nucleus consists of neutrons and positively charged protons.

Atoms containing the same number of protons have identical chemical properties and are known as elements. Elements with a different number of neutrons are known as isotopes. There are 88 naturally occurring elements some examples of which are oxygen, iron, sulphur, uranium and radon gas.

Some atoms are radioactive (they are called radionuclides) and the nucleus of such atoms can change structure (lose energy); in so doing the energy is emitted as radiation in three main forms:

alpha rays, beta rays and gamma rays.

This process is termed radioactive decay and the resulting daughter product, a new element, is formed as a result. These radiations can interact with surrounding matter to produce positively and negatively charged particles (a type of electricity). This process is called ionisation, hence the term “ionising radiation”. X-rays are also known as ionising radiation and they are identical to gamma rays except they are emitted by electrons, not by the nucleus.


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What are the Properties of Ionising Radiation?

Alpha rays and beta rays are sub-atomic particles that travel at close to the speed of light (300,000,000 metres per second). Alpha rays can be stopped (energy absorbed) by a piece of paper, while beta rays can be stopped by one or two centimetres of human tissue.

Gamma rays and X-rays are waves of energy similar to visible light, except they have more energy and are invisible. They travel at the speed of light and penetrate matter more easily than the particulate radiations.

What Units are used to Measure Radioactivity?

Radiation is measured in decays (disintegrations) per second which corresponds to the number of nuclei losing energy each second. One Becquerel (abbreviation Bq) is equal to one decay per second: one megabequerel is equal to one million disintegrations per second. The human body is naturally radioactive due to the presence of radioactive potassium: A 70 kilogram person would contain about 3500 Bq.

How Does Radiation Interact with Matter?

When the energy from radiation is absorbed by matter, chemical changes occur at the atomic level. If the exposure is large enough these changes can be readily observed. For example, if glass is heavily irradiated it changes colour. Some precious stones are coloured for commercial purposes using this method. When the body is subjected to a medical X-ray the bones absorb most of the energy and a photographic film can then give an image of the skeleton. The amount of radiation absorbed per gram of matter is called the “absorbed dose”.

What Units Are Used To Measure Absorbed Dose?

Absorbed dose is measured in grays (abbreviation Gy). One gray corresponds to one joule of radiation energy deposited in one kilogram of matter. (Note: It would require 320,000 joules of energy to boil one kilogram (one litre) of water). This is a large unit and the milligray (mGy), which is one thousandth of a gray, is more commonly used.

When radiation interacts with living tissue the effect it has varies with the type of radiation. Alpha rays are 20 times more effective than beta and gamma rays at causing tissue damage. To allow for this, the dose in grays is multiplied by an effectiveness factor and the new units are called sieverts (abbreviation Sv) and the dose is called the “equivalent dose”. A one milligray dose of alpha rays is equal to 20 mSv (millisieverts) of equivalent dose. A one milligray dose of beta rays is equal to 1 mSv equivalent dose because the effectiveness factor is 1 for beta rays. In most cases the effectiveness factor is unity and the dose in grays is equal to the dose in sieverts.


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How Does Radiation Interact with the Human Body?

When radiation is absorbed in the body it causes chemical reactions to occur which can alter the normal functions of the body. At high doses (above 1 sievert) this can result in massive cell death, organ damage and possibly death to the individual. At low doses (less than 50 mSv) the situation is more complex.

The body is made up of different cells. For example we have brain cells, muscle cells, blood cells etc. It is the genes within a cell that determine how a cell functions. If damage occurs to the genes then it is possible for a cancer to occur. This means the cell has lost the ability to control the rate at which it reproduces.

Radiation can cause this effect and at low doses it is the only known deleterious health effect. This type of event is very unlikely to occur, and an estimate of its frequency can only be obtained by measuring the effect at higher doses and calculating the probability at low doses.

Annex A of the application document gives more detail of the health effects of radiation.

The Natural Background

The effect of radiation on health must be discussed within the context of the natural background. Background radiation consists of cosmic rays from space and radiation present in the earth from when it was formed. Cosmic radiation increases with altitude and so airline pilots receive a higher exposure from this source; the dose rate at 12,000 metres being about 150 times the sea level dose. The terrestrial radiation comes from naturally occurring radioisotopes of potassium and rubidium and from decay products of uranium and thorium. On average two thirds of the dose people receive comes from terrestrial sources. Most of this dose comes from the gas, radon, which is a decay product of uranium and thorium. Radon emanates from the soil and tends to concentrate in buildings.

Source Of Exposure Exposure

Total Natural Radiation (Average UK) 2.2 mSv per year

Seven Hour Aeroplane Flight 0.02 to 0.07 mSv

Chest X-Ray 0.04 mSv

Cosmic Radiation Exposure of Domestic Airline Pilot 2 mSv per year

Examples of Exposure to Ionising Radiation


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Exposure Limits

The International Commission on Radiological Protection (ICRP) has set the following limits on exposure to ionising radiation:

The general public shall not be exposed to more than 1 mSv per annum (over and above natural background).

Occupational exposure shall not exceed 20 mSv per annum.

These limits exclude exposure due to background and medical radiation.

More restrictive targets than these limits are proposed by the authorisation application.

In this application the dose criteria for workers is <1mSv/yr during operations and they should on average receive much less than this. For the public the long term dose criteria is broadly equivalent to less than 0.02mSv/yr – which is equivalent to a 1% increase compared to the average background and is less than received during a typical transatlantic flight.

For this application:

<1 mSv/year is the worker maximum dose criteria ~ 6 transatlantic flights per year

0.02 mSv/year is the public dose guidance level ~ less than 1 transatlantic flight per year

Monitoring Of Radiation Exposure

People who are occupationally exposed to ionising radiation can be monitored with a dosemeter which is worn as a badge attached to clothing. At monthly intervals the dosemeter is sent to a laboratory where the radiation exposure can be read.

Samples of air, foodstuffs, soils, water and soils can be analysed to give an estimate of public exposure.


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Supporting Information for the Application


July 2009 23

1.0 Introduction 1.0.1 This document provides supporting information for an application for

authorisation under the Radioactive Substances Act (RSA) 1993, for disposal of solid Low Level Radioactive waste (LLW) with a specific activity of less than or equal to 200 Bq/g, including High Volume Very Low Level Waste (HV-VLLW), at the East Northants Resource Management Facility, operated by Augean plc.

1.0.2 The waste has a very low radioactivity content which this application

demonstrates would present a very low risk if disposed. 1.0.3 This document provides information to the Environment Agency, as regulator, in

order that they can consider the application for authorisation. This document is also a public document.

1.0.4 The key sections of the document are:

The summary.

The main body of text provides information relating to the application in accordance with the guidance on contents issued by the Environment Agency (ref 1).

A glossary with explanations of special terms.

An Annex containing further introductory information on radiation and radioactivity.

Annexes containing risk assessments for the application which examine the safety of the proposed waste disposals.

An Annex containing background information on the risk assessment methodology.

A copy of the application form for the disposal authorisation (under sect 13 of the RSA).

1.0.5 This document contains proposed arrangements and conditions which are

subject to regulatory approval and changes. If the application is granted the conditions that apply will be those established by the authorisation and by detailed supporting operational documentation prepared to address the authorisation.


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2.0 Authorisation 2.1 Background 2.1.1 Landfill disposal is considered a valid option (ref 3, 18, 19, 20) for the disposal of

LLW with a low specific activity from the nuclear industry and from other non-nuclear industry sources such as hospitals.

2.1.2 The disposal will be authorised under the Radioactive Substances Act 1993 (ref

4). The permitting arrangements are currently under review to incorporate the approach within the Environmental Permitting Regulations 2010 (ref 19).

2.1.3 The proposed LLW (<=200 Bq/g) for disposal contains very small amounts of

radioactivity. The waste can be handled safely by humans in direct contact with the material in a manner similar to other low hazard wastes. The material is a radioactive waste in accordance with legal definitions but, in the case of this application, it contains radioactivity of less than the bottom 5% of the range of low level radioactive wastes. The waste does not need special security measures.

2.1.4 LLW arises from the decommissioning and remediation of nuclear industry sites

and from non-nuclear industry sources, such as hospitals.

2.2 What is LLW?

2.2.1 Typical LLW proposed for the ENRMF landfill is rubble, soils, crushed concrete, bricks and metals that arise from demolition of buildings that were previously used for nuclear research or power generation. A large programme of work to decommission the nuclear legacies created since the 1940’s is currently underway in the UK that will generate significant volumes of LLW. The UK Nuclear Industry LLW strategy and supporting inventories (ref 20) provide detailed information on the potential types and nature of the wastes.

2.2.2 During decommissioning, the highest radioactive hazards are removed prior to

demolition of structures. What remains after decommissioning is a mixture of construction materials/soils either that can be proven clean or which sometimes contain trace levels of radioactivity. Efforts are made to separate out radioactivity, to sort wastes, to recycle materials and to reuse materials. The wastes that remain with trace levels of radioactivity after this process are LLW.

2.2.3 The LLW which is the subject of this application is:

A solid material which is slightly radioactive. The proposed waste has a radioactivity content that is in the bottom 5% of the range of wastes that are defined as Low Level Waste (LLW) (using the upper limit of acceptance at the LLWR of 4,000 to 12,000 Bq/g).


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The proposed LLW waste will have a radioactivity content of less than or equal to 200 Bq/g. Where Bq/g is Becquerel per gram, a Becquerel is a measure of radioactivity equivalent to 1 disintegration per second and hence Bq/g is a measure of the “concentration” of radioactivity, also called specific activity.

The lower limit of LLW for man made substances is currently 0.4 Bq/g below which the material is not subject to specific regulatory control. Other exemption/exclusion levels may apply to particular nuclides/radioelements. For this authorisation application, the waste is a LLW in the range:

- a specific activity greater than an applicable exemption/exclusion level and up to 200 Bq/g total specific activity.

If wastes of less than the exemption/exclusion level are mixed in with the LLW as an inevitable result of their production, in a manner that makes separation impracticable, then these would also be treated as LLW.

The total specific activity would be averaged appropriately in order to be representative of the individual waste package and in any case over not more than 4 tonnes.

The LLW may contain waste which were it not classified as a radioactive waste would be classified as Inert, Non-Hazardous or Hazardous.

The LLW may contain waste which would be defined as High Volume – Very Low Level Waste (HV-VLLW) in accordance with policy (ref 3), but is not limited to that definition.

2.3 Strategic Need 2.3.1 Disposal routes in the UK for LLW are limited and often the only option available

is disposal to the LLW repository near to the village of Drigg in Cumbria. The LLW repository does not have capacity for the volumes of LLW that will be generated from broad decommissioning of the nuclear industry and it is not thought to be a sustainable use of the repository, which has been designed and engineered for materials with radioactivity content in the higher range of activity of LLW. The strategic need for alternative fit for purpose disposal routes is established and detailed within the UK nuclear industry LLW strategy (ref 20) and for non-nuclear industry users in UK government policy (ref 3).

2.3.2 The strategic drivers for new LLW disposal routes are:

Decommissioning: A disposal route for LLW with low specific activity will make it possible to decommission many nuclear industry and non-nuclear industry legacies across the UK. The lack of such a route may hold-up decommissioning and increase costs for the taxpayer.

Sustainability: It is government policy that LLW management solutions should be provided earlier rather than later. The provision of a new LLW


July 2009 26

management option will allow current stockpiles of waste to be disposed safely and hence not leave the issue to future generations.

Technical Benefit: Some of the waste that will be labelled as LLW from the nuclear industry will be essentially clean demolition materials that for technical reasons cannot be proven to be clean for free release. Such projects are currently difficult to undertake and the provision of a new LLW management option will enable such projects to proceed.

Regional Scale: One approach for the provision of new LLW management options is to provide specialised landfill facilities at the site of origin. However, this could result in many relatively small landfills being constructed across the UK at the existing nuclear sites, many of which are not in favourable geological settings for such uses. As is the case for conventional wastes, it is reasonable to propose that regional solutions which balance transport distance with economies of scale are worth consideration. The transport of LLW with low specific activity does not present challenging hazards because of the very low levels of radioactivity involved. The amounts are small compared to conventional wastes and will be generated slowly over several decades. A LLW disposal route at the East Northants Resource Management Facility site could serve multiple nuclear industry sites. There are few such well engineered sites in the UK. The UK Nuclear Industry Strategy (ref 20) notes that whilst transport and proximity are important considerations when considered on a national level the issue is not a strong differentiator between options because the additional impact to transport infrastructure or carbon emissions is low. The proposal in this case would not result in a net increase in traffic to the site because the annual tonnage capacity limit is unchanged.

International Experience: Other countries that have progressed with the clean-up of their nuclear legacies have found great benefit from having waste routes for LLW disposal to landfill. There are examples from the USA and both Spain and France have recently opened such a route.

UK Government Policy: Advisory committees in the UK have examined the case for provision of such waste routes (ref 5) and concluded that government policy should be supportive. Government policy in this area has recently been revised and enables the provision of landfill waste routes for LLW under appropriate circumstances (ref 3). The UK Nuclear Industry LLW Strategy is supportive of the option (ref 20).

Low Level Waste Repository (LLWR) Acceptance Criteria: The LLW Repository Ltd criteria for waste acceptance at the disposal facility near the village of Drigg, Cumbria states that: “Waste shall not be Consigned for disposal if reasonably practicable measures could be adopted to segregate it from other arisings such that disposal is possible as any of the following: very low level waste, low level waste in domestic refuse or as a special precautions disposal at suitable landfill sites”. This is consistent with government policy.


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3.0 Policy and Regulatory Background 3.0.1 This section is not intended to be a comprehensive review and reference should

be made to the source documents for further detail. Points of particular relevance to the application for authorisation are made.

3.1 Radioactive Substances Regulation 3.1.1 Regulation of LLW is summarised by the Environment Agency in “Considerations

for Radioactive Substances Regulation under the RSA 1993 at Nuclear Sites...” (ref 2). Although the East Northants Resource Management Facility is not and will not be a nuclear licensed site, the provisions apply to wastes that arise from the nuclear industry who operate on nuclear licensed sites.

The Environment Agency is responsible under the Radioactive Substances Act 1993 for regulating all disposals of LLW from nuclear sites in England and Wales.

The Environment Agency issues authorisations (permits) which include limits and conditions.

The Environment Agency regulates the “source” site which generates or transfers the LLW and the “destination” disposal site which receives the waste in the case of solid waste disposal.

In addition to issuing or varying authorisations, the Environment Agency periodically reviews authorisations, carries out inspections, investigates incidents, assesses public exposure and has powers of enforcement.

3.1.2 The guidance on requirements for authorisation (ref 18) for near-surface disposal facilities for solid radioactive wastes has recently been issued. It is proposed that this application falls outside of the scope of that guidance because the application does not involve a facility solely for the disposal of solid radioactive waste and can be assessed using simple conservative approaches. However, the application has been prepared to be consistent with the guidance. The following list describes key relevant points from the guidance and where they are addressed by the application:


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Objective/Principle/Requirement from Guidance on Requirements for Authorisation

Addressed by…

The fundamental protection objective is to ensure that all disposals of solid radioactive waste to facilities on land are made in a way that safeguards the interests of people and the environment, now and in the future, commands public confidence and is cost-effective.

Section 5.1 and 8.0

Principle1: Solid radioactive waste shall be disposed of in such a way that the assessed radiological risks to people and the environment in the future are no greater than the risks that would be acceptable at the time of disposal.

Section 8.0

Principle 2: Both at the time of disposal and in the future, the radiological risks to people and the environment from a disposal of solid radioactive waste shall be as low as reasonable achievable under the circumstances prevailing at the time of disposal, taking into account economic and societal factors and the need to manage any non-radiological hazards.

Section 11.1

Principle 3: Both at the time of disposal and in the future, the standard of protection to people and the environment against radiological hazards from a disposal of solid radioactive waste shall be no less stringent than the nationally acceptable standard at the time of the disposal.

Section 5.1 and 8.0

Principle 4: The level of protection to people and the environment against any non-radiological hazards associated with disposing of solid radioactive waste shall be no less stringent than that provided by the nationally acceptable standard for such hazards from the disposal of any other waste at the time of disposal for wastes that present a non-radiological but not a radiological hazard.

Section 9.2 and 12.0

Principle 5: Both at the time of disposal and in the future, unreasonable reliance shall not be placed on human action to protect the public and the environment against radiological and any non-radiological hazards from a disposal of solid radioactive waste.

Section 8.0

R1 and R2 n/a

R3: The developer should take the lead on dialogue with the potential host community, other interested parties and the general public.

Section 16.0

R4: An application under RSA 93 relating to a proposed disposal of solid radioactive waste should be supported by an environmental safety case.

Section 8.0

R5: The developer/operator of a disposal facility for solid radioactive waste should foster and nurture a positive environmental safety culture at all times and should have a management system, organisational structure and resources sufficient to provide the following functions: (a) planning and control of work; (b) the application of sound science and good engineering practice; (c) provision of information; (d) documentation and record keeping; (e) quality management.

Section 15.0

R6: During the period of authorisation of a disposal facility for solid Section 8.0


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Objective/Principle/Requirement from Guidance on Requirements for Authorisation

Addressed by…

radioactive waste, the effective dose from the facility to a representative member of the critical group should not exceed a source-related dose constraint of 0.3 mSv/year.

R7: After the period of authorisation, the assessed radiological risk from a disposal facility to a person representative of those at greatest risk should be consistent with a risk guidance level of 10-6 per year (i.e. 1 in a million per year).

Section 8.0

R8: The developer/operator of a near-surface disposal facility should assess the potential consequences of human intrusion into the facility after the period of authorisation on the basis that it is likely to occur. The developer/operator should, however, consider and implement any practical measures that might reduce the chance of its happening. The assessed effective dose to any person during and after the assumed intrusion should be consistent with a dose guidance level in the range of around 3 mSv/year to around 20 mSv/year.

Section 8.0

R9: The choice of waste acceptance criteria, how the selected site is used and the design, construction, operation, closure and post-closure management of the disposal facility should ensure that radiological risks to members of the public and to the environment, both during the period of authorisation and afterwards, are as low as reasonably achievable (ALARA), taking into account economic and social factors.

Section 5.0 and 9.0

R10: The developer/operator should carry out an assessment to show that the radiological effects of a disposal facility on the accessible environment are acceptably low, both during the period of authorisation and afterwards.


R11: The developer/operator of a disposal facility for solid radioactive waste should demonstrate that the disposal system provides adequate protection against non-radiological hazards.


R12: n/a

R13: The developer/operator of a disposal facility for solid radioactive waste should make sure that the site is used and the facility is designed, constructed, operated and capable of closure so as to avoid unacceptable effects on the performance of the disposal system.

Section 2.0 and 9.0

R14: The developer/operator of a disposal facility for solid radioactive waste should establish waste acceptance criteria consistent with the assumptions made in the environmental safety case and with the requirements for transport and handling, and demonstrate that these can be applied during operations at the facility.

Section 9.0

R15: In support of the environmental safety case, the developer/operator of a disposal facility for solid radioactive waste should carry out a programme to monitor for changes caused by construction, operation and closure of the facility.

Section 5.8


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3.1.3 The HPA have recently issued their advice on Radiological Protection Objectives for the Land-Based Disposal of Solid Radioactive Wastes (ref 14) which provides overlapping guidance to the guidance on requirements for authorisation. Where there is a conflict between these guidance documents the guidance on requirements for authorisation has been used.

3.2 The Radioactive Substances Act 1993 3.2.1 The Radioactive Substances Act (RSA) 1993 defines what a radioactive waste is,

establishes that users of radioactive material/waste must be registered (section 6 and 7) and establishes arrangements for the authorisation of disposal and accumulation of radioactive waste (section 13). The Environment Agency provides regulation of RSA 1993 in England and Wales.

3.2.2 This application for authorisation of disposal and for registration as a user of

radioactive materials is made under the RSA 1993. 3.2.3 Note that the East Northants Resource Management Facility is an existing

permitted hazardous waste landfill under the Landfill Regulations 2002. The LLW authorisation, should one be granted, is additional to the existing permits and may be considered a separate but overlapping regime. The total tonnage capacity and the broad environmental impact of the landfill will be unaffected by the permitting of this type of waste.

3.2.4 Where wastes are currently acceptable under the existing permits it is proposed

that these arrangements will not be altered by any new authorisation granted under RSA 1993. The LLW authorised under RSA 1993 will have to comply with both the existing risk assessments (which underpin the existing permit conditions, as far as they can be applied and referred to in the new permit) and the RSA authorisation.

3.2.5 LLW is radioactive waste and is therefore not a “controlled” waste in England and

Wales (ref 7 and 11). The regime of regulation that applies to conventional landfill sites encompasses controlled wastes and does not cover LLW. The Hazardous Waste Regulations exclude radioactive waste except in the special case where the waste is exempt from the requirements of the RSA 1993 for disposal but is still a radioactive waste. This can occur for some so-called “exempt” wastes that are <0.4 Bq/g and contain man-made radionuclides. Such wastes do not need to be treated as LLW (ref 8).

3.2.6 A hazardous waste landfill site such as the East Northants Resource

Management Facility can receive LLW if authorised under RSA 1993. LLW is not a controlled waste and therefore disposal alongside other wastes does not represent co-disposal of inert or non-hazardous waste with hazardous waste. If the LLW also has hazardous waste properties then dual application of the hazardous waste landfill risk assessment and the RSA authorisation conditions provides adequate regulation. In cases where the LLW does not have hazardous


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waste properties the risk assessment underpinning the existing permits is conservative because LLW with non hazardous or inert properties will have a lower non-radiological risk than the existing risk assessment allows for.

3.2.7 An approach to regulation could be for the RSA authorisation to replicate the

relevant conditions of the existing permit/risk assessments in addition to further conditions specific to the LLW. This could be achieved by reference to the existing risk assessments which underpin the current landfill permit and this approach is assumed in the remainder of this document.

Existing Permit under the

Landfill Regulations 2002 New Authorisation

under RSA 1993 for LLW disposal which

also refers to existing risk assessment

constraints A Hazardous Waste stream Applies Does not apply because

the waste is not a LLW LLW that does not also

possess Hazardous Waste properties

Does not apply because the waste is not a controlled waste

Applies

LLW that does also possess Hazardous Waste properties

Does not apply because the waste is not a controlled waste

Applies

The parallel regimes of the Landfill Regulations and the Radioactive Substances Act

3.2.8 The majority of the wastes will be LLW with other properties that are non-

hazardous or inert. This raises the question of why a hazardous waste site such as the East Northants Resource Management Facility should be authorised as opposed to a non-hazardous site. The reasons for proposing this arrangement are:

Use of a hazardous site will allow the small amounts of LLW that are also landfillable hazardous waste to be consigned (for example, asbestos gaskets in radioactively contaminated waste ventilation ducts) and this helps prevent such wastes becoming orphaned.

Hazardous waste sites have to be engineered using Best Available Techniques (BAT) that are to the highest standard for landfill and therefore represent a good choice for LLW which have to be disposed of using Best Practicable Means (BPM) that also meet a high standard.

Hazardous landfills have well developed operational procedures for the acceptance and handling of difficult wastes. For example, the East Northants Resource Management Facility has a comprehensive laboratory with qualified chemists who will be able to manage the acceptance and disposal of the wastes. The facility utilises waste handling methods that are identical to those required for emplacing LLW.


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Hazardous waste landfills receive wastes which are essentially incombustible and which pass a flammability waste acceptance test. This reduces the probability of a landfill fire to a very low level.

Hazardous wastes sites are not common in the UK and are therefore much less likely to be “forgotten” by future societies. This is useful in minimising the risk that future generations will be exposed to the waste. (Note that the risk assessment assumes that the site is forgotten as a worst case assumption and demonstrates that the disposal is safe regardless of this.)

The probability of inadvertent human intrusion is reduced by the following features of the East Northants Resource Management Facility :

Depth of the disposal horizon. In the majority of cases the LLW will be buried several metres into the hazardous waste disposals which will act as a visual indicator for future generations that the area is a waste site.

The site is in an area of low mineral resource potential (the clay has been extracted) which makes it less likely that future generations will wish to dig deep into the waste.

The nature of the site as a hazardous site makes it more likely that records and knowledge of the site will be retained by future generations.

The engineered cover design provided for hazardous waste sites makes penetration into the waste less likely.

The landform design, whilst being sympathetic with the surroundings, is nonetheless relatively obvious as a non natural feature.

3.2.9 For the purposes of this authorisation the applicant wishes the site to be treated as a “non-nuclear premises” and for the authorisation to be held by the applicant for the site (ref 18, 9.2.16). The applicant does not wish for the disposal authorisation to be held by the consignor(s).

3.3 Risk

3.3.1 Radiation protection and regulation is concerned with ensuring the protection of humans from the risks presented by ionising radiation whilst taking into account the benefits offered by a particular process.

3.3.2 In the case of the East Northants Resource Management Facility authorisation

the focus is to assess the risks presented by the disposal of LLW to the workers and the public. The potential benefits of such a waste disposal route have been outlined above. The overall regulatory assessment is a balancing of risks and benefits.


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3.3.3 Any nuclear site wishing to send waste to such a disposal route will have to carry out an assessment of risks and benefits using a framework called “Best Practicable Environmental Option”.

3.3.4 The risk presented by ionising radiation is directly related to the amount of

radiation to which a person is exposed. The radioactivity exposure is called “dose”. Internationally accepted systems of relating risk to dose have been established and are used by the UK (refs 12, 14, 16).

3.3.5 Risk guidance levels/criteria are established by UK policy and internationally

accepted good practice. The amount of LLW that can be disposed in a landfill is limited by the risk guidance levels/criteria and subject to further optimisation through consideration of the process as a whole.

3.3.6 The risk arising from a waste disposal can be calculated for the short and long

term to both workers and the public. This can then be used to establish the amount of radioactivity that can be disposed in the landfill (which is called the radiological capacity) without exceeding the risk criteria.

3.3.7 The authorised radiological capacity is set to ensure that the resulting dose

presents a very low risk. This is achieved through prospective dose estimation/risk assessment calculations. Once the disposal process is operating, workplace and environmental monitoring can be used to check the actual exposure of humans to radioactivity.

3.3.8 It is not necessary to calculate the exposure of all humans exposed to the

radioactivity as long as the humans that might be most exposed are assessed. This is called the “critical group” methodology. In practice, a few groups may be assessed to ensure coverage of workers and the public both in the long and short timescales.

3.4 UK Government Policy 3.4.1 Policy on radioactive waste is set out in the White Paper, Review of Radioactive

Waste Management Policy, Cm 2919. In respect of LLW, the policy has been amended by the Policy for the Long Term Management of Solid Low Level Radioactive Waste (ref 3).

3.4.2 The policy allows for disposal of LLW at specified landfill sites, provided that this

meets regulatory requirements. The maximum specific activity of the LLW (<=200 Bq/g) that will be accepted at the site has been derived based on safety of handling using current operational practice, optimisation of the risk assessment and from the needs of the consignors.

3.4.3 The policy also establishes that disposal to an appropriately engineered facility

with no intent to retrieve should be the end point for LLW that remains after


July 2009 34

reasonable attempts have been made to avoid, reduce, recycle and reuse the material.

3.4.4 The government policy advises that a risk target of 10-6/year (one in a million) of

developing a fatal cancer or serious hereditary defect should be used as an objective in the design process. Where the regulators are satisfied that best practicable means have been adopted by the operator to limit risks and the estimated risks to the public (now and in the future) are below this target, then no further reductions in risk should be sought. The guidance has been developed further in guidance of requirements for authorisation (ref 18) discussed above.

3.5 Basic Safety Standards Directive 1996 (BSSD) and The

Radioactive Substances Direction 2000 3.5.1 European law influences the regulation of LLW disposal as follows:

A requirement is established to achieve the ALARA (as low as reasonably achievable) principle. All exposures to ionising radiation of any member of the public and of the population as a whole resulting from the disposal of LLW are kept as low as reasonably achievable, economic and social factors being taken into account (ALARA).

To achieve ALARA the stated maximum doses to individuals resulting from a defined source are 0.3 mSv/year from any single new source (mSv is milliSievert a measure of radiation dose; for comparison the background natural radiation dose in the UK is 2.2 mSv/year or more). This applies to current discharges that could be altered by changes to current operating arrangements. Government policy in Cm 2919 proposes a threshold of 0.02 mSv/year as equivalent to an annual risk of death of around 1 in a million/year (10-6/yr).

The range 0.3 mSv/yr to 0.02 mSv/yr is the dose range over which a process can be optimised in accordance with the ALARA principle. If doses are below 0.02 mSv/yr, the regulators should not seek to secure further reductions provided they are satisfied that the operator is using best practicable means to limit discharges.

Regardless of the above targets there is a UK legal limit derived from European law for the exposure of nuclear workers and the public from all man-made sources of radioactivity (other than medical exposure). The dose limit for members of the public is 1 mSv/yr and the dose limit for workers is 20mSv/yr. In the case of the ENRMF the landfill and other workers would be managed to a dose limit of 1 mSv/yr as specified by the site’s radiation protection plan.

3.5.2 BSSD defines the optimisation principle. In current practice a concept called the

use of Best Practicable Means or BPM is used to demonstrate in part that optimisation has been addressed. The EA is currently consulting on whether


July 2009 35

radioactive substances regulation should be brought under the Environmental Permitting regime. In the latter case the Best Available Techniques or BAT concept would be applied to the LLW management process.

3.5.3 BSSD sets out requirements on the EA in relation to permitting which have been

addressed by this application including: dose limits, dose constraints, authorisation conditions designed to be protective of human health, authorisation limits which have in-built safety factors, flexible authorisation limits and proposals for environmental monitoring.

3.5.4 BSSD sets out the requirements that management of radioactive waste disposal

should be undertaken following consultation by an operator with a Qualified Expert. For this application qualified experts are provided under contract to the site operator by the Health Protection Agency and the site operator has used suitably qualified and experienced advisors to prepare the application obtained from Galson Sciences and UKAEA.

3.6 Environmental Permitting Regulations 2007 3.6.1 The existing East Northants Resource Management Facility is permitted under

the Environmental Permitting(England and Wales) Regulations 2007. These regulations set out a pollution control regime for landfills.

3.6.2 To operate the landfill, a permit was issued under the Pollution Prevention and

Control (PPC) Regulations 2000. These regulations have just been rationalised into the Environmental Permitting Regulations 2007. A new Environmental Permit was issued for the site in March 2009. PPC and EPR seeks to improve environmental protection by introducing measures to reduce or prevent emissions to air, land and water.

3.6.3 The existing landfill has been built and is operated within these regulations and

hence the pollution prevention measures which exist are of direct use in minimising pollution from the LLW.

3.6.4 The EPR regime does not currently incorporate radioactive materials because

they are not controlled wastes. However an activity may be controlled by both EPR and the Radioactive Substances Act 1993 and that regulators will ensure that the two regimes do not impose conflicting obligations on the same matter.

3.6.5 The existing landfill is permitted under these regulations and that permit applies

conditions. In order for LLW to be permitted for disposal in the landfill they require to be permitted under the Radioactive Substances Act and it is proposed that the risk assessment constraints that apply to the existing landfill should be replicated within the RSA authorisation where applicable and in a non conflicting manner.


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3.6.6 This application document does not include detailed information concerning the existing permit and the underpinning risk assessments because these documents have been previously submitted through the regulator.

3.7 Conservation (Natural Habitats and Conservation) Regulations 1994

3.7.1 The Habitats Regulations require the Environment Agency to be satisfied that the

integrity of designated “European sites” (sites with certain ecological value) will not be adversely affected by relevant permissions issued by the Agency. These regulations have been addressed by the existing permit for the landfill.

3.8 Ionising Radiations Regulations 1999 3.8.1 Ionising radiations occur as either electromagnetic rays (such as X-rays and

gamma rays) or particles (such as alpha and beta particles). Radiation occurs naturally (e.g. from the radioactive decay of natural radioactive substances such as radon gas and its decay products) but can also be produced artificially. People can be exposed externally, to radiation from a radioactive material or a generator such as an X-ray set, or internally, by inhaling or ingesting radioactive substances. Wounds that become contaminated by radioactive material can also cause radioactive exposure.

3.8.2 Everyone receives some exposure to natural background radiation and much of

the population also has the occasional medical or dental X-ray. The Health and Safety Executive (HSE) is concerned with the control of exposure to radiation arising from the use of radioactive materials and radiation generators in work activities in the nuclear industry; waste; medical and dental practice; manufacturing; construction; engineering; paper; offshore drilling; education (colleges, schools) and non-destructive testing industries.

3.8.3 The main legal requirements enforced by HSE are detailed in the Work with

ionising radiation: Ionising Radiations Regulations 1999 Approved code of practice and guidance.

3.8.4 The regulations will apply to the workers at the landfill and visitors to the site.

The regulations require risk assessment and specialist advice from a Radiation Protection Adviser to be enacted prior to work with ionising radiations. The advice received will result in a safe system of work at the site to limit, control and measure exposure.

3.8.5 The very low amounts of radioactivity in the LLW mean that relatively simple

arrangements will be sufficient and that the workers will be treated as members of the public for purposes of dose limitation.


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3.9 Nuclear Industry LLW Strategy 3.9.1 The UK Nuclear Industry Strategy has been published for consultation by the

Nuclear Decommissioning Authority (NDA). The strategy underpins the strategic need for new waste management options for LLW disposal. A key theme of the strategy is “development and use of new fit for purpose management and disposal routes, so waste producers have more choice in determining and implementing waste management routes”.

3.10 Other 3.10.1 Article 37 of the Euratom Treaty requires member states to the European

Commission to provide sufficient information about plans to dispose of radioactive waste to allow the Commission to decide whether the plans could cause radioactive contamination of the water, soil or airspace of another Member State. It is assumed by this application that Article 37 submissions, where required, are implemented by the consigning nuclear industry sites.

3.10.2 The 1992 OSPAR convention and related national strategies seeks progressive

and substantial reductions of discharges, emissions and losses of radioactive substances to the marine environment. The strategies use a dose guidance level for the public of 0.02 mSv/yr which is consistent with that used in this application. The use of BAT methods is proposed and is consistent with the approach used for this application.

3.10.3 The UK ratified Joint Convention on the Safety of Spent Fuel Management and

on the Safety of Radioactive Waste Management (IAEA 1997) sets out a framework for radioactive waste management. This application has been made in a manner consistent with the objectives of this convention.

3.10.4 The IAEA publish good practice guidance in relation to radioactive waste

management. This application is consistent with IAEA guidance. 3.10.5 The ICRP is an independent advisory body that provides recommendations on

radiation protection. In the UK the HPA provide advice on the recommendations of the ICRP. The HPA’s advice concerning the most recent publication ICRP 103 changes the previous advice to use a single source dose constraint of 0.15 mSv. Whilst this application uses the 0.3 mSv figure which is current to UK policy, it is not used for calculating the radiological capacity of the landfill and adoption of the 0.15 mSv constraint would make no impact on the key dimensions of the proposal.

3.10.6 The Town and Country Planning (Environmental Impact Assessment)

Regulations 1999 may require the environmental effects of development of radioactive waste facilities to be assessed as part of the planning approval process under the Town and Country Planning Act 1990. A separate planning approval process is being pursued for the ENRMF in order for it to receive LLW.


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3.10.7 The new regulations to transpose the 2006 Groundwater Directive expected in

2009 may apply to radioactive substances. 3.10.8 The EA has previously issued guidance concerning BPM and BPEO in relation to

waste management options. The current move is to replace these concepts with BAT, this being underpinned by developing Radioactive Substances Regulation Environmental Principles. This application has been prepared taking into account these developments in so far as practicable.

3.10.9 The Nuclear Safeguards Act 2000 establishes arrangements through which the

UK accounts for and protects nuclear materials (plutonium, uranium and thorium). The consigning nuclear sites would provide accountancy for such materials up to the point of disposal.

3.10.10 Radioactive waste is transported to a disposal facility under strict controls in

accordance with Dangerous Goods Regulations which are based upon the transport regulations issued by the IAEA. For this application the site will not accept unpackaged wastes even where such wastes are compliant with the transport regulations. The Dangerous Goods Safety Advisor (DGSA) for the consigning sites ensures compliance with the regulations and the contract conditions. For this application the receiving site will also have an appropriately qualified DGSA to provide advice on auditing, checking and receiving packages.

3.10.11 The Environmental Permitting Regulations 2010 may be applied to Radioactive

Substances Regulation and are currently under consultation. 3.10.12 An assessment of the radiological impact on species other than humans may

be required to address the Environment Act 1995 and the Conservation Regulations 1994. This application has incorporated such an assessment.


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4.0 Site Background Information

4.1 Site Description and Local Environment 4.1.1 The landfill site lies approximately 2.5km north of the village of King’s Cliffe in the

East Northamptonshire District of the County of Northamptonshire at National Grid Reference TF010 001 (ref 15) (Figure 1). The setting is generally rural with a majority of the land surrounding the landfill site comprising open farmland or woodland. The only properties in the immediate vicinity of the landfill comprise a terrace of three houses (Westhay Cottages) and Westhay Farm with associated agricultural and commercial buildings. These properties are all located to the east of the eastern boundary of the landfill site on the other side of the site access road.

4.1.2 Landfilling operations at East Northants Resource Management Facility

commenced in 2002. 4.1.3 The closest village to the site is Duddington, approximately 2.2km to the

northwest and King’s Cliffe village which lies approximately 2.5km to the south (Figure 1). Collyweston village is approximately 3.3 km to the north of the site. The only other development within the vicinity of the site is the RAF airfield at Wittering which lies approximately 800m to the northeast. This is an operational airfield used for pilot training.

4.1.4 The landfill lies within the Rockingham Forest/Lower Nene Valley Special

Landscape Area, a local designation adopted by the County Council in 1974. This is an area of relatively level to gently undulating land at an elevation of approximately 85m above ordnance datum. The predominant land uses within the immediate area of the site are agriculture and woodland.

4.1.5 Part of the Collyweston Great Woods to the north of the site is designated a Site

of Special Scientific Interest (SSSI) and a National Nature Reserve (NNR). 4.1.6 The A47 road lies approximately 1 km to the north of the site, with the Stamford

Road, an unclassified road linking the A47 to the village of King’s Cliffe, passing along the eastern boundary of the site (Figure 1).

4.1.7 The geology of the site consists of an upper clay strata formed from glacial till

and estuarine mudstone to a depth of approximately 11.5m, underlain by Jurassic limestone. The clay strata have been partially worked as part of the quarrying operations on site, with overburden materials stockpiled to the western end of the site and when required silica clay is exported off site.

4.1.8 There are no main water courses in proximity to the landfill site, the closest being

Willow Brook (3km south) and the River Welland (2.5km west), the landfill site being approximately on the watershed between these two.


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4.2 Business Plans and Site Development Plans 4.2.1 The landfill site is operated as a hazardous waste landfill with a number of

ancillary and related waste activities on the site. The disposal rate of the engineered landfill cells with hazardous and inert (for cover etc.) waste is permitted at a maximum rate of 249,999 tonnes/year (Figure 2).

4.2.2 It is envisaged that landfill operations will continue until approximately 2013,

dependant on the actual importation rate. The site will be progressively restored and once complete will undergo a defined scheme of capping and restoration. In accordance with the extant planning permission the landfill site will be restored principally to grasslands for ecological and agricultural afteruse.

4.2.3 The proposal for LLW disposal at the site would not be envisaged to change the

total capacity of the site or the physical features that contributed to the original landfill permitting decision.

4.2.4 Operating details for the site are not presented here and are available in the

supporting documentation for the existing permitted operations (ref 15). The operating arrangements and culture at the site are consistent with the arrangements proposed for LLW disposal in this application.

4.3 Existing Permits 4.3.1 The East Northants Resource Management Facility landfill is operating under an

Environmental Permit (TP 3430GW) issued May 2009, for the disposal of hazardous waste. The site commenced operations in 2002 under a PPC Permit and was originally a co-disposal site for the disposal of non-hazardous and hazardous wastes. Since the beginning of 2004, the site has received predominantly hazardous waste and the practice of co-disposal has ceased. The site is therefore now a hazardous only site apart from the need for suitable cover materials. The permit boundary covers an area of 17.27 hectares with some 14.36 hectares for the maximum extent of the cells.

4.3.2 The wastes accepted at East Northants Resource Management Facility cover a

broad spectrum of those defined as hazardous under the European Waste Catalogue subject to the hazardous waste acceptance criteria. These criteria in particular exclude explosive, flammable, corrosive and infectious materials.


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5.0 Radioactive Waste Disposal Proposal 5.0.1 This section provides an outline of the proposed arrangements for the LLW

disposal process. After granting of the authorisation this outline would be developed into detailed operational written safe systems of work in accordance with the authorisation conditions.

5.0.2 It is useful to consider four distinct phases for the timeline of the facility:

The operational phase when the facility is receiving waste.

The active institutional control (aftercare) phase which covers the time from closure of the facility to the time when provisions for active aftercare ceases (60 years or greater in the case of the East Northants Resource Management Facility in accordance with the existing permit).

The passive control period over which records are expected to inform future generations of the presence of radioactive waste.

The uncontrolled phase when all records might be expected to have been lost.

5.1 Principles and Dose Criteria 5.1.1 Dose criteria for LLW disposal are well established in the UK (ref 18) and

considerable good practice guidance exists. In order of decreasing dose, the criteria can be “limits” which are legally established, “constraints” which are levels established by approved practice or “criteria”/“targets” /”guidance levels” which are good practices established by guidance. The dose criteria used for this authorisation application are:

For workers the legal dose limit is 20 mSv/year, and the criterion used for this application is 1 mSv/year, which is the same as the current legal limit for the public. This is an operational criterion and is not used to set the radiological capacity of the landfill because the exposure arises in a manner unrelated to the total capacity of the site. This criterion does affect some of the authorisation conditions, in particular external dose limits on packages and the limit on specific activity. This criterion will be used for radiation protection purposes during operation of the facility.

For the public a legal dose limit of 1 mSv/year and a dose constraint of 0.3 mSv/year would be used during the operational phase. The aim would be to ensure through radiation protection measures and monitoring that no person received more than the dose constraint during the operational phase. This is a constraint and is not used to set the radiological capacity of the landfill because this is considered to be an


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upper bound on the region of dose optimisation. This constraint will be used for radiation protection purposes during operation of the facility.

For all persons during the post-closure phases, for natural processes, the dose guidance level used to set the radiological capacity is 0.02 mSv/yr which corresponds to a risk of 10-6/yr or 1 in a million per year. This is used to set the radiological capacity of the landfill as a whole and not for occupational radiation dose protection. The same criterion is also used as a design target for the operational phase for public exposure.

Inadvertent intrusion into the site in the future is not certain to occur and therefore this event has a low probability of occurrence. The dose guidance level used is 3 mSv/year, which is the lower end of the range indicated by the guidance on requirements for authorisation and HPA guidance (ref 14,18). This is used for direct physical intrusion scenarios and for intrusion by extraction borehole at the site boundary.

It is assumed that following closure of the landfill and the end of the aftercare period the continued ability of the design to meet the risk target does not depend on actions of future generations to maintain integrity of the disposal system.

Following closure of the landfill, it is assumed for the purposes of conservative risk assessment that society prevents intrusion into the waste form for at least 60 years after closure which is consistent with the current financial provision for the long term aftercare of the landfill. In practice the site will be under the control of the existing Environmental Permit until the Environment Agency is satisfied that the site no longer represents a significant risk of harm to human health and pollution of the environment. This period will almost certainly be considerably longer than 60 years. Beyond the 60 year period, it is assumed conservatively and for the purpose of risk assessment that humans may penetrate into the landfill in a manner that results in continuous exposure without realisation of the hazards present. Note that this risk assessment applies only to the LLW component of the waste and not to other hazardous wastes that may be present in the landfill at this time and which have already been permitted for the site on the basis of other risk assessments. 60 years is considered reasonable given that the landfill in question is an extensively engineered and capped hazardous waste permitted site. The assumption of inadvertent intrusion which goes unrealised and which results in humans living on the exposed waste form is considered conservative.

The risk assessment is based upon the principles and scenarios in the “SNIFFER SPB model” (Annex E) as adjusted for this site specific case (Annex B) and as corrected to address developments required to the original SNIFFER model. The SNIFFER model was developed in conjunction with the UK regulatory authorities.

The use of a high quality modern hazardous waste engineered landfill and the stated operational arrangements are designed to ensure that the risk of radiological exposure to members of the public is as low as reasonably


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achievable. This is achieved through the application of best practicable means.

The facility design is already established and is that provided already for the landfill to operate as a permitted hazardous waste disposal facility. Operational arrangements specific to the LLW are established to augment existing arrangements (Sections 5.4 to 5.6).

The summarised dose criteria are:

Operational

Legal Dose Limit for Workers 20 mSv/yr

Legal Dose Limit for the Public 1 mSv/yr

Dose Criterion for Workers for this application <1 mSv/yr

Dose Constraint for the Public from a single source 0.3 mSv/yr

Design Dose Criterion for the Public for this application <0.02 mSv/yr

Post Closure and Aftercare Period

Long term risk guidance level for public exposure 1 in a million/yr

Long term dose guidance and design level for public exposure 0.02 mSv/yr

Dose guidance level for exposure from intrusion events 3 mSv/yr – 20 mSv/yr (The range reflects the gradation between long term and transient events)

5.2 Sources of Waste 5.2.1 The LLW that is to be disposed of under the East Northants Resource

Management Facility authorisation will arise from within the UK. The waste must conform to the landfill waste acceptance criteria as established by the RSA disposal authorisation and, where required, the consigning site must have an appropriate transfer authorisation issued under RSA 1993.

5.2.2 The waste may arise from:

Non-nuclear industry sources: For example, waste derived from hospitals or other non-nuclear licensed users of radioactivity.

Nuclear industry sources: For example, wastes derived from decommissioning of nuclear power stations and research centres.

5.2.3 It is not possible prior to nearer to the time of receipt of the wastes to describe the form, amounts or types of wastes. The general nature of the waste inventory is described in the national inventories for radioactive waste (ref 20). It is


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proposed that if the waste has an underpinning justification for disposal established by the consigning site and it meets the waste acceptance criteria and the waste acceptance criteria of the existing permit as reflected in the RSA authorisation, then the waste is acceptable. This would include wastes that if they were not radioactive would be classified as Inert, Non-Hazardous or Hazardous.

5.3 Road Transport 5.3.1 The following outline arrangements are proposed and will be detailed in the

operating arrangements for the process which will be developed if the authorisation is approved.

5.3.2 The main legislation covering the safe transport of the LLW material is The Carriage of Dangerous Goods…Regulations 2007 (ref 13). The emphasis of the regulations is for the safe management of each stage of the transport chain. Annex A of the ADR contains a section specific to package design to provide the main element of safety in normal and accident conditions.

5.3.3 The onus is on the consignor and carrier of the waste from the source site to ensure that it is transported in accordance with the transport regulations. Specialist advice must be sought from an appropriately trained person holding certification as a Dangerous Goods Safety Advisor (DGSA). Part of the waste acceptance arrangements at the landfill will be checks to ensure that the records and physical condition of the packages meet the transport regulations upon arrival at the landfill site. This is standard practice at the landfill for all waste accepted. Nuclear industry sites are experienced in these arrangements and have developed practices to ensure they are implemented.

5.3.4 LLW contains low amounts and concentrations of radioactivity which mean that they can be transported safely in standard packages used in the transportation of dangerous substances. In some cases the amount of radioactivity will be so low that the packages will be exempt from the regulations. Some of the lower activity wastes will be transportable in “excepted” packages as defined under the regulations and the remainder will be transportable in “industrial” packages. Even where wastes could be transported unpackaged as low specific activity materials in accordance with the regulations, all wastes will be contained in sealed packages.

5.3.5 Typical packages will be either:

Flexible Intermediate Bulk Containers. These are usually called “bulk bags”. The bags would be transported singly stacked on an enclosed freight vehicle and would be handled using pallets or integral lifting loops. It is normal to use double sealed bags. The bags would be placed into the disposal void using mechanical handling equipment.


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Non reusable type approved waste transport drums (200 litre nominal capacity). The drums would be handled on pallets or using drum handling equipment. The drums would be placed into the disposal void using mechanical handling equipment.

Single items may be wrapped and sealed.

5.3.6 Loose or exposed LLW will not be accepted at the landfill site.

5.3.7 Under the transport regulations the consignee (the receiving landfill site) has duties. The landfill site must have a staff training plan in place and a quality assurance programme (operating arrangements) to ensure that regulations are being adhered to. This would entail checks by the landfill operator on receipt of a shipment that the records are correct (all shipments would be pre-notified and pre-accepted for shipment), that the shipment is in accordance with the regulations, that the packages are in good order and that the external dose rate is in accordance with regulations (in addition to the waste acceptance criteria for dose rate). A quarantine arrangement would be implemented for non-compliant consignments.

5.3.8 The consignee must maintain a radiation protection programme in accordance with the Ionising Radiations Regulations.

5.3.9 The consignee must ensure appropriate segregation of the packages, upon receipt, from persons and other dangerous goods in accordance with the regulations.

5.3.10 The consignee must maintain an operating arrangement for emergencies and spillages and provide information to affected parties.

5.4 Pre-acceptance and Assay

5.4.1 The following outline arrangements are proposed and will be detailed in the operating arrangements for the process which will be developed if authorisation is approved.

5.4.2 Wastes that will be delivered at the landfill under the authorisation will be pre-

notified both for radioactive transport purposes and for waste acceptability against the waste acceptance criteria. Prior to dispatch of the waste from the consignor a package of information concerning the characteristics of the waste will be submitted by the consignor for acceptance by the landfill. Augean will check the characterisation information to ensure that the waste is adequately described and that the waste meets the waste acceptance criteria and that the landfill has adequate radiological capacity to receive the waste.

5.4.3 The waste will be characterised so as to facilitate their subsequent management,

including waste disposal. Arrangements for characterisation would be regulated


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by the authorisations for transfer issued to the source sites and established by the contract between the consignor and the disposal facility. Nuclear industry sites operate waste characterisation methodologies in accordance with industry good practice guidance.

5.4.4 Characterisation will include relevant physical, chemical and radiological

properties. 5.4.5 Wastes generated within a well-defined process or which can be demonstrated to

have self-similar characteristics may be characterised as a waste stream. This may mean that reduced characterisation of individual packages is required. However, the radioactive composition and specific activity of each individual waste package would always be reported and averaged over the waste package (or 4 tonnes whichever is the smaller).

5.4.6 Certain characterisation is required by the existing permits for the landfill in

relation to the receipt of hazardous wastes. Such characterisation will be provided for LLW in so far as the conditions of existing permits are referred to by the RSA authorisation for LLW disposal.

5.4.7 Radioactivity related characterisation information for each individual package will

include:

An assessment of the amount, concentration and isotopic composition of the radioactivity in each individual package. This could be obtained, for example, by radiochemical analysis and gamma spectrometry of a representative sample, using “fingerprinting” where applicable or using radiochemical analysis and bulk gamma spectrometry. The history and nature of occurrence of the waste will be taken into account by the source site when designing characterisation approaches. The isotopic composition must be adequate to characterise the waste in accordance with the list of nuclides (heads of chains) used in the risk assessment (Annex B).

The waste characterisation methodology used to obtain the measurements and a justification that this meets a best practicable means approach. The landfill site operator will review the characterisation methodology as part of the waste acceptance process.

The quality assurance methodology used by the consignor to validate the radioactivity measurements.

The external radiation dose required by the transport regulations and at 1 metre from the package on all sides and the top. The justification that this meets the waste acceptance criteria and transport regulations.

The surface contamination clearance records.

Unique identification labelling of each waste package as required under the transport regulations.


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5.4.8 Other characterisation information will include:

The generic information required for basic characterisation by the existing permit for the hazardous waste stream: Source and origin of the waste The process producing the waste Waste treatments applied to the waste The composition of the waste and an assessment against relevant

limit values The appearance of the waste Any equivalent codes applicable to the waste (EWC etc.),

although this would not apply legally to LLW Any hazardous properties according to Schedule 3 of the

Hazardous Waste Regulations. The relevant properties which make the waste hazardous should

that apply A demonstration that the waste is not prohibited For waste streams where each package is not characterised and it

is argued that some of the characteristics are the same across the waste stream; the compositional range for the individual packages.

A measurement of the weight of each package.

The acceptability and characterisation of the waste against the hazardous waste landfill acceptance criteria under existing permits where these are referred to by the RSA authorisation or other non-radiological criteria established by the RSA authorisation. For a hazardous waste landfill site this might generically include:

compliance with the banned substance list,

compliance with the waste acceptance limits for hazardous waste disposal,

the designated leach testing of the waste/waste streams, where applicable,

Details of any pre-conditioning/treatment of the wastes that has been utilised and details of compliance with the “three-point test” for pre-treatment.

The amount of voidage in the waste package (where relevant, for example in the case of wrapped items).

Justification that the waste hierarchy of avoid, reduce, recycle and reuse has been applied to the waste.

Information relating to the safe transport of the waste as required under the transport regulations.


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5.4.9 Each package or self-similar group of packages will have a representative sample taken at the time of packing. The sample will be retained by the source site and be identified uniquely as linked to the package. The nature of LLW means that sampling and analysis has to be carried out using specialist laboratories and that often the source site will be better equipped to manage this than the landfill. Samples may be subject to transport regulations.

5.4.10 The provision of sealed samples is designed to enable the landfill operator or

regulators to request check analysis (compliance testing) without the requirement to open the main package, thereby avoiding double handling and unnecessary exposure of loose waste at the landfill.

5.4.11 Samples will be retained by the source site for 1 year after disposal of the

package and then disposed to the landfill. 5.4.12 Wastes arriving at the landfill will be subject to the following on site verification:

The shipment will be checked while still on the vehicle against the pre-notified characterisation information for consistency.

The external dose rate at 1 metre will be checked.

The packages will be visually checked for integrity.

The transport documentation will be checked for compliance with the transport regulations.

The characterisation documentation will be checked to ensure the waste has been pre-accepted and is compliant.

Receipt records will be generated.

The waste packages will not be opened or sampled at the landfill in order to minimise unnecessary exposure.

5.5 Accumulation and Quarantine 5.5.1 The following outline arrangements are proposed and will be detailed in the

operating procedures and instructions for the process which will be prepared if authorisation is approved. It is noted that receipt of unacceptable waste packages is very unlikely because of the stringent pre-acceptance and transport arrangements applied to the wastes. The arrangements for quarantine are provided as a contingency measure.

If a waste consignment fails to be acceptable upon receipt at the site entrance and can safely be returned to the consignor, it will be refused entry to the site.


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If a waste consignment fails to be acceptable upon receipt and may not be safe to return to the consignor (for example a package has been damaged, or a dose rate exceeded) the landfill site operator will:

Consult the consignor and available information to enable a safe response plan to be generated.

Enact the contingency arrangements where required as outlined below.

In cases where safe to do so, move the consignment and offload to a designated quarantine area for LLW.

Inform the Environment Agency.

For the unlikely case that these contingency measures are executed, the disposal contract between the landfill and the consignor will establish responsibility for remedial action. The consignor will take responsibility for remedial action in co-ordination with the landfill operator. The consignor will complete any necessary regulatory notifications, investigations, remedial action planning and remedial works. Such actions will be subject to specific safety planning prior to execution such that no significant risks are imposed on the landfill operators or public. The consignors of significant volumes of LLW are operators of nuclear licensed sites and will have considerable resources that they can bring into play in order to ensure effective remedial action.

A designated quarantine area will be provided that is marked, physically demarcated off from the rest of the site and well segregated by distance from persons on the site. The design of the quarantine area will be part of the radiation protection plan for the site established to meet the requirements of the Ionising Radiations Regulations. For illustration, the quarantine area might consist of a lockable steel HISO freight container set to one side of the main traffic routes and suitably labelled. The quarantine area shall be so designed to ensure that the dose at the perimeter does not exceed 2 microSv/h.

LLW will not be intentionally accumulated. Wastes received to the site will be placed in the landfill void and covered before the end of the working day. During the working day any accumulations of LLW required for operational reasons will be kept together in designated and marked locations that avoid human exposure through distance. If, for any exceptional reason, LLW cannot be disposed and covered on the day of arrival they will be stored in a designated and segregated area and the Environment Agency informed. The storage area shall be so designed to ensure that the dose at the perimeter does not exceed 2 microSv/h.


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5.6 Disposal, Waste Emplacement, Compaction, Cover and Handling

5.6.1 The following outline arrangements are proposed and will be detailed in the

operating arrangements for the process which will be developed if authorisation is approved.

Waste will be disposed to the landfill void as soon as practicable after receipt. The waste will be moved to the landfill working face along roads made of suitable hardcore materials.

The waste packages will be lifted using mechanical equipment with air conditioned cabins (P3 filter equivalent) and placed into the landfill at the base of the working face.

Immediately after placement of the load the waste will be covered with at least a 300mmm thickness of suitable cover on all exposed surfaces. If the doserate at 1 metre above the emplaced and covered waste is greater than 2 microSv/hr further cover will be added until this doserate is achieved.

A bowser to dowse the waste with water will be on standby wherever waste is unloaded in case of spillage.

A record will be kept of the waste disposal location.

Waste will be disposed of in the current working cell or cells and will be spread throughout the landfill void of that cell without deliberate concentration into one location.

Waste will not be co-located with incompatible other wastes; for example, other wastes that could damage the package during emplacement.

Traffic over existing wastes will be on suitable cover tracks in order to avoid vehicle penetration to the waste layer.

The most likely point at which a load could be dropped or damaged would be during emplacement. Emergency procedures will be enacted to deal with a dropped load situation should this occur.

No loose, unpackaged or exposed wastes will be handled under normal operating conditions.

5.7 Worker Radiation Protection 5.7.1 The following outline arrangements are proposed and will be detailed in the

operating arrangements for the process which will be developed if authorisation is approved.


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5.7.2 The landfill operator will develop a written safe system of work (radiation protection plan) to implement controls and arrangements in accordance with the Ionising Radiations Regulations to ensure worker radiation protection. A plan has been prepared by the appointed Radiation Protection Advisor, the HPA (Annex C).

5.7.3 The system of work will include:

Arrangements for radiation protection for all operations that take place within the boundaries of the landfill site including transport, receipt, quarantine, accumulation, disposal and post-disposal operations.

The system of work will aim to ensure that dose to individuals is optimised and below the dose limit and dose constraint.

The system of work will be based on the principle that landfill operators are not specialist nuclear workers and that they should be protected to limits that would apply to members of the public during the operational phase.

The system of work shall aim to achieve the ALARA principle regardless of limits and constraints that apply.

Actual radiation exposures from direct radiation will be monitored and the systems subject to annual review and improvement.

The system of work will be based upon a prior risk and dose assessment of the potential exposures which will consider routine and accident conditions.

The system of work will document roles and responsibilities, dose assessment and optimisation, surface contamination assessment, segregation, other protective measures, emergency response arrangements, training provision, information provision, competency assessment and quality assurance.

Arrangements for employing the services of a qualified Radiation Protection Adviser to assist with the establishment and operation of the system of work.

Arrangements for monitoring packages and conveyances for radiation dose and arrangements for ensuring surface contamination clearance.

Arrangements for workplace monitoring, where recommended by the radiological protection advisor.

Arrangements for worker monitoring.

Arrangements for recording and reporting exposures.

Arrangements for limiting the external dose received by landfill site workers through primary limitations of the radiation dose waste acceptance criteria, coupled with time of exposure, distance from


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package and additional shielding (for example soil cover on disposed wastes).

5.7.4 The main source of radiation exposure to the landfill workers under normal

conditions will be external radiation from the packages during handling and emplacement. The controls are:

Regardless of the requirements of the transport regulations the maximum dose rate at a specified distance from the package will be limited to a value that ensures that given the likely number of shipments per year the dose constraint of 1 mSv/year will not be exceeded under routine operational conditions.

The maximum concentration of radioactivity in the package is limited by the waste acceptance criteria and conditions of authorisation.

The package is sealed.

Packages are designed to withstand specified drop tests to withstand accidents during emplacement or are robust to the conditions of handling at the landfill.

Incident procedures will be enacted to minimise exposure for accident conditions, such as dropped loads.

The packages will be handled at a distance using mechanical equipment with air conditioned cabins (P3 filter equivalent).

Accumulated packages will be set aside in a quarantined area that is segregated by distance and barriers.

Packages will normally be emplaced in the disposal void immediately upon receipt.

The working zone and face of the disposal area will be covered with an adequate covering of cover material (soils) after each emplacement operation or at the end of the working day in order to reduce external radiation dose to trivial levels.

A bowser to dowse the waste with water will be on standby wherever waste is unloaded in case of spillage.

There will be no loose handling or tipping of wastes.

Packages will be placed in the disposal void and will not be tipped.

Monitoring of the package, workplace and worker for external radiation will be carried out in accordance with an established plan commensurate with risks.

Monitoring of working areas for surface contamination including occasional reassurance monitoring of, for example, wheel wash, traffic routes, gateways and change rooms will be undertaken in accordance with an established plan commensurate with risks.


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Designation of areas (if applicable) under the Ionising Radiations Regulations.

5.8 Environmental Radioactivity Monitoring 5.8.1 The following outline arrangements are proposed and will be detailed in the

operating arrangements for the process which will be developed if authorisation is approved. The monitoring under the Environmental Permit is described in detail in a series of Monitoring and Action Plans (MAPs). The MAPs set out the parameters, frequencies, methodologies and reporting of monitoring for the landfill. Contingency action plans are included in the event that a limit specified in the Permit is exceeded. It is anticipated that radiological monitoring will be added to the MAPs.

5.8.2 The following environmental radiological monitoring is proposed:

Annual radiochemical analysis of groundwater to current monitoring schedules as described under the existing permit for several existing boreholes close to the site. Analysis would be for gamma spectrometry, gross alpha / beta in waters and 3H in aqueous samples.

Annual radiochemical analysis of leachate. Analysis would be for gamma spectrometry, gross alpha / beta in waters and 3H in aqueous samples.

Quarterly radiochemical analysis of surface water discharge. Analysis would be for gamma spectrometry, gross alpha/beta in waters and 3H in aqueous samples.

Annual radiochemical analysis of the landfill gas flare stack emission for the radioactive gases identified in the risk assessment.

Quarterly radiochemical analysis for dust deposited on a powered static air sampler paper at one predominantly downwind location on the site boundary to include gamma spectrometry and gross alpha/beta. (This would be baselined against equivalent samples taken in the period prior to first waste receipt).

Annual analysis of randomly selected surface soils from four points around the site boundary to include gamma spectrometry and gross alpha/beta. (This would be baselined against equivalent samples taken in the period prior to first waste receipt).


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6.0 Waste Disposal History 6.0.1 The original Permit for the site was issued in 2002. The Permit was reviewed in

early 2009 and a new Permit issued in May 2009. Landfilling commenced in 2002 in a series of 5 Phases (Figure 2). The site was originally operated as a co-disposal facility in which Special Wastes were disposed of with non-hazardous biodegradable wastes. This principle of co-disposal relies on the biological activity in the biodegradable wastes to ameliorate the Special Wastes. Phases 1 and 2 of the site were operated by co-disposal. The use of co-disposal ended with the implementation of the Landfill Regulations in 2004. Under the 2004 Regulations the facility became a landfill site for the acceptance of hazardous wastes. Therefore Phases 3 onwards have received hazardous and inert wastes only.

6.0.2 To date Phases 1, 2 and 3 have been completed capped and restored. Phase 4

is operational and Phase 5 needs to be excavated and engineered (Figure 2). Approximately 650,000m3 of waste has been disposed of and there is approximately 700,000m3 of void remaining.

6.0.3 Baseline samples of leachate and groundwater have been analysed for

radioactivity and results are given in Annex I. The baseline sampling showed no enhancements of radioactivity in the groundwater samples. The leachate samples showed a slight enhancement for Tritium, which is commonly found in leachate from landfills across the UK. The levels of Tritium were lower than those usually found in landfill leachates (Annex I).


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7.0 Proposals for Liquid and Gaseous Discharges 7.0.1 The proposal involves no specific authorised liquid or gaseous discharge routes.

Inadvertent discharge to the air from gas generation from the waste form has been included in the risk assessment which shows that the radioactive emissions will be negligible (Annex B). Discharge to groundwater has been included in the risk assessment which shows that the potential emissions are very low and will not result in exceedance of relevant constraint limits (Annex B).

7.0.2 A specific risk assessment has been provided for any radioactive exposure or

environmental impact arising from leachate management practices (Annex B). Leachate is currently treated offsite. It is not reasonable to model the impact from leachate using the same conservative assumptions concerning leachate activity levels that are used for the groundwater modelling. The optimal approach for controlling any impact that may arise from leachate management is to place an authorisation limit on leachate activity levels such that the consequent impact is insignificant. Should the leachate ultimately have higher activity levels than the authorisation limit, then an alternative treatment would be considered which could include a detailed risk assessment and revised limit (see 8.6).


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8.0 Radioactive Waste Disposal Consequence Assessment and Radiological Capacity

8.0.1 To determine the amount of radiological material that can be disposed of within

the site without exceeding the proposed dose criteria a series of consequence and risk assessments have been undertaken. The consequence and risk assessment for the disposal of LLW at the East Northants Resource Management Facility is included in Annex B. Supplementary information on the assessment methodology is included in Annex E.

8.0.2 An additional assessment of direct radiation dose to workers is included in

Annexes D and H. 8.0.3 The primary assessment has been carried out using the SNIFFER 2006 model

for assessing the suitability of controlled landfills to accept disposals of solid low level radioactive waste. A range of 43 nuclides has been considered with their associated daughters. The methodology for the other supplementary assessments is described in the Annexes.

8.0.4 A summary of the scenarios considered follows: Scenario Annual Dose Criteria

Used for Assessment Public Worker Assessment Capacity

Constraint? Pre-Closure – expected to occur Direct Radiation Exposure from Waste Handling and Emplacement

20 mSv/yr Worker (Ionising Radiation Radiations) 1 mSv/yr Worker (Operational Criterion) 0.02 mSv/yr Public (GRA)- Lower Bound 0.3 mSv/yr Public (GRA) – Upper Bound

√ √ 8.1 Annex D and H

Not used to define landfill capacity

Exposure from Gas Generation from the Landfill


√ √ 8.2 Annex B (5.5)

Considered as a constraint to landfill capacity

Pre-Closure – not expected to occur Dropped Load of Waste (and hypothetical aircraft impact )

20 mSv/yr Worker (Ionising Radiation Radiations) 1 mSv/yr Worker (Operational Criterion) 0.02 mSv/yr Public

√ √ 8.3 Annex C



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Scenario Annual Dose Criteria Used for Assessment

Public Worker Assessment Capacity Constraint?

(GRA)- Lower Bound 0.3 mSv/yr Public (GRA) – Upper Bound 3 mSv/yr Public for aircraft intrusion

Wound Exposure 20 mSv/yr Worker (Ionising Radiation Radiations) 1 mSv/yr Worker (Operational Criterion)

x √ 8.4 Annex C


Exposure from Fire See discussion at 8.5

√ √ Not assessed Discussed in Annex B


Pre Closure and Aftercare Period – expected to occur Exposure from Leachate Processing Offsite – Sewage Works


√ √ 8.6 Annex B

Not Considered as a constraint to landfill capacity Used to set authorisation conditions for leachate discharge

Pre Closure and Aftercare Period – not certain to occur Exposure from Leachate - Spillage


√ √ 8.7 Annex B

Not used as a constraint on landfill capacity because worst case constraint is larger than the physical landfill

Exposure from Leachate - Aerosols


√ √ 8.8 Annex B

Not used as a constraint on landfill capacity or leachate discharge concentration because the case in section 8.6 is more constraining

Post-Closure – expected to occur Exposure by Using Groundwater at Nearest Abstraction Point

0.02 mSv/yr Public (GRA) √ x 8.9 Annex B

Considered as a potential constraint to landfill capacity


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Public Worker Assessment Capacity Constraint?

Exposure from Gas Generation from the Landfill

0.02 mSv/yr Public (GRA) √ x 8.10 Annex B

Considered as a potential constraint to landfill capacity

Exposure to Wildlife from all sources

10 microgray/hr x x 8.11 Annex B


External dose from emplaced wastes

0.02 mSv Public (GRA) √ x 8.12 Annex B

Not considered as a constraint to landfill capacity because the resulting doses are trivial

Post –Closure not expected to occur Exposure by Using Groundwater from a Borehole Constructed at the Boundary of the Landfill

3 mSv Public (GRA and HPA)

√ x 8.13 Annex B

Has the potential to be a constraint to landfill capacity

Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill

3 mSv Public or Worker (GRA and HPA)

√ √ 8.14 Annex B

Has the potential to be a constraint to landfill capacity

Other potential scenarios are discussed in Annex B and have not been assessed for documented reasons. 8.0.5 Some other potential scenarios are discussed in Annex B and have not been

assessed for documented reasons. Additionally:

- The aircraft crash scenario is discussed with the dropped load scenario.

- A scenario involving drilling into the waste form for construction of new sampling or leachate wells is not discussed because this would be executed with knowledge under appropriate regulations with appropriate precautions as necessary.

- The effects of very long term climate change are not assessed because the site is already permitted as a hazardous site and LLW disposal gives rise to no additional considerations in respect of flooding, coastal erosion or sea level rises. Future glaciation would have similar or lesser effects than the “residential intrusion scenario” considered in 8.14.

- The effects of seismic events. The engineered containment structures at the site are not formed of brittle materials such as concrete that may fracture as a result of a severe earthquake. The HDPE and clay lining materials have a high shear strength and have the flexibility to withstand


July 2009 59

the stresses which would be imposed during the types of earthquake which occur in the UK.

- Transport accident scenarios are not discussed because transport is outside of the scope of the authorisation and is regulated under an existing regime of Dangerous Goods Regulations. Transport accidents on the site are considered as part of the dropped load scenario and a transport accident involving leachate is specifically considered.

8.06 The full set of results is given in the Annexes and the following conclusions are

reached from the assessment:

8.1 Pre-Closure – expected to occur Direct Radiation Exposure from Waste Handling and Emplacement

8.1.1 The dose criteria are the legal limit to workers of 20 mSv/yr, the site criterion of 1

mSv/yr for workers, the dose guidance level of 0.02 mSv/yr for the public and the dose constraint for the public of 0.3 mSv/yr.

8.1.2 The impacted group is landfill workers and the public near to the site. The

emplaced waste can only affect the landfill workers because there is no line of sight for direct radiation from landfill void.

8.1.3 The assessment is contained within Annexes D and H. 8.1.4 Waste Emplacement: The scenario is the external radiation exposure of workers

in the vicinity of the waste emplaced in the landfill after it has been covered. 8.1.5 The assessment is contained within Annex H.

8.1.6 Annex H illustrates the dose rate for varying cover thicknesses using two

illustrative cases, one of which is a worst case. The advice of the radiation protection advisor is that the maximum radiation dose 1 m above the covered waste should be less than 2 microSv/hr in order to ensure the occupational dose is considerably less than the dose criterion of 1 mSv/yr.

8.1.7 Annex H demonstrates that for most cases a 300mm thick cover layer will more

than achieve the dose rate. For the worst case of waste containing Co-60 at 200 Bq/g a cover layer of 700mm would be required to achieve the dose rate, but this is exceptional.

8.1.8 The proposed authorisation condition is that a minimum cover layer of 300mm be

utilised and that if the dose rate 1 m above the waste is still greater than 2 microSv/hr then further cover will be added in order to achieve the dose rate. The minimum cover layer of 300mm is adequate to ensure daily physical protection of the waste.


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8.1.9 Additional ALARA precautions are that all wastes are handled by machines and

operatives generally do not enter the operational area on foot. On most days the only reason to enter the operational area on foot is for final inspection at the end of the day and health physics monitoring. Workplace monitoring will confirm actual doses and enable dose limitation to be managed.

8.1.10 The original SNIFFER model uses occupational external dose as a constraint to

set the radiological capacity of the landfill but since this dose is specific to workers during the operational phase and can be managed through occupational radiation dose protection practices this is not considered necessary. Hence the external dose assessment for waste emplacement has not been used to constrain the overall radiological capacity.

8.1.11 Waste Handling: The scenario is the external radiation exposure to workers from

their occupancy near to a waste package prior to disposal. 8.1.12 The SNIFFER model does not include this scenario and it has been assessed in

Annex D. Annex D considers the external radiation dose for a series of cases and package types. The hypothetical worst case is identified to be a waste flexible type container with 200 Bq/g of Co-60. A flexible container carrying Co-60 at 200 Bq/g is an unlikely case and another case is included in Annex D to illustrate more typical exposures.

8.1.13 The hypothetical worst case dose identified in Annex D is 14.5 microSv/hr at 1

metre from the package face. However the radiation protection advisor has advised that the maximum dose at 1 metre from a package should be less than 10 microSv/hr in order to ensure the occupational dose is considerably less than the dose criterion of 1 mSv/yr. Thus 10 microSv/hr will be used as an acceptance criterion and constrains the contents of the package to this limit.

8.1.14 The proposed authorisation condition is that the dose at 1 metre from the

package face must be less than 10 microSv/hr. This would be measured by the consignor prior to sending the package and would be checked by the consignee upon arrival of the package.

8.1.15 Additional ALARA precautions are that dose can be measured directly and

managed actively to prevent unnecessary exposure. As illustrated in Annex D the field dose drops quickly with distance from the package and hence the simple precaution of managing occupancy time and distance is practicable.

8.1.16 This dose is specific to workers during the operational phase and can be

managed through occupational radiation dose protection practices, hence it is not used to constrain overall radiological capacity.

8.1.17 There is an additional scenario that a member of the public stands at a distance

in direct line of sight of a waste package/shipment and hence receives direct radiation exposure. This can be estimated by considering the waste as a single


July 2009 61

point source with a 10 microSv/hr doserate at 1 metre, assuming that the member of the public is located 50 metres from the waste. The doserate at 50 metres can be estimated from:

D1=D2 (X2

2/X12)

Where, D1 and D2 = doserate at positions 1 and 2 X1 and X2 = the distance from the source at positions 1 and 2 This gives an estimated maximum doserate at 50 metres of 4E-3 microSv/hr. If

the person stands in that location for 8 hours per day and there is waste at the maximum level in that location every day then the person would receive 12 microSv per year. This is based on conservative assumptions and is within the 20 microSv per year lower bound dose criterion.

Under the same assumptions but with a 100 metre distance to the person, the

maximum estimated dose would be 3 microSv/yr. These calculations do not take into account the significant shielding afforded by

the soil screen bund at the boundary of the site.

8.2 Pre-Closure – expected to occur Exposure from Gas Generation from the Landfill



8.2.2 The impacted groups during the pre-closure phase are the public and workers. 8.2.3 The assessment is contained within Annex B, section 5.5. 8.2.4 The scenario is radioactive gas release from the landfill in the pre-closure phase.

Annex B indicates that the worst case is for Ra-226 in both the case of the worker and the public. For this worst case and using a worker dose criterion of 1 mSv/yr the capacity of the landfill to take only Ra-226 would be 8.4E9 MBq (42 million tonnes at 200 Bq/g). For this worst case and using a public dose criterion of 0.02 mSv/yr the capacity of the landfill to take only Ra-226 would be 81E6 MBq (403 thousand tonnes at 200 Bq/g).

8.2.5 This scenario has the potential to constrain landfill capacity for the public

exposure case but not for the worker case (because the landfill is physically smaller than the radiological capacity).


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8.3 Pre-Closure – not expected to occur Dropped Load of Waste

Dropped Load 8.3.1 The dose criteria are the legal limit to workers of 20 mSv/yr, the site criterion of 1


8.3.2 The impacted groups during the pre-closure phase are the public and workers. 8.3.3 The scenario is not contained within the SNIFFER model and has been

separately addressed in Annex C, which is a radiological risk assessment for occupational exposure completed by the HPA.

8.3.4 The conclusion is that with appropriate precautions the worker exposure can be

kept with the site criterion under the unlikely circumstance of a dropped container which gave rise to a release.

8.3.4 This scenario is not used to constrain landfill capacity. 8.3.5 To augment the calculations in Annex C the following table gives exposure to

both workers and the public under the following assumptions using the UKAEA dropped load methodology from the safety assessment handbook (ref 22).

8.3.6 The assumptions are:

- A one cubic metre flexible container of wastes is dropped and spills 10% of its contents through broken seams.

- The bag is filled with a dry solid. - The bag contains the maximum concentration of a single nuclide at 200

Bg/g. - The bag weighs 1 tonne. - The distance to the nearest public is 50m and the event duration is 30

minutes. - The worker remains very close to the dropped waste without taking

precautions or retreating for at least 30 minutes. - The atmospheric conditions are worst case, still conditions.


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8.3.7 Dose from inhaling material discharged from a dropped container is given by:

DF

DBCRFRFIDose

Rninh

baginh

21,

where I is the inventory of radionuclide Rn releasable (Bq) RF1 is the release fraction (-) RF2 is the respirable fraction (-) C is the dispersion coefficient (s m-3). B is the breathing rate (m3 s-1) DF is the decontamination factor (-) Dinh is the dose coefficient for inhalation of radionuclide, Rn (Sv Bq-1). Parameter Description Value Units inventory of radionuclide in the bag 200E6 Bq I inventory of radionuclide Rn

releasable 20E6 Bq

RF1 release fraction 1E-3 - RF2 respirable fraction 0.1 -

Worker 5 C dispersion coefficient Public 1.7E-2

s m-3

B breathing rate 3.3E-4 m3 s-1 DF decontamination factor 1 -


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Doses from Dropped Load Scenario

Radionuclide

Inhalation Dose Coefficent (Sv/Bq)

Worker Dose (microSv)

Public Dose (microSv)

H-3 2.60E-10 8.58E-04 2.92E-06 C-14 5.80E-09 1.91E-02 6.51E-05 Cl-36 7.30E-09 2.41E-02 8.19E-05 Fe-55 7.70E-10 2.54E-03 8.64E-06 Co-60 3.10E-08 1.02E-01 3.48E-04 Ni-63 4.80E-10 1.58E-03 5.39E-06 Sr-90 1.62E-07 5.35E-01 1.82E-03 Nb-94 1.10E-08 3.63E-02 1.23E-04 Tc-99 1.30E-08 4.29E-02 1.46E-04 Ru-106 6.60E-08 2.18E-01 7.41E-04 Ag-108m 3.70E-08 1.22E-01 4.15E-04 Sb-125 5.46E-09 1.80E-02 6.13E-05 Sn-126 3.12E-08 1.03E-01 3.50E-04 I-129 3.60E-08 1.19E-01 4.04E-04 Ba-133 3.10E-09 1.02E-02 3.48E-05 Cs-134 6.80E-09 2.24E-02 7.63E-05 Cs-137 3.90E-08 1.29E-01 4.38E-04 Pm-147 5.00E-09 1.65E-02 5.61E-05 Eu-152 4.20E-08 1.39E-01 4.71E-04 Eu-154 5.30E-08 1.75E-01 5.95E-04 Eu-155 6.90E-09 2.28E-02 7.74E-05 Pb-210 9.99E-06 3.30E+01 1.12E-01 Ra-226 1.95E-05 6.44E+01 2.19E-01 Ac-227 5.69E-04 1.88E+03 6.38E+00 Th-229 2.56E-04 8.45E+02 2.87E+00 Th-230 1.00E-04 3.30E+02 1.12E+00 Th-232 1.70E-04 5.61E+02 1.91E+00 Pa-231 1.40E-04 4.62E+02 1.57E+00 U-232 4.69E-05 1.55E+02 5.26E-01 U-233 9.60E-06 3.17E+01 1.08E-01 U-234 9.40E-06 3.10E+01 1.05E-01 U-235 8.50E-06 2.81E+01 9.54E-02 U-236 3.20E-06 1.06E+01 3.59E-02 U-238 8.01E-06 2.64E+01 8.99E-02 Np-237 5.00E-05 1.65E+02 5.61E-01 Pu-238 1.10E-04 3.63E+02 1.23E+00 Pu-239 1.20E-04 3.96E+02 1.35E+00 Pu-240 1.20E-04 3.96E+02 1.35E+00 Pu-241 2.30E-06 7.59E+00 2.58E-02 Pu-242 1.10E-04 3.63E+02 1.23E+00 Am-241 9.60E-05 3.17E+02 1.08E+00 Cm-243 3.11E-05 1.03E+02 3.49E-01 Cm-244 2.71E-05 8.94E+01 3.04E-01

8.3.8 Only in the case of Ac-227 does the assessment fail to meet the site criterion for

workers. Ac-227 is very unlikely to be present at 200 Bg/g given the low


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occurrence of this nuclide. The above calculations assume that the bag is filled with a loose dry material that disperses readily, that the package fails and that the worker does not respond correctly. These are highly conservative assumptions, especially given the operational precautions proposed in Section 5.6.

8.3.9 A key measure to mitigate dropped load dispersion events will be to engineer the waste containers such that they withstand or substantially withstand accidental drops during handling. Where drums are used these will be rated under existing dangerous good transport regulations for radioactive material to withstand a drop test. Flexible containers may only be used where this is acceptable under dangerous goods transport regulations and these regulations specify isotope specific limits designed to ensure public safety.

8.3.10 The dropped bag scenario is not used to establish radiological capacity of the landfill because it is independent of the total tonnage received.

Aircraft Impact 8.3.11 The event could be considered an intrusion in which case the 3-20 mSv/yr dose

criteria would apply. 8.3.12 The impacted groups during the pre-closure phase are the public and workers.

The event is assessed for the pre-closure phase but could also apply to the post-closure phase for the public if the landfill closure cap (at least 1.5m thick) did not provide full protection from the impact.

8.3.13 The scenario is not contained within the SNIFFER model and has been

separately addressed below. 8.3.14 This scenario is not used to constrain landfill capacity because it is independent

of tonnage received. The scenario has a very low probability of occurrence. 8.3.15 The following gives exposure to both workers and the public under the following

assumptions using the UKAEA release methodology from the safety assessment handbook (ref 22). The approach used is to assume an amount of material is physically displaced by crater formation through impact of a high velocity military aircraft. This is considered a reasonable scenario given the presence of an RAF base close to the landfill when compared to much less likely scenarios involving heavy civilian aircraft.

8.3.16 Due to the complexity of such an event this assessment can only be considered

as a scoping calculation based on conservative assumptions. 8.3.17 The assumptions are:

- The aircraft hits an area of exposed waste and forms a crater.


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- The crater size can be estimated from theoretical models for estimated impact parameters such as densities, impact velocity, impact angle, missile dimensions and target density/type (Ref 21). Scoping calculations indicate that crater sizes of 300 cum are conceivable. Actual crater sizes from impacts due to Harrier jets (the type of aircraft currently based at RAF Wittering) reveal a wide variation from virtually no displacement to significant craters dependent on the nature of the event. A record (ref 23) notes a Harrier jet impact forming a crater of approximately 300 cum. For comparison, the Lockerbie B747 impact formed a crater of 560 cum (ref 24).

- The displaced waste contains the maximum concentration of a single nuclide at 200 Bg/g. The chosen nuclide is Pu-239 which has a conservative inhalation dose coefficient and yet is a credible nuclide to occur at the concentration assumed.

- The density of the displaced waste is 1.5 t/cum. 300 cum or 450 tonnes are displaced. Giving rise to displacement of 90,000 MBq.

- The distance to the nearest public is 200m and the event has 30 minute release duration. This is on the basis that immediate evacuation of the near zone would occur from such an extreme event and within the very near zone immediate fatality due to impact would be likely.

- The effect of fire on dispersal is not included (refer Section 8.5)..

- The worker exposure is the same as the public exposure because workers would evacuate quickly to the same distance.

- The atmospheric conditions are worst case still conditions and mixing is not assumed to be enhanced by fire.

8.3.18 Dose from inhaling material discharged from displaced material:

DF

DBCRFRFIDose

Rninh

baginh

21,

where I is the inventory of radionuclide Rn releasable (Bq) RF1 is the release fraction (-) RF2 is the respirable fraction (-) C is the dispersion coefficient (s m-3). B is the breathing rate (m3 s-1) DF is the decontamination factor (-) Dinh is the dose coefficient for inhalation of radionuclide, Rn (Sv Bq-1).


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Parameter Description Value Units inventory of radionuclide 90,000E6 Bq RF1 release fraction 1E-3 - RF2 respirable fraction 0.1 -

Worker 1.5 E-3 C dispersion coefficient Public 1.5 E-3

s m-3

B breathing rate 3.3E-4 m3 s-1 DF decontamination factor 1 -

8.3.19 The resulting dose would be approximately 0.5 mSv. Such a calculation could

have a relatively wide range of uncertainty, but this conservative scoping estimate indicates that public and worker legal dose limits would not be exceeded and the 3 mSv/yr intrusion dose limit would not be exceeded by this low probability extreme event.

8.4 Pre-Closure – not expected to occur Wound Exposure 8.4.1 The dose criteria are the legal limit to workers of 20 mSv/yr and the site criterion

of 1 mSv/yr for workers. 8.4.2 The impacted groups during the pre-closure phase are the landfill site workers. 8.4.3 The scenario is not contained within the SNIFFER model and has been

separately addressed in Annex C, which is a radiological risk assessment for occupational exposure completed by the HPA.

8.4.4 The conclusion is that wound exposures are unlikely and can be further reduced

in likelihood and impact through simple precautions. It is very likely this will be effective in maintaining individual exposures within the site criterion.

8.4.5 This scenario is not used to constrain landfill capacity


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8.5 Pre-Closure – not expected to occur Exposure from Fire

8.5.1 The dose criteria are not defined within the guidance on requirements for

authorisation for this scenario but could be taken to be the worker dose target of 1 mSv/yr and the public dose constraint of 0.3 mSv/yr.

8.5.2 The impacted groups during the pre-closure phase are the public and workers. 8.5.3 The scenario is transient and for practical purposes can only occur whilst the

wastes are not covered with the final capping layer. The lack of biodegradable wastes makes fires very unlikely after the cap is in place.

8.5.4 Furthermore, some of the exposure pathways considered in the SNIFFER fire

model would not arise in practice because intervention would occur. So for example exposures from any subsequent deposition would be controlled by remedial activities.

8.5.5 Although an aircraft crash could lead to a fire, the fire would mostly consume

aircraft fuel and wreckage. The main feature of an aircraft impact which could lead to exposure would be the physical displacement of material and this is considered in Section 8.3.

8.5.6 The waste in the landfill, the cover materials and the LLW are essentially

incombustible. The current waste acceptance criterion for the landfill largely excludes organic material and includes a flammability test. As such it is difficult to conceive that the fire scenario included in the SNIFFER model can occur for this type of landfill and it has not been utilised to constrain landfill capacity.

8.6 Pre Closure and Aftercare Period – expected to occur Exposure from Leachate Processing Offsite – Sewage Works



8.6.2 The impacted groups are sewage workers and the public impacted by the

sewage works. 8.6.3 The scenario is addressed in Annex B. 8.6.4 Leachate levels at the ENRMF are maintained by pumping excess leachate to

tankers and transporting this leachate to a water treatment plant at Avonmouth. An initial assessment of the potential impacts from routine, off-site leachate management has been made using the Environment Agency’s methodology and


July 2009 69

the assumption that doses from water treatment would be similar to doses from sewage treatment. The Environment Agency’s methodology allows for a range of exposure groups affected by releases to a public sewer, depending on the discharge route for treated effluent. For this assessment, only the groups associated directly with operation of the treatment plant, farming of land conditioned by sludge or using the estuary are considered. These groups and the relevant exposure pathways are:

Sewage treatment workers (adults only)

External irradiation from radionuclides in raw sewage and sludge Inadvertent inhalation and ingestion of raw sewage and sludge containing radionuclides

Farming family living on land conditioned with sewage sludge Consumption of food produced on land conditioned with sludge and incorporating radionuclides External irradiation from radionuclides in sludge conditioned soil Inadvertent inhalation and ingestion of sludge conditioned soil

Fisherman’s family (estuary/coastal water receives treated effluent from sewage works, typically via a river)

External irradiation from radionuclides deposited in sediments Consumption of fish incorporating radionuclides

8.6.5 The results in Annex B are expressed as specific dose per MBq/year of activity in

the leachate treated. The groundwater assessment uses a very pessimistic assumption regarding leachate concentrations which it is not appropriate to use in this assessment. There is no empirical evidence on which to base leachate activity concentrations.

8.6.6 The worst case result is for Th-232 with the “farming family” public exposure

group. If we assume a dose constraint of 0.3 mSv/yr for the public during the operational phase, the results indicate that the maximum allowable leachate discharge per year is 216 MBq if the leachate only comprised Th-232.

8.6.7 The proposed approach is that this scenario is not used to constrain radiological

capacity, but that it is used to derive authorisation discharge limits for the leachate which can then be subsequently refined when empirical monitoring results become available. Based upon the above approach the authorisation limits for individual nuclides if the leachate comprised 100% of that nuclide, with a 0.3 mSv/y dose criterion are given in the following table. In practice authorisation discharge limits will be set after discussion with the Environment Agency and then optimized through operational experience such that restrictive dose criterion are met.


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Maximum Leachate Authorisation Limits for Individual Nuclides

Nuclide Farming family Specific dose

(microSv / y per MBq / y) Leachate Authorisation Limit

MBq/yr

H-3 2.83E-06 106E+6

C-14 4.72E-03 63E+3

Cl-36 7.78E-02 3.9E+3

Fe-55 1.33E-03 225E+3

Co-60 7.78E-01 385

Sr-90 2.17E-02 14E+3

Tc-99 2.83E-01 1060

Ru-106 3.06E-03 98E+3

I-129 6.11E-02 4909

Cs-134 1.17E-01 2564

Cs-137 1.00E-01 3000

Pm-147 1.67E-05 18E+6

Eu-152 2.67E-01 1123

Eu-154 2.72E-01 1102

Eu-155 5.06E-03 59E+3

Pb-210 5.33E-01 562

Ra-226 5.56E-01 540

Th-230 1.28E-02 23E+3

Th-232 1.39E+00 215

U-234 1.17E-03 256E+3

U-235 7.78E-03 38E+3

U-238 2.06E-03 145E+3

Np-237 7.22E-02 4155

Pu-238 2.00E-02 15E+3

Pu-239 2.28E-02 13E+3

Pu-240 2.28E-02 13E+3

Pu-241 3.39E-04 885E+3

Pu-242 2.22E-02 13E+3

Am-241 3.94E-02 7614

Cm-243 6.67E-02 4497

Cm-244 1.78E-02 16E+3


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8.7 Pre Closure and Aftercare Period – not certain to occur Exposure from Leachate - Spillage



8.7.2 Notwithstanding any radioactive components, landfill leachate poses a hazard to

the environment if spilt and any road accident involving loss of an entire load would be subject to mitigation measures. Leachate that did enter water resources would also become diluted. For this assessment, it is conservatively assumed that an entire tanker load of leachate (30 m3 of leachate) reaches a small reservoir (2 x 106 m3) that is used for drinking water, irrigation and fishing.

8.7.3 The scenario is addressed in Annex B, table 5.5. 8.7.4 The worst case is if the leachate comprises only Ra-226. The public dose

constraint of 0.3 mSv can be used because this event is low probability and clean-up actions would in reality be taken to largely mitigate the event altogether. The resulting radiological capacity for the crops exposure case is then 71E6 MBq or 356,000 tonnes at 200 Bq/g. Given the mitigation measures noted above this scenario does not constrain radiological capacity.

8.8 Pre Closure and Aftercare Period – not certain to occur

Exposure from Aerosols 8.8.1 The dose criteria are the legal limit to workers of 20 mSv/yr, the site criterion of 1


8.8.2 There is potential, during leachate management or spillage, for the production of

aerosols which could lead to doses via the inhalation pathway. 8.8.3 The assessment is presented in Annex B, table 5.6. The results are presented in

terms of specific dose (µSv y-1 per MBq per hour). The worst case result is if the leachate aerosol comprises only Ac-227, for public exposure. Assuming 1600 hours exposure per year and a dose guidance level of 0.02 mSv/yr, the resulting leachate discharge limit would be 7002 MBq/yr. This case results in levels which are orders of magnitude higher than the case considered in section 8.6 above and hence this scenario is not a constraint on either radiological capacity or leachate discharge concentration.


July 2009 72

8.9 Post-Closure – expected to occur Exposure by Using Groundwater at Nearest Abstraction Point

8.9.1 The dose criterion is the dose guidance level of 0.02 mSv/yr for the public. 8.9.2 The scenario only impacts members of the public. 8.9.3 The scenario is assessed in Annex B, table 5.1 (1500m irrigation case). 8.9.4 If a well or river is used for irrigation, then doses can result from ingestion of

foodstuffs raised on contaminated soil, inhalation of dust from the soil, and external exposure to the soil. Drinking of contaminated water from a well or river is also a potential exposure pathway. If contaminated groundwater discharges to surface water (spring, river, sea), then ingestion of foodstuffs from the surface water is a potential exposure pathway. This scenario considers such exposure from the nearest abstraction point.

8.9.5 The following table shows the results of the assessment on the assumption of

using the public dose guidance level of 0.02 mSv/yr and the tonnage capacities assuming 200 Bq/g of each nuclide. These capacities are for the case where all of the waste in the landfill is comprised of the single nuclide at the maximum concentration. The information has been sorted is order of ascending capacity with the most restricted capacity nuclides at the top of the table.

8.9.6 With the exception of I-129 and Np-237 the tonnage capacities are larger than

the physical size of the landfill void. This scenario has the potential to constrain landfill capacity.

Groundwater Pathway, Borehole, 1500m Irrigation Case (Table 5.1 Annex B)

Radionuclide

Specific Dose (microSv/yr per MBq)

Radiological Capacity (MBq) (0.02 mSv/yr dose criterion)

Tonnage Capacity (tonnes) (@200 Bq/g concentration)

I-129 1.38E-05 1.45E+06 7.25E+03

Np-237 1.22E-06 1.64E+07 8.20E+04

Cl-36 6.52E-08 3.07E+08 1.53E+06

Th-232 4.04E-08 4.95E+08 2.48E+06

Pa-231 3.60E-08 5.56E+08 2.78E+06

Pu-242 1.69E-08 1.18E+09 5.92E+06

Ra-226 1.56E-08 1.28E+09 6.41E+06

Pu-239 1.54E-08 1.30E+09 6.49E+06

Th-229 1.44E-08 1.39E+09 6.94E+06

Sn-126 1.20E-08 1.67E+09 8.33E+06


July 2009 73

Groundwater Pathway, Borehole, 1500m Irrigation Case (Table 5.1 Annex B)

Radionuclide




Pu-240 1.04E-08 1.92E+09 9.62E+06

Th-230 8.24E-09 2.43E+09 1.21E+07

U-233 4.62E-09 4.33E+09 2.16E+07

U-234 4.35E-09 4.60E+09 2.30E+07

U-235 4.35E-09 4.60E+09 2.30E+07

U-238 4.35E-09 4.60E+09 2.30E+07

U-236 4.22E-09 4.74E+09 2.37E+07

Tc-99 2.12E-09 9.43E+09 4.72E+07

C-14 2.06E-09 9.71E+09 4.85E+07

Cm-244 1.29E-09 1.55E+10 7.75E+07

Nb-94 3.76E-10 5.32E+10 2.66E+08

Cm-243 1.82E-10 1.10E+11 5.49E+08

Pu-241 1.59E-10 1.26E+11 6.29E+08

Am-241 5.37E-11 3.72E+11 1.86E+09

Pu-238 1.86E-12 1.08E+13 5.38E+10

Ag-108m 1.68E-17 1.19E+18 5.95E+15

U-232 5.07E-20 3.94E+20 1.97E+18

Ni-63 7.94E-21 2.52E+21 1.26E+19

Sr-90 3.04E-24 6.58E+24 3.29E+22

Cs-137 1.28E-25 1.56E+26 7.81E+23

Pb-210 1.25E-25 1.60E+26 8.00E+23

Ac-227 5.66E-26 3.53E+26 1.77E+24

H-3 3.66E-30 5.46E+30 2.73E+28

Ba-133 1.44E-31 1.39E+32 6.94E+29

Eu-152 2.77E-32 7.22E+32 3.61E+30

Eu-154 3.28E-35 6.10E+35 3.05E+33

Co-60 1.21E-39 1.65E+40 8.26E+37

Eu-155 3.63E-41 5.51E+41 2.75E+39

Sb-125 4.81E-42 4.16E+42 2.08E+40

Cs-134 1.82E-43 1.10E+44 5.49E+41

Fe-55 1.04E-43 1.92E+44 9.62E+41

Pm-147 3.41E-44 5.87E+44 2.93E+42

Ru-106 3.44E-46 5.81E+46 2.91E+44


July 2009 74

8.10 Post-Closure – expected to occur Exposure from Gas Generation from the Landfill

8.10.1 The dose criterion is the dose guidance level of 0.02 mSv/yr for the public. 8.10.2 The scenario is relevant to members of the public (post closure workers are

treated as members of the public in this case). 8.10.3 The scenario is assessed in Annex B, table 5.7, the “resident after closure case”. 8.10.4 The scenario is the release of radioactive gas in the post-closure phase. The

results indicate that the worst case is for Ra-226 which using the 0.02 mSv/yr dose criterion gives a capacity of 7E6 MBq or 35,000 tonnes at 200 Bq/g.

8.10.5 This scenario has the potential to constrain landfill capacity.

8.11 Post-Closure – expected to occur Exposure to Wildlife from all sources

8.11.1 A dose criterion for screening purposes of 10 microGy/hr has been used in accordance with the ERICA tool.

8.11.2 The scenario relates to a representative range of organisms and wildlife groups. 8.11.3 The scenario is the release of radionuclides into the environment. A set of

pessimistic assumptions have been used for a hypothetical release. 8.11.4 The scenario is described and assessed in Annex B, section 5.7. Tables 5.10,

5.11 and 5.12 give the results. 8.11.5 The conclusion is that the exposure to the range of organisms and wildlife groups

is below the screening dose criteria and therefore does not need further assessment. This scenario does not constrain landfill capacity

8.12 Post-Closure – expected to occur External dose from emplaced wastes

8.12.1 The dose criterion is the dose guidance level of 0.02 mSv/yr for the public. 8.12.2 The scenario applies only to members of the public post-closure. 8.12.3 The scenario is that persons walking on the closed waste site will experience

direct radiation exposure through the cover materials. The scenario is included in the SNIFFER model and is assessed in Annex B, table 5.2, “Public dose 1.5m


July 2009 75

cap”. The assumption is that the waste is shielded by a 1.5m thick cap, although in practice this is likely to be a conservative assumption.

8.12.4 The results shows that the resulting doses from a worst case where the landfill is

filled with the worst case nuclide at the maximum concentration, are far below the dose criterion. This case does not constrain landfill capacity.

8.13 Post –Closure not expected to occur Exposure by Using Groundwater from a Borehole Constructed at the Boundary of the Landfill

8.13.1 The dose criterion is the lower dose guidance level of 3 mSv/yr for the public for

an intrusion scenario. 8.13.2 The scenario applies only to members of the public post-closure and is unlikely to

occur given the presence of the hazardous landfill. 8.13.3 The scenario is that a new groundwater abstraction point is licensed at the

boundary of the landfill site. The scenario is assessed in Annex B, table 5.1”Site boundary drinking”.

8.13.4 The resulting radiological capacities and tonnage capacities (based on the

maximum concentration of 200 Bq/g) are shown below in order of ascending capacity with the most restricted capacity nuclides at the top of the table.

8.13.5 With the exception of I-129 and Np-237 the tonnage capacities are larger than the

physical size of the landfill void. This scenario has the potential to constrain landfill capacity.


July 2009 76

Groundwater Pathway, Site Boundary Drinking Case, (Table 5.1 Annex B)

Radionuclide


Radiological Capacity (MBq) (3 mSv/yr dose criterion)


I-129 3.02E-04 9.93E+06 4.97E+04

Np-237 9.52E-05 3.15E+07 1.58E+05

Th-232 3.59E-06 8.36E+08 4.18E+06

Pa-231 3.08E-06 9.74E+08 4.87E+06

Cl-36 1.69E-06 1.78E+09 8.88E+06

Pu-242 1.51E-06 1.99E+09 9.93E+06

Pu-239 1.37E-06 2.19E+09 1.09E+07

Ra-226 1.36E-06 2.21E+09 1.10E+07

Th-229 1.29E-06 2.33E+09 1.16E+07

Pu-240 9.33E-07 3.22E+09 1.61E+07

Th-230 7.35E-07 4.08E+09 2.04E+07

Sn-126 4.87E-07 6.16E+09 3.08E+07

U-233 4.13E-07 7.26E+09 3.63E+07

U-234 3.89E-07 7.71E+09 3.86E+07

U-238 3.89E-07 7.71E+09 3.86E+07

U-235 3.87E-07 7.75E+09 3.88E+07

U-236 3.78E-07 7.94E+09 3.97E+07

Tc-99 1.52E-07 1.97E+10 9.87E+07

C-14 1.39E-07 2.16E+10 1.08E+08

Cm-244 1.15E-07 2.61E+10 1.30E+08

Pu-241 3.62E-08 8.29E+10 4.14E+08

Cm-243 1.63E-08 1.84E+11 9.20E+08

Am-241 1.08E-08 2.78E+11 1.39E+09

Nb-94 9.07E-09 3.31E+11 1.65E+09

Pu-238 1.67E-10 1.80E+13 8.98E+10

Ag-108m 1.51E-12 1.99E+15 9.93E+12

U-232 6.19E-14 4.85E+16 2.42E+14

Ni-63 3.47E-15 8.65E+17 4.32E+15

Sr-90 2.19E-17 1.37E+20 6.85E+17

Pb-210 1.43E-18 2.10E+21 1.05E+19

Cs-137 9.32E-19 3.22E+21 1.61E+19

Ac-227 6.70E-19 4.48E+21 2.24E+19


July 2009 77

Groundwater Pathway, Site Boundary Drinking Case, (Table 5.1 Annex B)

Radionuclide




H-3 6.89E-23 4.35E+25 2.18E+23

Ba-133 3.25E-24 9.23E+26 4.62E+24

Eu-152 4.84E-25 6.20E+27 3.10E+25

Eu-154 7.98E-28 3.76E+30 1.88E+28

Co-60 3.79E-32 7.92E+34 3.96E+32

Eu-155 1.29E-33 2.33E+36 1.16E+34

Sb-125 2.03E-34 1.48E+37 7.39E+34

Cs-134 8.12E-36 3.69E+38 1.85E+36

Fe-55 4.45E-36 6.74E+38 3.37E+36

Pm-147 1.47E-36 2.04E+39 1.02E+37

Ru-106 1.67E-38 1.80E+41 8.98E+38

8.14 Post –Closure not expected to occur Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill

8.14.1 The dose criterion is the lower dose guidance level of 3 mSv/yr for the public and workers for an intrusion scenario.

8.14.2 The scenario applies to members of the public and workers post-closure and is

uncertain to occur. 8.14.3 The scenario is assessed in Annex B, table 5.3 in the “Intruder, 60 years case

and the Resident 60 years case”. The scenario is that either workers or members of the public intrude into the waste. For the public case the scenario includes residence on the waste material after intrusion.

8.14.4 The resulting radiological capacities and tonnage capacities (based on the

maximum concentration of 200 Bq/g) are shown below in order of ascending capacity with the most restricted capacity nuclides at the top of the table.

8.14.5 These scenarios have the potential to constrain landfill capacity.


July 2009 78

Intrusion (Table 5.3, Annex B) "Intruder 60 year Case"

Radionuclide




Th-232 1.87E-04 1.60E+07 8.02E+04

Sn-126 1.34E-04 2.24E+07 1.12E+05

Ra-226 1.08E-04 2.78E+07 1.39E+05

Th-229 9.48E-05 3.16E+07 1.58E+05

Nb-94 7.52E-05 3.99E+07 1.99E+05

Pa-231 6.82E-05 4.40E+07 2.20E+05

Ag-108m 5.13E-05 5.85E+07 2.92E+05

Pu-239 3.84E-05 7.81E+07 3.91E+05

Pu-240 3.82E-05 7.85E+07 3.93E+05

Pu-242 3.54E-05 8.47E+07 4.24E+05

Th-230 3.49E-05 8.60E+07 4.30E+05

Am-241 2.79E-05 1.08E+08 5.38E+05

Np-237 2.47E-05 1.21E+08 6.07E+05

Ac-227 2.14E-05 1.40E+08 7.01E+05

Pu-238 2.03E-05 1.48E+08 7.39E+05

Pu-241 1.47E-05 2.04E+08 1.02E+06

U-232 8.99E-06 3.34E+08 1.67E+06

U-235 8.96E-06 3.35E+08 1.67E+06

Cs-137 5.60E-06 5.36E+08 2.68E+06

U-233 3.89E-06 7.71E+08 3.86E+06

U-238 3.88E-06 7.73E+08 3.87E+06

U-234 3.29E-06 9.12E+08 4.56E+06

Cm-243 3.03E-06 9.90E+08 4.95E+06

Pb-210 2.19E-06 1.37E+09 6.85E+06

Cm-244 1.57E-06 1.91E+09 9.55E+06

Eu-152 1.42E-06 2.11E+09 1.06E+07

U-236 1.38E-06 2.17E+09 1.09E+07

I-129 1.06E-06 2.83E+09 1.42E+07

Eu-154 2.41E-07 1.24E+10 6.22E+07

Ba-133 1.65E-07 1.82E+10 9.09E+07

Sr-90 9.59E-08 3.13E+10 1.56E+08


July 2009 79

Intrusion (Table 5.3, Annex B) "Intruder 60 year Case"

Radionuclide




Cl-36 3.02E-08 9.93E+10 4.97E+08

Co-60 1.27E-08 2.36E+11 1.18E+09

Tc-99 1.09E-08 2.75E+11 1.38E+09

C-14 7.05E-09 4.26E+11 2.13E+09

Ni-63 8.68E-10 3.46E+12 1.73E+10

Eu-155 8.04E-11 3.73E+13 1.87E+11

H-3 4.52E-12 6.64E+14 3.32E+12

Sb-125 5.68E-13 5.28E+15 2.64E+13

Cs-134 6.83E-15 4.39E+17 2.20E+15

Fe-55 4.85E-17 6.19E+19 3.09E+17

Pm-147 3.54E-17 8.47E+19 4.24E+17

Ru-106 1.39E-26 2.16E+29 1.08E+27


July 2009 80

Intrusion (Table 5.3, Annex B) "Resident 60 year Case"

Radionuclide




Ra-226 5.40E-06 5.56E+08 2.78E+06

Pa-231 1.66E-06 1.81E+09 9.04E+06

I-129 1.36E-06 2.21E+09 1.10E+07

Sn-126 4.90E-07 6.12E+09 3.06E+07

Th-232 4.75E-07 6.32E+09 3.16E+07

Cl-36 4.28E-07 7.01E+09 3.50E+07

Tc-99 3.68E-07 8.15E+09 4.08E+07

Sr-90 3.31E-07 9.06E+09 4.53E+07

Nb-94 2.44E-07 1.23E+10 6.15E+07

Ag-108m 1.68E-07 1.79E+10 8.93E+07

Th-230 1.56E-07 1.92E+10 9.62E+07

Pb-210 1.28E-07 2.34E+10 1.17E+08

Th-229 8.84E-08 3.39E+10 1.70E+08

Np-237 4.65E-08 6.45E+10 3.23E+08

Cs-137 2.61E-08 1.15E+11 5.75E+08

U-235 2.47E-08 1.21E+11 6.07E+08

Pu-239 1.86E-08 1.61E+11 8.06E+08

Pu-240 1.85E-08 1.62E+11 8.11E+08

Ac-227 1.81E-08 1.66E+11 8.29E+08

Pu-242 1.76E-08 1.70E+11 8.52E+08

U-232 1.57E-08 1.91E+11 9.55E+08

Am-241 1.57E-08 1.91E+11 9.55E+08

Pu-238 9.86E-09 3.04E+11 1.52E+09

C-14 8.68E-09 3.46E+11 1.73E+09

Pu-241 8.29E-09 3.62E+11 1.81E+09

U-238 6.87E-09 4.37E+11 2.18E+09

Eu-152 4.63E-09 6.48E+11 3.24E+09

Cm-243 4.49E-09 6.68E+11 3.34E+09

U-233 4.32E-09 6.94E+11 3.47E+09

U-234 3.71E-09 8.09E+11 4.04E+09

U-236 3.14E-09 9.55E+11 4.78E+09

Cm-244 9.45E-10 3.17E+12 1.59E+10


July 2009 81

Intrusion (Table 5.3, Annex B) "Resident 60 year Case"

Radionuclide




Eu-154 7.87E-10 3.81E+12 1.91E+10

Ba-133 5.63E-10 5.33E+12 2.66E+10

Ni-63 2.60E-10 1.15E+13 5.77E+10

H-3 1.28E-10 2.34E+13 1.17E+11

Co-60 4.19E-11 7.16E+13 3.58E+11

Eu-155 2.74E-13 1.09E+16 5.47E+13

Sb-125 1.89E-15 1.59E+18 7.94E+15

Cs-134 2.76E-17 1.09E+20 5.43E+17

Fe-55 8.13E-18 3.69E+20 1.85E+18

Pm-147 4.55E-19 6.59E+21 3.30E+19

Ru-106 7.44E-29 4.03E+31 2.02E+29


July 2009 82

8.15 Results of the Assessment Scenario Annual Dose Criteria

Used for Assessment Assessment Results

Direct Radiation Exposure from Waste Handling and Emplacement


8.1 Annex D and H

Exposure from the emplaced wastes is constrained by a site rule limiting dose rate. Exposure from handling waste packages is constrained by a site rule limiting dose rate.

Exposure from Gas Generation from the Landfill – Pre Closure


8.2 Annex B (5.5)

This scenario may limit radiological capacity for certain cases.

Dropped Load of Waste (and hypothetical aircraft impact )


8.3 Annex C

Worker exposure and public exposure is within dose targets.

Wound Exposure 20 mSv/yr Worker (Ionising Radiation Radiations) 1 mSv/yr Worker (Operational Criterion)

8.4 Annex C

Worker exposure is within dose targets.

Exposure from Fire See discussion at 8.5

Annex B Not used to limit radiological capacity because the waste is essentially incombustible.

Exposure from Leachate Processing Offsite – Sewage Works


8.6 Annex B

Will be used to establish leachate discharge concentration limits.

Exposure from Leachate - Spillage

20 mSv/yr Worker (Ionising Radiation Radiations)

8.7 Annex B

Does not restrict landfill capacity because the landfill is smaller than the most restrictive case.


July 2009 83


Assessment Results

1 mSv/yr Worker (Operational Criterion) 0.02 mSv/yr Public (GRA)- Lower Bound 0.3 mSv/yr Public (GRA) – Upper Bound

Exposure from Leachate - Aerosols


8.8 Annex B

This scenario is less restrictive that the case at 8.6 which will result in more restrictive leachate discharge limits.

Exposure by Using Groundwater at Nearest Abstraction Point

0.02 mSv/yr Public (GRA) 8.9 Annex B


Exposure from Gas Generation from the Landfill – post closure

0.02 mSv/yr Public (GRA) 8.10 Annex B


Exposure to Wildlife from all sources

10 microgray/hr 8.11 Annex B

Not restrictive to landfill capacity.

External dose from emplaced wastes

0.02 mSv Public (GRA) 8.12 Annex B

Not restrictive to landfill capacity.

Exposure by Using Groundwater from a Borehole Constructed at the Boundary of the Landfill

3 mSv Public (GRA and HPA)

8.13 Annex B


Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill

3 mSv Public or Worker (GRA and HPA)

8.14 Annex B


8.15.1 A number of sensitivity studies have been undertaken (Annex B) to examine

uncertainties associated with the assessment results.


July 2009 84

8.16 Landfill Radiological Capacity 8.16.1 The actual radiological capacity depends on the proportions of the different

isotopes in the mixture of the entire waste that is disposed.

8.16.2 Section 6 of Annex B gives the radiological capacity for a typical and well-defined waste stream from the Harwell site, for illustration, on the assumption this was the only waste stream sent to the site and presents nuclide specific capacities for the various scenarios.

8.16.3 As discussed above the scenarios which could constrain landfill capacity are:

- Exposure from Gas Generation from the Landfill

- Exposure by Using Groundwater at Nearest Abstraction Point

- Exposure by Using Groundwater from a Borehole Constructed at the Boundary of the Landfill

- Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill

8.16.4 The table below identifies the specific doses for each of these cases as derived from the results in Annex B. The table identifies the most restrictive specific dose in each case (shown in yellow highlight) and gives the radiological capacity for that case for that nuclide only. Those shown in red outlined border are the nuclides which could have capacities that are smaller than the remaining physical capacity of the landfill (~1 million tonnes) and hence may be restrictive.


July 2009 85

Radionuclide

Specific Dose (microSv/yr per MBq) Groundwater Pathway, Borehole, 1500m Irrigation Case (Table 5.1 Annex B)


Specific Dose (microSv/yr per MBq) Groundwater Pathway, Site Boundary Drinking Case, (Table 5.1 Annex B)


Specific Dose (microSv/yr per MBq) Intrusion (Table 5.3, Annex B) "Intruder 60 year Case"


Specific Dose (microSv/yr per MBq) Intrusion (Table 5.3, Annex B) "Resident 60 year Case"


Specific Dose (microSv/yr per MBq) Gas Generation from the Landfill Combined Pre-Closure and Post Closure Public


H-3 3.66E-30 5.46E+30 6.89E-23 4.35E+25 4.52E-12 6.64E+14 1.28E-10 2.34E+13 1.13E-07 1.77E+08

C-14 2.06E-09 9.71E+09 1.39E-07 2.16E+10 7.05E-09 4.26E+11 8.68E-09 3.46E+11

Cl-36 6.52E-08 3.07E+08 1.69E-06 1.78E+09 3.02E-08 9.93E+10 4.28E-07 7.01E+09

Fe-55 1.04E-43 1.92E+44 4.45E-36 6.74E+38 4.85E-17 6.19E+19 8.13E-18 3.69E+20

Co-60 1.21E-39 1.65E+40 3.79E-32 7.92E+34 1.27E-08 2.36E+11 4.19E-11 7.16E+13

Ni-63 7.94E-21 2.52E+21 3.47E-15 8.65E+17 8.68E-10 3.46E+12 2.60E-10 1.15E+13

Sr-90 3.04E-24 6.58E+24 2.19E-17 1.37E+20 9.59E-08 3.13E+10 3.31E-07 9.06E+09

Nb-94 3.76E-10 5.32E+10 9.07E-09 3.31E+11 7.52E-05 3.99E+07 2.44E-07 1.23E+10

Tc-99 2.12E-09 9.43E+09 1.52E-07 1.97E+10 1.09E-08 2.75E+11 3.68E-07 8.15E+09

Ru-106 3.44E-46 5.81E+46 1.67E-38 1.80E+41 1.39E-26 2.16E+29 7.44E-29 4.03E+31

Ag-108m 1.68E-17 1.19E+18 1.51E-12 1.99E+15 5.13E-05 5.85E+07 1.68E-07 1.79E+10

Sb-125 4.81E-42 4.16E+42 2.03E-34 1.48E+37 5.68E-13 5.28E+15 1.89E-15 1.59E+18

Sn-126 1.20E-08 1.67E+09 4.87E-07 6.16E+09 1.34E-04 2.24E+07 4.90E-07 6.12E+09

I-129 1.38E-05 1.45E+06 3.02E-04 9.93E+06 1.06E-06 2.83E+09 1.36E-06 2.21E+09

Ba-133 1.44E-31 1.39E+32 3.25E-24 9.23E+26 1.65E-07 1.82E+10 5.63E-10 5.33E+12

Cs-134 1.82E-43 1.10E+44 8.12E-36 3.69E+38 6.83E-15 4.39E+17 2.76E-17 1.09E+20

Cs-137 1.28E-25 1.56E+26 9.32E-19 3.22E+21 5.60E-06 5.36E+08 2.61E-08 1.15E+11

Pm-147 3.41E-44 5.87E+44 1.47E-36 2.04E+39 3.54E-17 8.47E+19 4.55E-19 6.59E+21


July 2009 86

Radionuclide











Eu-152 2.77E-32 7.22E+32 4.84E-25 6.20E+27 1.42E-06 2.11E+09 4.63E-09 6.48E+11

Eu-154 3.28E-35 6.10E+35 7.98E-28 3.76E+30 2.41E-07 1.24E+10 7.87E-10 3.81E+12

Eu-155 3.63E-41 5.51E+41 1.29E-33 2.33E+36 8.04E-11 3.73E+13 2.74E-13 1.09E+16

Pb-210 1.25E-25 1.60E+26 1.43E-18 2.10E+21 2.19E-06 1.37E+09 1.28E-07 2.34E+10

Ra-226 1.56E-08 1.28E+09 1.36E-06 2.21E+09 1.08E-04 2.78E+07 5.40E-06 5.56E+08 2.85E-06 7.02E+06

Ac-227 5.66E-26 3.53E+26 6.70E-19 4.48E+21 2.14E-05 1.40E+08 1.81E-08 1.66E+11

Th-229 1.44E-08 1.39E+09 1.29E-06 2.33E+09 9.48E-05 3.16E+07 8.84E-08 3.39E+10

Th-230 8.24E-09 2.43E+09 7.35E-07 4.08E+09 3.49E-05 8.60E+07 1.56E-07 1.92E+10 7.30E-08 2.74E+08

Th-232 4.04E-08 4.95E+08 3.59E-06 8.36E+08 1.87E-04 1.60E+07 4.75E-07 6.32E+09

Pa-231 3.60E-08 5.56E+08 3.08E-06 9.74E+08 6.82E-05 4.40E+07 1.66E-06 1.81E+09

U-232 5.07E-20 3.94E+20 6.19E-14 4.85E+16 8.99E-06 3.34E+08 1.57E-08 1.91E+11

U-233 4.62E-09 4.33E+09 4.13E-07 7.26E+09 3.89E-06 7.71E+08 4.32E-09 6.94E+11

U-234 4.35E-09 4.60E+09 3.89E-07 7.71E+09 3.29E-06 9.12E+08 3.71E-09 8.09E+11 2.02E-11 9.90E+11

U-235 4.35E-09 4.60E+09 3.87E-07 7.75E+09 8.96E-06 3.35E+08 2.47E-08 1.21E+11

U-236 4.22E-09 4.74E+09 3.78E-07 7.94E+09 1.38E-06 2.17E+09 3.14E-09 9.55E+11

U-238 4.35E-09 4.60E+09 3.89E-07 7.71E+09 3.88E-06 7.73E+08 6.87E-09 4.37E+11 2.57E-12 7.78E+12

Np-237 1.22E-06 1.64E+07 9.52E-05 3.15E+07 2.47E-05 1.21E+08 4.65E-08 6.45E+10

Pu-238 1.86E-12 1.08E+13 1.67E-10 1.80E+13 2.03E-05 1.48E+08 9.86E-09 3.04E+11 1.64E-15 1.22E+16

Pu-239 1.54E-08 1.30E+09 1.37E-06 2.19E+09 3.84E-05 7.81E+07 1.86E-08 1.61E+11


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Radionuclide











Pu-240 1.04E-08 1.92E+09 9.33E-07 3.22E+09 3.82E-05 7.85E+07 1.85E-08 1.62E+11

Pu-241 1.59E-10 1.26E+11 3.62E-08 8.29E+10 1.47E-05 2.04E+08 8.29E-09 3.62E+11

Pu-242 1.69E-08 1.18E+09 1.51E-06 1.99E+09 3.54E-05 8.47E+07 1.76E-08 1.70E+11 6.46E-21 3.10E+21

Am-241 5.37E-11 3.72E+11 1.08E-08 2.78E+11 2.79E-05 1.08E+08 1.57E-08 1.91E+11

Cm-243 1.82E-10 1.10E+11 1.63E-08 1.84E+11 3.03E-06 9.90E+08 4.49E-09 6.68E+11

Cm-244 1.29E-09 1.55E+10 1.15E-07 2.61E+10 1.57E-06 1.91E+09 9.45E-10 3.17E+12

Restrictive Cases for Radiological Capacity on an Individual Nuclide Basis


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8.16.5 The likely conclusion is that the landfill can receive unlimited amounts of these nuclides at up to 200 Bq/g without exceeding the radiological capacity before reaching the physical capacity. For those highlighted nuclides their effect in practice will be reduced by being parts of mixtures with other nuclides and at average concentrations less than 200 Bq/g. Several of the most restrictive nuclides are short half-life and will generally not feature in decommissioning wastes in significant concentrations.

8.16.6 Notwithstanding the overall conclusion that capacity is not particularly restricted

in this case, the proposal is that the capacity of the landfill is subject to a total capacity limit combined with a series of other conditions. The total capacity limit would apply from the date of issue until closure of the landfill or until the capacity is reached. The landfill would receive no more LLW under the permit once the capacity limit is reached. The capacity limit cannot be expressed as a single number because it depends on the mixture received up to any point in time, so the proposal is for a continuously revised capacity limit based on individual nuclides (including appropriate daughter chains). The total capacity limit would be established using an authorised spreadsheet model agreed with the regulator. The spreadsheet model would represent the most restrictive case from the risk assessment and would produce as an output the remaining capacity of the landfill on an individual nuclide basis given the exact wastes received to that point in time. Prior to accepting any further waste the model would be used by the landfill operator to determine that the consignment would not lead to a breach of the total capacity limit. This approach has a number of features:

The approach requires a comprehensive level of waste characterisation by the consignors, but this is considered practicable and is optimal for ensuring public health is not impacted by imprecise waste assay. This is also sustainable because future generations will receive comprehensive information on the disposed nuclides enabling them to make informed decisions.

The approach cannot be expressed as a simple number and hence may be less transparent to the public, but the approach is highly transparent and detailed to the regulator.

The approach is “modern” in the sense that it aligns the authorisation with the risk assessment. This is in line with proposals to use risk based approaches for RSA exemptions orders, waste definitions, clearance definitions and nuclear site delicensing conditions. Hence the criteria used to produce the waste, to categorise the waste and to dispose of the waste are based on a consistent risk based approach that can be expressed in common terms of risk and dose.

The approach is based on the total life cycle of the facility. This addresses a potential public concern that the authorised capacity may “creep” upwards at the point of annual reviews. Authorisation creep of this type was identified as a concern from the pre-application stakeholder workshops.


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The approach drives the correct behaviour down to the consignor in respect of the waste hierarchy because the approach enables remaining nuclide capacity (hence price) to be directly related to the environmental risk of that nuclide.

The approach addresses the optimisation principle because the remaining capacity of the landfill is continuously optimised in a manner that ensures the overall risk guidance levels are not exceeded and the model will enable the operator and regulator to make informed choices.

The capacity of the landfill is administered in a manner that ensures dose limits and constraints will not be exceeded with a significant safety factor to account for uncertainties.

The approach maintains flexibility to account for the uncertain overall inventory of decommissioning wastes to be produced as required by regulatory guidance (ref 19).

8.16.7 For the purposes of authorisation this approach can be described as maintaining a condition in which the remaining radiological capacity for a particular nuclide or wastestream, which is proposed for receipt, is greater than zero, taking into account the wastestream received to that point in time. Where:

Radiological capacity is the amount of radioactive material that can be consigned to a site without any of the potentially exposed groups considered receiving a dose above a specified criterion, for the specific scenario. For a single radionuclide, the radiological capacity (in Bq) can be easily calculated by dividing the dose criterion (expressed in Sv) by the maximum specific dose for that radionuclide (expressed in Sv/Bq). For mixtures of individual radionuclides, the capacity can be simply apportioned (e.g., half of the overall capacity to each of two radionuclides). In the case of waste streams, however, in which the proportions of different radionuclides are fixed, the calculation of capacity must consider both the specific dose and the activity ratios.

The radiological capacity for radionuclide Rni in a waste stream (RCi) is given by:

DCfSD

fRC

ii

ii

where:

fi is the fraction of the overall activity arising from Rni (such that fi=1) SDi is the specific dose from Rni DC is the dose constraint Furthermore: Total activity limit for each radionuclide (Bq) = Dose Limit (Sv/y).Waste Activity (Bq)

Dose Estimate (Sv/y)


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With the additional constraint that the total dose from all of the radionuclides must not exceed the relevant dose limit:

Σi Qi / Qi,l <= 1

Where Qi (Bq) is the actual activity of radionuclide i to be disposed and Qi,l (Bq) is the activity limit for radionuclide i if it were the only radionuclide to be disposed of.

8.16.8 Not every isotope has been assessed in the model although a full range of

behaviours is encompassed by those which have been modelled. If another isotope type is required to be disposed, the model will be re-examined prior to acceptance of the waste.


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9.0 Radioactive Waste Disposal Proposed Authorisation Conditions and Waste Acceptance Criteria

9.0.1 The following are proposed authorisation conditions and waste acceptance

criteria subject to development during the authorisation process:

9.1 Potential Conditions Arising from the Standard RSA Authorisation Template

The use of Best Practicable Means for operation, management and maintenance.

Maintaining equipment and systems provided for the waste disposal process in good repair.

The requirement for a management system, organisational structure and sufficient resources to achieve compliance.

The requirement for training of staff in respect of the conditions of the authorisation, the operating techniques and emergency action plans.

The requirement for written operating arrangements.

The requirement for audit and review of arrangements.

The requirement for sampling, testing, calculations and analysis to determine compliance.

The requirement to keep records.

The requirement to inform the EA of certain matters of compliance.

9.2 Potential Conditions Arising from the Existing Landfill Permit

and the Landfill Regulations 2002

Inclusion of the underpinning limits established by existing risk assessments for the existing Landfill Permit for hazardous waste disposal, where applicable. Including:

Compliance with appropriate waste acceptance criteria for hazardous waste disposal, including compliance with the prohibited substance list, compliance with the waste acceptance limits for hazardous waste disposal established by underpinning risk assessments and for the appropriate designated leach testing of waste packages or self-similar waste streams.

Prohibited wastes include: liquids, explosive, corrosive, oxidising, flammable or highly flammable wastes, infectious clinical wastes, unknown chemical substances.


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Limit values are prescribed for certain leach parameters and total organic carbon. The limit values vary between granular or monolithic waste forms (LLW could be either type). The limit values appropriate for the LLW would be established through reference to the underpinning risk assessments for the landfill design.

Special arrangements apply to handling of asbestos bearing wastes.

Compliance with site specific authorisation conditions.

Compliance with the total tonnage limits placed on the landfill by the existing permit (It is important in order to maintain the integrity of the existing environmental impact assessment that the total physical capacity of the landfill is unchanged).

Prohibition of deliberate dilution or mixing to achieve waste acceptance criteria.

9.3 Conditions Arising from the Site Specific Risk Assessment and

Industry Practice

LLW will not be loose handled or tipped at the site.

LLW will be transported to the site in radioactive materials compliant sealed packages and in packages suitable for handling at the landfill site.

Waste will undergo an agreed pre-acceptance, pre-notification, receipt and disposal process in accordance with operating arrangements.

The paper work for each consignment and the waste load will be inspected to confirm that they are consistent with the waste booked into the site.

Wastes will be placed on the same day as receipt, or will be suitably quarantined where this is not practicable.

Wastes will be covered by at least 300m thickness of suitable cover after each emplacement campaign or at the end of the working day such that there is no exposed face. Sufficient cover will be used to ensure the doserate at 1 metre above the waste is less than 2 microSv/hr.

The waste will not be trafficked or compacted without a covering protective layer of suitable cover adequate to protect the waste from exposure to the surface.

A radiation protection plan and scheme of environmental monitoring will be operated in accordance with agreed operating arrangements.

The maximum concentration of radioactivity in any package will be 200 Bq/g averaged over the package and in any case not averaged over more than 4 tonnes.

The minimum concentration of radioactivity averaged in any package shall be such that the waste is defined as low level or very low level radioactive waste.


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Exempt and excluded wastes are not a relevant waste to the permit. It is proposed that if wastes of less than a relevant exemption or exclusion order are mixed in with the LLW as an inevitable result of their production then these would also be treated as LLW. Should the RSA exemption orders which define the boundary of Exempt and LLW wastes be revised, then the authorisation would automatically incorporate such changes.

The waste will be a solid waste as defined under the Landfill Regulations. Liquid wastes and slurries etc. are prohibited.

The radiological capacity will be monitored and complied with.

The waste will consist of waste material that is deemed to be contaminated and not be primary contaminant material (such material would be normally recoverable and hence not be a waste).

Notwithstanding the requirements of the existing permit under the Landfill Regulations which concern acceptability of chemical hazards in respect of hazardous waste - the waste will not be capable of generating toxic or explosive gases, vapours or fumes that would be harmful to persons involved in the waste process.

Notwithstanding the requirements of the existing permit under the Landfill Regulations which concern acceptability of chemical hazards in respect of hazardous waste – the waste will not contain pressurised gas receptacles as defined within the Carriage of Dangerous Goods…Regulations 2004 (or as amended).

LLW containing putrescible materials (materials liable to be readily decomposed by micro-organisms, excluding wood and paper) will be excluded in so far as is reasonably practicable.

Conditions for acceptance will dictate that the consignor ensures that external non-fixed contamination levels on waste packages will be as low as reasonably practicable throughout the process and in any case not more than 4 Bq/cm2 beta/gamma and 0.4 Bq/cm2 alpha averaged over an area of 300cm2 (as derived from normal industry practice).

External dose rates throughout the process will be as low as reasonably practicable, shall be in accordance with the transport regulations and shall not exceed 0.01 mSv/hr (10 microSv/hr) at 1m from the waste package.

LLW with hazardous properties that would mean it would be a hazardous waste if it were not radioactive, will comply with all the relevant conditions of the RSA authorisation in respect of non-radiological hazards.

The capacity of the landfill is subject to a total capacity limit combined with a series of other conditions. The total capacity limit would apply from the date of issue until closure of the landfill or until the capacity is reached. The landfill would receive no more LLW under the permit once the capacity limit is reached. The capacity limit cannot be expressed as a single number because it depends on the mixture received up to any point in time, so the proposal is for a continuously revised capacity limit based on individual


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nuclides (including appropriate daughter chains). The total capacity limit would be established using an authorised spreadsheet model agreed with the regulator. The spreadsheet model would represent the most restrictive case from the risk assessment and would produce as an output the remaining capacity of the landfill on an individual nuclide basis given the exact wastes received to that point in time. Prior to accepting any further waste the model would be used by the landfill operator to determine that the consignment would not lead to a breach of the total capacity limit. An example spreadsheet to illustrate the model is included in Annex G. It is considered that a condition as proposed below will provide robust and enforceable means of regulating the site capacity:

Radiological capacity is the amount of radioactive material that can

be consigned to the site without any of the potentially exposed groups considered receiving a dose above a specified criterion, for the specific scenario.


DCfSD

fRC

ii

ii

where:

fi is the fraction of the overall activity arising from Rni

(such that fi=1) SDi is the specific dose from Rni DC is the dose constraint

Furthermore:

Total activity limit for each radionuclide (Bq) = Dose Limit (Sv/y).Waste Activity (Bq)

Dose Estimate (Sv/y)

With the additional constraint that the total dose from all of the radionuclides must not exceed the relevant dose limit:

Σi Qi / Qi,l <= 1

Where Qi (Bq) is the actual activity of radionuclide i to be disposed

and Qi,l (Bq) is the activity limit for radionuclide i if it were the only radionuclide to be disposed of.

Values for the above parameters are obtained from the suitability assessment (Annex B).


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Any isotopes not modelled in the risk assessment will be modelled prior to acceptance and incorporated into the approved spreadsheet model for radiological capacity.

The conditions for the aftercare period and revocation of the authorisation, including the provisions for closure of the authorisation at the time the disposals cease and any provisions for monitoring during the aftercare period.


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10.0 BPEO Assessment for Disposal of LLW from Nuclear Sites

10.1 BPEO (Best Practicable Environmental Option)

10.1.1 The Royal Commission on Environmental Pollution defined the Best Practicable Environmental Option (BPEO) as:

“The outcome of a systematic and consultative decision-making procedure which emphasises the protection and conservation of the environment across land, air and water.

The BPEO procedure establishes, for a given set of objectives, the option that provides the most benefits or the least damage to the environment as a whole, at acceptable cost, in the long term as well as the short term.”

10.1.2 The Environment Agency requires the use of the BPEO methodology by nuclear

industry sites in order to underpin their waste management choices and strategies.

10.1.3 A BPEO study involves a rational consideration of all the options against a series

of criteria. Importantly it involves extensive consultation. 10.1.4 Any nuclear industry site that wished to send waste to the East Northants

Resource Management Facility under an authorisation would be required to have an underpinning BPEO study that justified this approach before they would be granted a transfer authorisation under the RSA 1993 by the EA.

10.1.5 The consigning site would in any case have to apply the waste management

hierarchy of avoid – minimise – recycle – reuse to any waste stream prior to consideration of disposal and would have to demonstrate the use of best practicable means in respect of waste generating activities. So, for example, a nuclear industry site is required to extensively sort wastes prior to dispatch in order to avoid unnecessary disposals.

10.1.6 The proposed approach is that the BPEO methodology would not be applied to

this authorisation application, but would be applied to each individual transfer authorisation application from nuclear industry sites.

10.1.7 Non-nuclear industry sites are not required to prepare BPEO assessments for

their waste streams, although these would normally be small volumes in comparison to the potential higher volumes from the nuclear industry.


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11.0 BPM and ALARA Assessment for the Proposed Radioactive Waste Disposal

11.1 ALARA 11.1.1 The “As Low as Reasonably Achievable” (ALARA) principle is concerned with

optimising radiation doses to humans. 11.1.2 “In relation to any particular source within a practice, the magnitude of individual

doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable, economic and social factors being taken into account.”

11.1.3 Conservative radiological assessments for workers and the public in the

operational and post-closure stages are presented in this report and demonstrate that it is likely that dose constraints, dose limits, design risk targets and design dose targets will be achieved. The design targets are set at levels beyond which further measures should only be considered necessary if they do not involve disproportionate costs.

11.1.4 Operational optimisation measures are described in this report and would be

developed as part of the radiation protection plan for the site. Feedback from workplace and environmental monitoring would be used to implement further optimisation measures if required in order to achieve actual exposures which are a fraction of the constraints and limits.

11.2 BPM 11.2.1 The Best Practicable Means (BPM) principle is essentially a consideration of

whether an adequate argument has been made that further measures to reduce risk are not needed because the measures cannot be implemented at a reasonable cost given economic and social factors.

11.2.2 Having carried out a BPEO study to consider what the right option to pursue is,

BPM is concerned with executing that option in the right way. 11.2.3 Whereas ALARA applies to dose optimisation, BPM applies to optimise

radioactive waste management. 11.2.4 Within a particular waste option, the BPM is that level of management and

engineering control that minimises, as far as practicable, the release of radioactivity to the environment whilst taking account of a wider range of factors, including cost-effectiveness, technological status, operational safety and social/environment factors.

11.2.5 To some extent the BPM concept overlaps with the BAT (Best Available

Techniques) concept that underpins the provision of new landfill designs under


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the Environmental Permitting Regulations (formerly the Landfill Regulations) and which applies to non-radioactive pollutants. Indeed, there is a view that that BAT and BPM are synonymous. Many BAT features of new landfill sites for hazardous waste such as the East Northants Resource Management Facility are effectively prescribed (for example, the permeability performance of barrier layers), whereas BPM features are not so specifically prescribed.

11.2.6 The BPEO study undertaken for the example Harwell waste stream is likely to be

typical for all decommissioning waste arising. That study indicates that shallow disposal in an engineered facility is likely to be the BPEO for most low level decommissioning wastes of the type proposed for ENRMF. Hence, BPM for such an option focuses on the design of the facility, whether it meets modern standards and whether any further straightforward improvements are feasible.

11.2.7 The current state of the wastes on the various nuclear sites, whilst adequately

controlled, is unarguably less satisfactory and less sustainable than final disposal. If the waste is not disposed to engineered facilities it will remain in above ground stores or in contaminated land areas and will present a higher risk to future generations. The proposed option represents a net reduction in risk from the current situation.

11.2.8 It is submitted that use of a modern standard hazardous waste landfill that has

been designed and implemented using BAT under recent legislative guidance represents BPM for the disposal of LLW of the type proposed for ENRMF. The reasoning is that the LLW has the same chemical properties whilst being no more mobile and are generally less reactive than the hazardous wastes for which the landfill was designed and that the landfill was designed in such a way as to prevent harm to humans and the environment. If a new specialist landfill were designed for the LLW of the type proposed for ENRMF it is unlikely to use engineering features and standards beyond those currently used to define BAT for modern hazardous waste landfills.

11.2.9 The risk assessments in this application support the case that the existing landfill

design will prevent harm arising from the LLW to an appropriate risk standard. 11.2.10Further limitations have been proposed on the disposal that are additional to the

BAT features of the existing landfill and these are described throughout this application and in particular in section 5, including :

Wastes will only be accepted for disposal if the source site (in the case of a nuclear industry site) demonstrates that the option is BPEO and that BPM has been used to apply the waste hierarchy and to characterise the waste.

The radiological capacity of the landfill has been back calculated to give a design risk target under the most restrictive future scenarios of <1 in a million/year for natural processes and 3 mSv/year for inadvertent intrusion, using conservative critical group modelling.


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The maximum concentration of radioactivity has been limited through proposed waste acceptance criteria to limit the effects of routine and accidental exposures from transport and emplacement operations such that dose constraints are achieved and improved upon.

Additional waste acceptance criteria have been proposed to further limit exposure. For example, a constraint on the external dose on the transport package has been proposed which is more constraining than required by transport regulations.

Operational arrangements have been proposed to further reduce exposure, including for example, no loose handling of materials, the use of suitable cover materials, the use of segregation arrangements, the use of contamination clearance and control arrangements, the use of personnel, workplace and environmental monitoring and the use of emergency arrangements.

11.2.11The likely result of these additional measures will be that the risk presented by

the waste disposal will be less than the design risk target over the long term and that occupational dose to workers will be well within the dose constraints.

11.2.12Can further measures be implemented?

The two principal possible further restrictions are to further limit the radiological capacity or to further limit the radioactivity concentration.

The long term risks are driven by the total activity disposed. Further

reductions in the capacity are not reasonably practicable because the capacity has been designed to a basic risk target that meets current guidance and in practice the further optimisation measures described above will reduce the risk still further. The capacities that result are of a size to be useful and economic for the decommissioning industry and reductions would make the waste route considerably less able to meet regional demand.

The short term risks are driven by the concentration of activity and

resulting doserate in any one package of waste for any given isotope type. The concentrations and doserates have been optimised through application of restrictive waste acceptance criteria and further restrictions are not required in order to achieve the occupational dose constraints for workers (based on those workers using the constraints applicable to the public). In practice the ALARA and BPM arrangements described above will further reduce exposures. The proposed radioactivity concentrations and doserate criteria are of practical use for the decommissioning industry and further reductions would severely limit the applicability of the waste route to solve the strategic drivers described above.


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12.0 Landfill Engineering and BAT Features of the Existing Landfill

12.0.1 The existing hazardous waste landfill at the East Northants Resource

Management Facility is authorised under the Landfill Regulations and the Pollution Prevention and Control Regulations.

12.0.2 The landfill is designed in accordance with these regulations and utilises Best

Available Techniques (BAT). These BAT design features and arrangements would also be utilised by the LLW and contribute to the BPM case above. Recently created hazardous waste landfill sites, such as the East Northants Resource Management Facility, have the highest level of BAT features (compared to non-hazardous or inert landfill sites).

12.0.3 The details of the BAT design features of the existing landfill are contained in ref

15 which gives hydrogeological, stability, landfill gas, environmental impact and nuisance risk assessments. The arrangements for construction design, waste acceptance, groundwater protection, landfill gas management, leachate management, landfill stabilisation, pollution prevention, nuisance prevention and quality assurance, construction quality assurance, maintenance, landfill capping, site restoration, operations, waste handling/placement, security, use of raw materials, secondary wastes, accident arrangements, monitoring, closure, aftercare and surrender are described in existing documentation for the landfill site.

12.0.4 These BAT features represent a solid foundation for the management of the LLW

and have been taken into account in the risk assessment for LLW disposal to the extent detailed in this document. The features are not described in detail in this document. An outline of the key landfill engineering features follows:

A full containment landfill engineering system designed to meet the requirements of the Landfill Regulations 2002 and 2004 (as amended). This requires a basal lining system with, or equivalent to having, a permeability of 1 x 10-9 m/s or lower and a thickness of no less than 5m or equivalent. For the basal liner the landfill incorporates a 1.5m thick layer of reworked clay with a maximum permeability of 3x10-10 m/s and a 2mm high density polyethylene synthetic liner. The sidewalls are formed from the in-situ clay materials with the liner placed over these.

The surface capping system comprises from the waste surface upwards:

300mm regulating layer

geosynthetic clay liner

1mm welded geomembrane

protector geotextile or drainage geocomposite

1m of restoration soils


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A leachate collection system

A gas collection system

Ancillary systems such as vehicle cleaning equipment

A surface water, groundwater and environmental monitoring system

Restoration of the site to grassland including wildflower meadow and agricultural grassland

Operational arrangements for site construction, operation, closure, restoration and aftercare.

12.0.5 The proposal for LLW disposal does not change the existing arrangements and

augments the arrangements for the LLW component.


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13.0 Waste Hierarchy and Waste Minimisation at Source 13.0.1 Producing sites will be required to demonstrate that the waste hierarchy has

been applied to the waste prior to acceptance by the landfill.

Avoid – Wastes are not generated if this is feasible, for example, maintaining separation of clean and radioactive materials in a building rather than deliberately mixing the wastes to produce a larger volume of lower concentration material.

Segregate – For example through application of BPM methods to characterise waste into exempt and LLW waste streams rather than mixing.

Prevent Spread – For example, preventing the spread of contamination to clean materials.

Recycle/Reuse – For example, reuse of contaminated or activated concrete as a construction material within the nuclear industry.

Clearance, Exemption and Exclusion – For example, use of the good practice to sort materials into classifications that can be defined as non-radioactive waste for the purposes of disposal.

Volume Reduction – For example, compacting wastes which are compactable to reduce disposal volume.

Disposal


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14.0 Summary of the Existing Environmental Statement for the Site and Impacts of the Proposal

14.0.1 The environmental impact statement for the existing landfill (ref 15) describes the

impacts from hazardous waste disposal operations. 14.0.2 It is submitted that the addition of a LLW stream to the inventory of waste

acceptable at the site makes no significant change to the existing environmental impact assessment or current/future use of the site.

14.0.3 The following table is a summarised version of table 15.1 from ref 15 which

summarises the existing environmental impact assessment and to which has been added comments on the changes introduced by the LLW stream.

Feature and Interest Existing Impact and

Mitigation Changes Resulting from LLW Disposal Authorisation

GROUNDWATER Aquifer – flow characteristics Change in recharge and flow

considered insignificant. None – the design of the existing landfill is unchanged

Aquifer – groundwater quality Residual impact from leachate considered to be insignificant due to engineered mitigation features

Insignificant additional risk as demonstrated by the risk assessment

SURFACE WATER Steams - flow Potential change in flow due to

new landform, considered to be insignificant after impact

No additional change in landform.

Steams - quality Potential impact from leachate and runoff, considered to be insignificant after impact.

Insignificant additional risk as demonstrated by the risk assessment

LANDSCAPE Westhay Cottages – visual receptor

Design and screening to mitigate impact.


Landscape Character Area – landscape receptor

Design and screening to mitigate impact.


TRAFFIC Traffic – Residents Change in traffic insignificant No additional change in traffic

quantity. Traffic - Motorists Roads have adequate

capacity No additional change in traffic quantity.

Traffic - Safety Motorists Mud on road mitigated through decontamination measures

Additional monitoring to check for contamination spread

NOISE Noise – Local community Some noise mitigation

measures applied to mitigate noise from landfilling operations

No additional noise


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ECOLOGY and NATURE CONSERVATION Woods, hedgerows, scrub, grassland, standing water, badgers, birds, amphibians, reptiles, bats, rare plant species.

Some preventative measures, surveys and habitat creation schemes used to mitigate potential impacts.

No additional land usage. The risk assessment shows that the impact will be insignificant.

AIR QUALITY Local property and global climate

Mitigation measures to control gas generation and odour/dust.

Risk from additional gas discharges assessed as insignificant. No handing of loose wastes. Emergency measures to address spillages and dropped loads.

HEALTH Health – people Risk of exposure to dusts,

aerosols, gas, contaminants, leachate, vermin, and litter mitigated through engineering design features and operational arrangements.

Health impact from LLW assessed to be insignificant during the operational and post-closure phases.

ARCHAEOLOGY Archaeology No known impacts. No additional impacts. RADIOACTIVITY (added to this table for this report) Health – people in the long and short term – workers and the public

Risk of exposure to radiation. The risk to workers has been reduced to low levels which are better than occupational dose targets and constraints through limits on waste acceptance and through operational arrangements. The risk to the public in the short and long term has been reduced to insignificant levels through limits on waste acceptance and through operational arrangements.


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15.0 Outline of Management and Operating Arrangements

15.0.1 Augean formed in 2004 is the UK’s market leader specialising in the management of hazardous waste. The Company provides a complete solution for the management of hazardous wastes and works in partnership with producers to provide long-term answers to the treatment and disposal of our more difficult to manage wastes. Augean operates proactively to ensure that regulatory standards are met and often exceeded. Best practice is considered normal practice. The Company currently owns more than 8 million cubic metres of void space, five treatment centres and employ over 150 people across 10 sites.

15.0.2 The locations of Augean facilities are shown below:

15.1 Augean Corporate Social Responsibility 15.1.1 Augean is committed to Corporate Social Responsibility as demonstrated through

the publication annually of a Corporate Social Responsibility report which measures our performance in respect of business, health and safety, our employees, our neighbours and the environment. An essential element of our approach to our business is our core business values.


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15.2 Augean’s Core Business Values:

Transparency we are open and transparent in all that we do

Integrity we are trustworthy and honest in all that we say and do and take responsibility for our own actions

Social and community responsibility

we recognise that our actions have a material impact on the communities in which we operate and take that responsibility extremely seriously

Environmental responsibility

we respect the environment and invest time and resource in protecting it

Technical excellence we employ skilled staff and use up-to-date techniques and equipment

Professionalism we are reliable and consistent and deliver excellent service

Respect we are friendly and courteous to colleagues, clients and suppliers

Passion we are proud of our company and dedicated to its purpose. We are enthusiastic, enjoy challenges and are eager for success

15.3 Management systems 15.3.1 Operational performance is maintained through a certified Integrated

Management System (IMS) delivering protection of health and safety, both internally and externally, and the management, protection and improvement of the environment for nature and our local communities. The IMS is certified by the British Standards Institute to the following standards:

IS0 9001 Quality management system ISO 14001 Environmental management system OHSAS 18001 Health and safety management system PAS 99 Integrated management system

15.3.2 Central to the Integrated Management System is the IMS Policy statement.


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15.4 Augean’s IMS Policy: Augean is committed to conducting its business operations in a responsible manner and we recognise the need to continually improve our operations where practical to do so in order to reduce our effects on the environment, ensure the safety and welfare of our personnel and neighbours, and ensure client satisfaction through service excellence.

We seek to exceed legal obligations and be among the leading exponents of good practice and technological development within the waste management industry. At no time shall we provide services that fall short of the professional integrity and objectivity that we understand our clients and stakeholders will require and every effort shall be sustained to ensure the accuracy, probity and surety of the services that we provide.

To achieve this and remain competitive, we pursue a programme of continuous improvement in all aspects of our business. To assist in achieving this high level of regulatory compliance, client satisfaction and operational improvement, corporate objectives shall be set on an annual basis. Realisation of set objectives is continuously monitored, reviewed and communicated throughout the company. To ensure a high standard of awareness within the company we provide our employees with continuous training to improve their skills and competencies. To maintain external awareness and good perception that the company actively liaises with regulatory bodies, environmental organisations, stakeholders, the local community and all other interested parties. The company shall encourage our supply chain and contractors to improve business standards through continual assessment. It is the policy of the company that the documented Business Management System detailed in the Business Manual and supporting administrative procedures are the normal basis of working and will be applied to all relevant work. Paul Blackler, Chief Executive


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15.5 Augean Organisation

15.5.1 Site Managers are responsible for the quality, health and safety and

environmental performance of their sites. The Site Manager reports directly to the Management Board which is ultimately responsible for performance. The Site Manager at East Northants Resource Management Facility is a holder of a Certificate of Technical Competence for the management of a hazardous landfill. Technical support and expertise is provided by the Technical Team specifically the Technical Manager who deals with Authorisation issues and legislative compliance, the monitoring team that monitors the environmental impact of the site in all media and the site chemists who provide laboratory facilities and determine the suitability of waste for acceptance at the site. The Technical Team reports to the Group Technical Director who is a member of the Management Board and advises the Board on health and safety and environment issues.

15.5.2 Augean employs a range of highly qualified professionals with expertise in

environmental and health and safety legislation, environmental management, chemistry, ecology, planning, engineering and waste management. As

Management Board

Site Manager Site Manager

Site Manager

Group Technical Director

Site Supervisor

Technical support Technical Manager Monitoring Manager

Site Chemists and Laboratory

Operatives


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necessary expertise is outsourced from external consultants. The Company maintains a list of approved consultants who are selected on the basis of qualification and experience and whose place on the list is dependent on good service.

15.6 Operational control 15.6.1 Through the IMS the aspects and impacts of the business have been

established. Risk assessments have been conducted for all operational activities and where necessary to ensure adequate operational control procedures have been developed and implemented. The Table below lists the main operational procedures relevant to this application. Additional procedures specific to disposal of LLW would be developed as required.

Reference No

Title

BM01 Business Manual BMS02 Customer Care Policy BMS05 Group Environmental Aspects Register BMS05 Environmental Aspects - Kings Cliffe BMS07 Group Register of Environmental Regulations BMS14 COSHH Register and forms - Kings Cliffe BMS18 Eye and Eye Sight Test Policy BMS19 Group Health and Safety Regulatory Register CBP01 Document Control CBP03 Training CBP04 Communication CBP08 Regulatory Compliance CBP09 Assessment of Enviornmental Effects CBP10 Supplier Evaluation CBP12 Control of Contractors CBP13 PPC Emergency Preparedness and Response CBP15 PPC Handling environmental and safety

complaints

CBP16 PPC non-conformance identification, investigation and implementation of corrective and preventative actions

CBP17 Monitoring and reporting CBP18 Internal Auditing CBP19 Management Review CBP27 Plant maintenance CBP40 Data Back up CBP41 Management of change CBP43 Permit to work instruction CPR 01 Sampling of hazardous waste CPR 02 Emergency preparedness and response CPR 03 Collection of Windblown Litter Risk Assessment


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CPR 04 Off loading of palletised waste Risk Assessment CPR 05 Environmental Monitoring Risk Assessment CQP01 Customer Complaints CQP02 Customer Feedback CQP03 Telephone Contact CSP01 COSHH CSP02 Risk Assessment CSP03 Health Surveillance CSP04 Fork Lift Trucks CSP05 Welding and flame cutting CSP06 Electrical safety CSP07 Manual Handling CSP08 Noise Control CSP09 Personal Protective Equipment CSP10 First Aid CSP11 Fire Safety CSP12 Lifting Operations and lifting equipment CSP13 New and expectant mothers CSP14 Young persons CSP15 Transport Safety CSP16 Display Screen Workstations CSP17 Violence at Works CSP18 Visitors CSP19 Housekeeping CSP20 Statutory Inspections CSP21 Health and Safety information CSP23 Consultation with employees CSP24 Disabled Persons CSP25 Pressure systems CSP26 Lone Working CSP27 Accident Investigation CSP28 Stress at Work CSP29 Carriage of Samples CSS01 Tipping Artic Systems CSS02 Disposal of hazardous waste CSS04 Water bowser (filling, use etc) CSS05 Unloading palletised loads CSS08 Wheel wash maintenance CSS10 Use of strimmers CSS11 Refuelling of plant and equipment CSS15 Towing vehicles CSS19 Errection and dismantling of litter fencing CSS22 Tipping artics containing asbestos waste CSS23 Tipping artics containing non-hazardous gypsum

wastes

IG01 Sampling of Waste


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KC01 Site Rules - Drivers KC01 Site Rules - Visitors and Contractors KC06 Emergency Plan LF001 Offices Risk Assessment LF002 Operational Areas Risk Assessment LF003 Plant Risk Assessment LF004 Site Traffic Risk Assessment LF005 Wheel cleaning facilities Risk Assessment LF006 Tipping Artics Risk Assessment MHR1 Rock Salt Bags Risk Assessment MHR11 Waste inspection and sampling Risk Assessment MHR2 Handling of Drummed Oils and Greases Risk Assessment MHR3 Handling and Installation of Leachate Pipes Risk Assessment MHR4 Unloading Palletised Waste Risk Assessment MHR5 Handling of Deliveries Risk Assessment MHR6 Water Bowser Risk Assessment MHR7 Erection and Dismantling of Litter Fencing Risk Assessment MHR8 Collection of Windblown Litter Risk Assessment PPC LF 02 Acceptance of hazardous waste to landfill PPC LF 10 Non-conforming waste loads PPC LF 11 Quarantined Waste Loads PPC LF 12 Cover control PPC LF 13 Pest control PPC LF 14 Litter control PPC LF 15 Noise control PPC LF 16 Odour control PPC LF 17 Control of dust and particulates PPC LF 18 Classification, assessment and accpetance of inert

wastes

PPC LF 19 Security Procedures SR1 Work on or above water Risk Assessment SR2 Tipping Artics Risk Assessment SR3 Steam Cleaning Plant and Equipment Risk Assessment SR4 Disposal of Asbestos Risk Assessment SR6 Leptospirosis Risk Assessment SR7 Tetanus Risk Assessment Disability Risk Assessment Risk Assessment Fire Risk Assessment Risk Assessment


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15.7 Arrangements Specific to LLW Disposal Operations 15.7.1 The following arrangements will be incorporated into the management system

specific to LLW disposal operations:

- A radiation protection plan and risk assessment as required by the Ionising Radiations Regulations, prepared by the site Radiological Protection Advisor and Qualified Expert. ENRMF, IRRs 1999, Radiation Risk Assessment for LLW, HPA (ref 16)

- An amendment to the site emergency plan to include response arrangements to identified fault scenarios including:

Dropped load

Contamination discovery

Non-compliant load

Dose above threshold discovery

Potentially contaminated person or wound

- A procedure for the receipt of waste, assay, quarantine, waste emplacement, coverage, record keeping and general LLW disposal operations

- A procedure for routine and periodic health surveillance monitoring for contamination and exposure

- Procedures for environmental monitoring incorporated into the MAPs

- A procedure for the pre-acceptance of waste including the conditions for acceptance for LLW for use in contractual arrangements with consignors

- Amendments to existing roles and responsibilities to add the roles:

Radiation Protection Advisor (Qualified Expert),

Radiation Protection Supervisor(s),

Dangerous Goods Safety Advisor (Class 7) 15.7.2 Augean have retained the service of the Health Protection Agency as

Radiological Protection Supervisor and to act as Qualified Expert in respect of such matters as training, advice, emergency response, provision and calibration of instrumentation, provision of health physics services, environmental and workplace monitoring, analysis, interpretation of specialist information etc.


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16.0 Stakeholder Consultation 16.0.1 Working in close co-operation with potential consignors of LLW and the

Environment Agency, Augean plc has undertaken a public consultation programme in support of this application for authorisation in accordance with Government and Local Planning Authority best practice guidance on this aspect of development.

16.0.2 The specialist development and planning communications company, Jennings

Nicholson Associates, have assisted Augean in this task. 16.0.3 The purpose of the communications programme has been to inform and educate

those affected by this application, to reassure the local communities of the non-threatening nature of what is proposed and to ensure that all the key stakeholders have an opportunity to voice their comments and concerns so that the company can address them during the authorisation process.

16.0.4 Augean has operated in the area of its East Northants Resource Management

Facility (formerly Kings Cliffe Landfill) since 2004. In that time it has built up a good working relationship and enhanced its corporate reputation with the local communities and those elected to represent them, as well as the statutory and non-statutory consultees. It will build on this foundation to engage all the key target audiences during the consultation process associated with this application.

16.0.5 The programme has been set out in a Communications Plan, which established

clear communications objectives, set out a carefully-timed phased programme to reflect milestones in the determination process and identified a variety of proven and effective mechanisms to promote the scheme and it key messages.

16.0.6 The programme has included: meetings of the local liaison committee, a one day

public surgery, meetings with official bodies and regulators and meetings with county, district and parish councils. The results of the feedback from these events has been analysed and used to refine this authorisation application.

16.0.7 The LLW disposal process is subject to a planning approval under Town and

Country Planning regulations which will involve further stakeholder dialogue.


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17.0 The Application Forms 17.1 Waste Disposal 17.1.1 A copy of the application form for a disposal authorisation (under sect 13 of the

Radioactive Substances Act 1993 (RSA) is included in Annex F.


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18.0 Conclusion 18.0.1 A proposed set of outline arrangements, waste acceptance criteria and potential

authorisation conditions has been described for a process to dispose of solid LLW wastes to East Northants Resource Management Facility.

18.0.2 A consequence and risk assessment has been carried out for the public and

workers in the long and short term. The radiological capacity of the landfill site has been back calculated in order to meet defined risk and dose targets. In addition, operational arrangements and constraints have been proposed using best practicable means to further reduce risk and optimise exposures.

18.0.3 The proposal is that the capacity of the landfill is subject to a total capacity limit combined with a series of other conditions. The total capacity limit would apply from the date of issue until closure of the landfill or until the capacity is reached. The landfill would receive no more LLW under the permit once the capacity limit is reached. The capacity limit cannot be expressed as a single number because it depends on the mixture received up to any point in time, so the proposal is for a continuously revised capacity limit based on individual nuclides (including appropriate daughter chains). The total capacity limit would be established using an authorised spreadsheet model agreed with the regulator. The spreadsheet model would represent the most restrictive case from the risk assessment and would produce as an output the remaining capacity of the landfill on an individual nuclide basis given the exact wastes received to that point in time. Prior to accepting any further waste the model would be used by the landfill operator to determine that the consignment would not lead to a breach of the total capacity limit.

18.0.4 It is submitted that the proposal for disposal of LLW is justified.


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References

1 Process and Information Document for: Applications for New Authorisations;.....issued under the Radioactive Substances Acts 1993 to Nuclear Sites in England and Wales, EA, 16/12/05,

Version 1

2 Considerations for Radioactive Substances Regulation under the RSA 1993 at Nuclear Sites in England and Wales, 16/12/05, EA

3 Policy for the Long Term Management of Solid Low Level Radioactive Waste in the UK, March 2007, DEFRA

4 Radioactive Substances Act 1993

5 RWMAC, Advice to Ministers on Management of Low Activity Solid Radioactive Wastes, 2003

6 Radioactive Substances Act 1960, A guide to administration of the Act

7 Environment Act 1990

8 Hazardous Waste Regulations 2005.

9 The Pollution, Prevention and Control Regulations 2000

10 The Landfill Regulations 2002 (as amended 2004 and 2005)

11 Implications of European Directives for the Disposal of Radioactive Wastes, DEFRA, October 2005

12 Radiological Assessment of Disposal of Large Quantities of Very Low Level Waste in Landfill Site, Chen, Kowe, Mobbs and Jones, HPA-RPD and Atkins, HPA-RPD-020, March 2007

13 The Carriage of Dangerous Goods and Use of Transportable Pressure Equipment Regulations 2007, No 1573

14 Documents of the HPA: Radiation Protection Objectives for the Land-Based Disposal of Solid Radioactive Wastes, RCE-8, February 2009.

15 Augean South Ltd., East Northants Resource Management Facility, Environmental Statement, Bullen Consultants, June 2005

16 ENRMF, IRRs 1991, Radiation Risk Assessment for LLW, HPA March 2009.

17 SNIFFER, UKRSR05: BPM for the Management of Radioactive Waste, 2005

18 Near-surface Disposal Facilities on Land for Solid Radioactive Wastes, Guidance on Requirements for Authorisation, February 2009


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19 Environmental Permitting Guidance Radioactive Substances Regulation (RSR), Draft Guidance for Consultation, May 2009.

20 UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear Industry: UK Nuclear Industry LLW Strategy, Consultation Document, June 2009, Nuclear Decommissioning Authority.

21 Impact Cratering: A Geologic Process, H.J.Melosh

22 UKAEA Safety Assessment Handbook

23 Report into Wisbech Air Crash, 1979, Hansard

24 Aircraft Accident Report No 2/90 (EW/C1094), Report on the accident to B747-121, N739PA, Lockerbie, 1988.


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Figures


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Glossary In the context of this Glossary, the term ‘waste’ refers, in general, to radioactive waste unless otherwise specified.

absorbed dose. See dose, absorbed.

activation. The process of inducing radioactivity. Most commonly used to refer to the induction of radioactivity in moderators, coolants, and structural and shielding materials, caused by irradiation with neutrons.

activation product. A radionuclide produced by activation. Often used in distinction from fission products. For example, in decommissioning waste comprising structural materials from a nuclear facility, activation products might typically be found primarily within the matrix of the material, whereas fission products are more likely to be present in the form of contamination on surfaces.

activity. The quantity A for an amount of radionuclide in a given energy state at a given time. The SI unit of activity is the reciprocal second (s–1), termed the Becquerel (Bq). Formerly expressed in curie (Ci), which is still sometimes used.

activity, specific. Of a radionuclide, the activity per unit mass of that nuclide. Of a material, the activity per unit mass or volume of the material in which the radionuclides are essentially uniformly distributed.

ALARP & ALARA. As low as reasonably practicable. As low as reasonably achievable. ALARP & ALARA describe approaches to optimisation. The optimisation principle states “in relation to any particular source within a practice, the magnitude of individual doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable (ALARA), economic and social factors being taken into account…” ALARA is incorporated in UK law via RSA 1993 (BSS) Direction 2000. ALARA & ALARP focus on impacts to people.

alpha bearing waste. See waste, alpha bearing.

analysis. Often used interchangeably with assessment, especially in more specific terms such as safety analysis. In general, however, analysis suggests a more narrowly technical process than assessment, aimed at understanding the subject of the analysis rather than determining whether or not it is acceptable. Analysis is also often associated with the use of a specific technique. Hence, one or more forms of analysis may be used in assessment.

analysis, consequence. A safety analysis that estimates potential individual or collective radiation doses to humans on the basis of radionuclide releases and transport from a nuclear facility (e.g. a waste storage facility or disposal site) to the human environment as defined by hypothetical release and transport scenarios.

analysis, deterministic. A simulation of the behaviour of a system utilizing one set of parameters, events and features. See also analysis, probabilistic.

analysis, probabilistic. A simulation of the behaviour of a system defined by parameters, events and features whose values are represented by a statistical distribution. The analysis gives a corresponding distribution of results. See also analysis, deterministic.

analysis, risk. An analysis of possible events and their probabilities of occurrence together with their potential consequences.

analysis, safety. An evaluation of the potential hazards associated with the implementation of a proposed activity.


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analysis, sensitivity. A quantitative examination of how the behaviour of a simulated system (e.g. a computer model) varies with change, usually in the values of its parameters. Two common approaches used are: parameter variation, in which the variation of the results is investigated for changes in one or more input parameter values within a range around selected reference or mean values, and perturbation analysis, in which the variations of the results with respect to changes in all the input parameter values are obtained by applying differential, integral or probabilistic analysis.

analysis, uncertainty. An analysis of the amount of variation in the results of assessments or analyses due to incomplete knowledge about the current and future states of a system.

aquifer. A water bearing formation below the surface of the earth that can furnish an appreciable supply of water for a well or spring.

area, controlled. A defined area in which specific protection measures and safety provisions are or could be required for controlling normal exposures or preventing the spread of contamination during normal working conditions, and preventing or limiting the extent of potential exposures.

argillaceous. The term applied to all rocks and substances composed of clay or having a notable proportion of clay in their composition.

assessment. The process, and the result, of analysing systematically the hazards associated with sources and practices, and associated protection and safety measures, aimed at quantifying performance measures for comparison with criteria. Assessment should be distinguished from analysis. Assessment is aimed at providing information that forms the basis of a decision whether something is satisfactory or not. Various kinds of analysis may be used as tools in doing this. Hence an assessment may include a number of analyses.

assessment, consequence. An assessment of the radiological consequences (e.g. doses and activity concentrations) of normal operation and possible accidents associated with a proposed or authorized facility or part thereof. This differs from risk assessment in that probabilities are not included in the assessment.

assessment, environmental (impact). An evaluation of radiological and nonradiological impacts of a proposed activity, where the performance measure is overall environmental impact, including radiological and other global measures of impact on safety and environment.

assessment, performance. An assessment of the performance of a system or subsystem and its implications for protection and safety at a planned or an authorized facility. This differs from safety assessment in that it can be applied to parts of a facility, and does not necessarily require assessment of radiological impacts.

assessment, risk. An assessment of the radiological risks associated with normal operation and potential accidents involving a source or practice. This will normally include consequence assessment and associated probabilities.

assessment, safety. An analysis to evaluate the performance of an overall system and its impact, where the performance measure is radiological impact or some other global measure of impact on safety. See also assessment, performance.

attribute. In the context of multi attribute decision aiding, attributes are features that the options possess which can be used to distinguish between the options in terms of advantages and disadvantages. For example, when choosing between types of lawnmower attributes might be; price, colour, weight, power source, fineness of cut, safety etc.

audit. A documented activity performed to determine by investigation, examination and evaluation of objective evidence the adequacy of, and adherence to, established procedures, instructions, specifications, codes, standards, administrative or operational programmes and other applicable documents, and the effectiveness of implementation.


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authorization. The granting by a regulatory body or other governmental body of written permission for an operator to perform specified activities. Authorization could include, for example, licensing, certification and registration. See also licence.

backfill. The material used to refill excavated portions of a repository (drifts, disposal rooms or boreholes) during and after waste has been emplaced.

background (radiation). The dose, dose rate or an observed measure related to the dose or dose rate, attributable to all sources other than the one(s) specified.

barrier. A physical obstruction that prevents or delays the movement of radionuclides or other material between components in a system, for example a waste repository. In general, a barrier can be an engineered barrier which is constructed or a natural (or geological) barrier.

barrier, intrusion. The components of a repository designed to prevent inadvertent access to the waste by humans, animals and plants.

barriers, multiple. Two or more natural or engineered barriers used to isolate radioactive waste in, and prevent radionuclide migration from, a repository. See also barrier.

borehole. A cylindrical excavation, made by a drilling device. Boreholes are drilled during site investigation and testing and are also used for waste emplacement in repositories and monitoring.

BPEO. Best Practicable Environmental Option. The outcome of a systematic and consultative decision-making procedure which emphasises the protection and conservation of the environment across land, air and water. The BPEO procedure establishes, for a given set of objectives, the option that provides the most benefits or the least damage to the environment as a whole, at acceptable cost, in the long term as well as the short term.

Bq/g A Becquerel (abbreviated as Bq) is the International System (SI) unit for the activity of radioactive material. One Bq of radioactive material is that amount of material in which one atom is transformed or undergoes one disintegration every second. A Gram (abbreviated as g) is a unit of mass. A Becquerel per Gram (abbreviated Bq/g) is therefore a measure of the concentration of radioactivity in a material.

characterization, site. Detailed surface and subsurface investigations and activities at candidate disposal sites to obtain information to determine the suitability of the site for a repository and to evaluate the long term performance of a repository at the site.

characterization, waste. Determination of the physical, chemical and radiological properties of the waste to establish the need for further adjustment, treatment, conditioning, or its suitability for further handling, processing, storage or disposal.

clay. Minerals that are essentially hydrated aluminium silicates or occasionally hydrated magnesium silicates, with sodium, calcium, potassium and magnesium cations. Also denotes a natural material with plastic properties which is essentially a composition of fine to very fine clay particles. Clays differ greatly mineralogically and chemically and consequently in their physical properties. Because of their large surface areas, most of them have good sorption characteristics.

cleanup. Any measures that may be carried out to reduce the radiation exposure from existing contamination through actions applied to the contamination itself (the source) or to the exposure pathways to humans. In a radioactive waste management context, cleanup has essentially the same meaning as rehabilitation, remediation and restoration.

clearance. Removal of radioactive materials or radioactive objects within authorized practices from any further regulatory control by the regulatory body.

clearance level. See level, clearance.


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closure. Administrative and technical actions directed at a repository at the end of its operating lifetime — for example covering the disposed waste (for a near surface repository) or backfilling and/or sealing (for a geological repository and the passages leading to it) — and termination and completion of activities in any associated structures.

commissioning. The process during which systems and components of facilities and activities, having been constructed, are made operational and verified to be in accordance with design specifications and to have met the required performance criteria. Commissioning may include both non-radioactive and radioactive testing.

compaction. A treatment method where the bulk volume of a compressible material is reduced by application of external pressure — hence an increase in its density (mass per unit volume).

conditioning. Those operations that produce a waste package suitable for handling, transport, storage and/or disposal. Conditioning may include the conversion of the waste to a solid waste form, enclosure of the waste in containers, and, if necessary, providing an overpack. See also immobilization.

conductivity, hydraulic, K. Ratio of flow rate n to driving force dh/dl (the change of hydraulic head with distance) for viscous flow of a fluid in a porous medium. This is the so-called constant of proportionality K in Darcy’s law and depends on both the porous medium and the fluid properties. See also permeability.

container, waste. The vessel into which the waste form is placed for handling, transport, storage and/or eventual disposal; also the outer barrier protecting the waste from external intrusions. The waste container is a component of the waste package. See also barrier; cask; waste package.

containment. Methods or physical structures designed to prevent the dispersion of radioactive substances.

contamination. (1) Radioactive substances on surfaces, or within solids, liquids or gases (including the human body), where their presence is unintended or undesirable, (2) the presence of such substances in such places or (3) the process giving rise to their presence in such places.

contamination, fixed. Contamination other than non-fixed contamination.

contamination, non-fixed. Contamination that can be removed from a surface during any handling activities, including routine conditions of transport.

control, institutional. Control of a waste site by an authority or institution designated under the laws of a country. This control may be active (monitoring, surveillance and remedial work) or passive (land use control) and may be a factor in the design of a nuclear facility (e.g. a near surface repository).

control, regulatory. Any form of control applied to facilities or activities by a regulatory body for reasons related to protection or safety.

controlled area. See area, controlled.

cover. A layer of material or materials placed over the waste packages or physical structures in a near surface repository. The main purpose of covers is to prevent ingress of surface water into the repositories and to reduce the likelihood of intrusion.

criteria. Conditions on which a decision or judgement can be based. They may be qualitative or quantitative and should result from established principles and standards. See also requirement; specifications.

critical group. A group of members of the public which is reasonably homogeneous with respect to its exposure for a given radiation source and given exposure pathway and is typical of


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individuals receiving the highest effective dose or equivalent dose (as applicable) by the given exposure pathway from the given source.

critical pathway. The dominant environmental route by which members of the critical group are exposed to radiation. For example, the critical pathway for iodine discharged with gaseous effluents is from pasture to cows and then to milk. Consumption of the milk by individuals gives rise to exposure to radiation.

decommissioning. Administrative and technical actions taken to allow the removal of some or all of the regulatory controls from a facility. This does not apply to a repository or to certain nuclear facilities used for mining and milling of radioactive materials, for which closure is used.

decontamination. The complete or partial removal of contamination by a deliberate physical, chemical or biological process.

depleted uranium. See uranium, depleted.

design. The process and result of developing a concept, detailed plans, supporting calculations and specifications for a facility and its parts.

desorption. See sorption.

deterministic analysis. See analysis, deterministic.

diffusion. The movement of atoms or molecules from a region of higher concentration of the diffusing species to regions of lower concentration, due to a concentration gradient.

discharge. A planned and controlled release of (usually gaseous or liquid) radioactive material to the environment.

discharge, authorized. A discharge in accordance with an authorization. See limit, authorized.

discharges, radioactive. Radioactive substances arising from a source within a practice which are discharged to the environment, generally with the purpose of dilution and dispersion.

disintegration per second. See also Bq/g. A Disintegration is any nuclear transformation that emits radiation. Radiation is energy in transit in the form of high speed particles and electromagnetic waves. We encounter electromagnetic waves every day. They make up our visible light, radio and television waves, ultra violet (UV), and microwaves with a large spectrum of energies. These examples of electromagnetic waves do not cause ionizations of atoms because they do not carry enough energy to separate molecules or remove electrons from atoms. LLW is a radioactive waste because it can emit ionizing radiation. Ionizing radiation is radiation with enough energy so that during an interaction with an atom, it can remove tightly bound electrons from their orbits, causing the atom to become charged or ionized. Examples are gamma rays and neutrons.

disposal. Emplacement of waste in an appropriate facility without the intention of retrieval. Some countries use the term disposal to include discharges of effluents to the environment.

disposal, near surface. See repository, near surface.

disposal, on-site. Disposal of the nuclear facility or portions thereof within the nuclear site boundary. It includes in situ disposal (entombment) where the nuclear facility is disposed wholly or partly at its existing location; or on-site transfer and disposal where the nuclear facility or portions thereof are moved to a repository at an adjacent location on the site.

disposal facility. Synonymous with repository.

distribution coefficient, Kd. The ratio of the amount of substance sorbed on a unit mass of dry solid to the concentration of the substance in a solution in contact with the solid, assuming equilibrium conditions. The SI units are: m3/kg.


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dose. A measure of the energy deposited by radiation in a target. Absorbed dose, committed equivalent dose, committed effective dose, effective dose, equivalent dose or organ dose, depending on the context. All these quantities have the dimensions of energy divided by mass.

dose, absorbed, D. The fundamental dosimetric quantity D. The unit is J/kg, termed the gray (Gy).

dose constraint. A prospective and source related restriction on the individual dose from a source, which provides a basic level of protection for the most highly exposed individuals from a source and serves as an upper bound on the dose in optimization of protection for that source. The UK government has set a maximum dose constraint value of 0.3 mSv/year when determining applications for discharge authorization from a single new source.

dose, effective, E. A summation of the tissue equivalent doses, each multiplied by the appropriate tissue weighting factor: The unit of effective dose is J/kg, with the special name sievert (Sv). The committed effective dose is the effective dose that will be received by the person over their lifetime as a result of radionuclides taken into the body e.g. by ingestion or inhalation.

dose, equivalent, HT. The radiation-weighted dose in a tissue or organ. This takes account of the different amounts of damage caused by different types of radiation eg alpha particles, gamma radiation. The unit of equivalent dose is J/kg, termed sievert (Sv).

dose limit. See limit, dose. The value of the effective dose or the equivalent dose to individuals from planned exposure situations that shall not be exceeded. For the purposes of discharge authorizations, the UK has (since 1986) applied a dose limit of 1 mSv/year to members of the public from all man-made sources of radioactivity (other than from medical applications).

effluent. Gaseous or liquid radioactive materials which are discharged to the environment. See also discharge, authorized.

emanation. Generation of radioactive gas by the decay of a radioactive solid.

engineered barrier. See barrier.

environmental (impact) assessment. See assessment, environmental (impact).

environmental impact statement. A set of documents recording the results of an evaluation of the physical, ecological, cultural and socioeconomic effects of a planned facility (e.g. a repository) or of a new technology.

environmental monitoring. See monitoring, environmental.

equivalent dose. See dose, equivalent.

exempt waste. See waste, exempt.

exemption. The determination by a regulatory body that a source or practice need not be subject to some or all aspects of regulatory control on the basis that the exposure (including potential exposure) due to the source or practice is too small to warrant the application of those aspects. See also level, clearance.

exemption & exclusion. A number of exemption orders have been made under RSA 1993 which specify the conditions under which materials or wastes, which are defined as radioactive under the Act, can be made Exempt or excluded from some or all provisions of the Act. An important exemption order is the Substances of Low Activity (SoLA) Exemption Order. SoLA establishes a limit of 0.4 Bq/g for certain radioactive wastes that in effect is the limit below which wastes are not treated specifically as a radioactive waste for purposes of disposal. The Phosphatic Substances, Rare Earths etc. Exemption Order and the Uranium and Thorium Exemption Order are also used to set the practical lower boundaries of what becomes LLW.


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exposure. The act or condition of being subject to irradiation. Exposure can either be external exposure due to sources outside the body or internal exposure due to sources inside the body.

exposure, normal. Exposure which is expected to occur under the normal operating conditions of a facility or activity, including possible minor mishaps that can be kept under control, i.e. during normal operation and anticipated operational occurrences.

exposure, potential. Exposure that is not expected to occur with certainty but that may result from an accident at a source or owing to an event or sequence of events of a probabilistic nature, including equipment failures and operating errors.

exposure pathway. A route by which radiation or radionuclides can reach humans and cause exposure. An exposure pathway may be very simple, for example external exposure from airborne radionuclides, or involve a more complex chain, for example internal exposure from drinking milk from cows that ate grass contaminated with deposited radionuclides.

facility. See nuclear facility.

fissile material. Uranium-233, uranium-235, plutonium-239, plutonium-241, or any combination of these radionuclides. Excepted from this definition is: (a) natural uranium or depleted uranium which is unirradiated, (b) natural uranium or depleted uranium which has been irradiated in thermal reactors only.

fission product. A radionuclide produced by nuclear fission.

fixed contamination. See contamination, fixed.

flow, unsaturated. The flow of water in unsaturated soil by capillary action and gravity.

fracture. A general term for any breaks in rock whether or not it causes displacement.

fuel, nuclear. Fissionable and fertile material used in a nuclear reactor for the purpose of generating energy.

geological barrier. See barrier.

gradient, hydraulic. The change in total hydraulic head per unit distance of flow in a given direction.

groundwater. Water that is held in rocks and soil beneath the surface of the earth.

half-life, T1/2. The time taken for the quantity of a specified material (e.g. a radionuclide) in a specified place to decrease by half as a result of any specified process or processes that follow similar exponential patterns to radioactive decay.

half-life, effective, Teff. The time taken for the activity of a radionuclide in a specified place to halve as a result of all relevant processes.

half-life, radioactive. For a radionuclide, the time required for the activity to decrease, by a radioactive decay process, by half.

Harwell. The UKAEA Harwell site in Oxfordshire is an ex-RAF WWII airbase that has been used since 1946 for nuclear research, mainly in support of civilian power generation. The site is now well advanced with decommissioning. The aim is to return the site to a delicensed status by 2025.

HVLA Waste. High Volume Very Low Level Activity Waste. See main text.

HV-VLLW. High volume very low level waste. A sub-category of LLW as defined in “Policy for the Long Term Management of Solid Low Level Radioactive Waste in the United Kingdom” (DEFRA, 2007).


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HPA (NRPB) The Health Protection Agency (HPA) is an independent body that protects the health and well-being of the population. The HPA includes the ex-National Radiological Protection Board (NRPB).

HSE. Britain's Health and Safety Commission (HSC) and the Health and Safety Executive (HSE) are responsible for the regulation of almost all the risks to health and safety arising from work activity in Britain.

hydraulic conductivity, K. See conductivity, hydraulic.

hydraulic gradient. See gradient, hydraulic.

hydraulic transmissivity. See transmissivity, hydraulic.

inadvertent human intrusion. Accidental intrusion into a disposal facility without prior knowledge of the presence of the facility or accidental intrusion, without prior knowledge, into an area adjacent to the facility in such a way that it degrades the environmental safety performance of the facility.

immobilization. Conversion of waste into a waste form by solidification, embedding or encapsulation. The aim is to reduce the potential for migration or dispersion of radionuclides during handling, transport, storage and/or disposal. See also conditioning.

inert waste. Material which does not undergo any significant physical, chemical or biological transformations; does not dissolve, burn or otherwise physically or chemically react, biodegrade or adversely affect other matter with which it comes into contact in a way likely to give rise to environmental pollution or harm to human health; and whose total leachability and pollutant content and the ecotoxicity of its leachate are insignificant and in particular do not endanger the quality of any surface water or groundwater. This is defined by UK waste legislation for non radioactive wastes.

in situ disposal. See disposal, on-site.

infiltration. The downward entry of water through the ground surface into soil or rock.

institutional control. See control, institutional.

intervention. Any action intended to reduce or avert exposure or the likelihood of exposure to sources which are not part of a controlled practice or which are out of control as a consequence of an accident.

leach rate. The rate of dissolution or erosion of material or the release by diffusion from a solid, this is hence a measure of how rapidly radionuclides may be released from that material. The term usually refers to the durability of a solid waste form but also describes the removal of sorbed material from the surface of a solid or porous bed.

leach test. A test conducted to determine the leach rate of a waste form. The test results may be used for judging and comparing different types of waste forms, or may serve as input data for a long term safety assessment of a repository. Many different test parameters have to be taken into account, for example water composition and temperature.

leachate. A solution that has been in contact with waste form and, as a result, may contain radionuclides.

level, clearance. A value, established by a regulatory body and expressed in terms of activity concentration and/or total activity, at or below which a source of radiation may be released from regulatory control. See also clearance.

level, exemption. A value, established by a regulatory body and expressed in terms of activity concentration and/or total activity, at or below which a source of radiation may be granted exemption from regulatory control without further consideration.


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licence. A legal document issued by the regulatory body granting authorization to perform specified activities related to a facility or activity. The holder of a current licence is termed a licensee. A licence is a product of the authorization process, although the term licensing process is sometimes used.

limit, authorized. A limit on a measurable quantity, established or formally accepted by a regulatory body. Authorized limit has been commonly used particularly in the context of limits on discharges. See also discharge, authorized.

limit, dose. The value of the effective dose or the equivalent dose to individuals from controlled practices that shall not be exceeded.

liner. (1) A layer of material placed between a waste form and a container to resist corrosion or any other degradation of a waste package. (2) A layer of clay, plaster, asphalt or other impermeable material placed around or beneath a repository or tailings impoundment to prevent leakage and/or erosion. (3) A structural component (made, for example, of concrete or steel) on the surface of a tunnel or shaft in a repository.

LLW. See waste, low and intermediate level. Low Level Radioactive Waste. With certain specific exceptions, LLW is defined as waste which has an activity concentration in the range 0.4 – 4,000 Bq/g for alpha emitters and 12,000 Bq/g for beta-gamma emitters. Where Bq/g is Becquerel per gram, a measure of activity within the SI system equivalent to 1 disintegration per second. Where an alpha emitter is a form of radioactive decay involving emission of alpha particles (a helium nucleus). Where beta decay is a type of radioactive decay involving the emission of electrons or positrons.

Low Level Waste Repository LLWR (Drigg LLW facility). The Drigg site, located 6 km south-east of Sellafield, has operated safely for over 40 years disposing of Low Level Radioactive Wastes (LLW) from the nuclear and general industries, universities and hospitals. Drigg is operated by BN-GS (ex.British Nuclear Fuels Limited (BNFL)).

long lived waste. See waste, long lived.

long term. In radioactive waste disposal, refers to periods of time which exceed the time during which active institutional control can be expected to last.

long term stewardship. Conducting, supervising, or managing something entrusted to one's care. In the context of nuclear waste sites the phrase encompasses the activities undertaken after closure of the site to maintain and monitor the wastes in the long term.

low level waste (LLW). See waste, low and intermediate level.

LSG. Local Stakeholder Group. A group of stakeholders that meet regularly in relation to a nuclear licensed site.

Isotope. Different forms of atoms of the same element that have different numbers of neutrons in their nuclei. An element may have a number of isotopes. For example, the three isotopes of hydrogen are protium, deuterium, and tritium. All three have one proton in their nuclei, but deuterium also has one neutron, and tritium has two neutrons. Different isotopes can have different radioactive properties and present different risks.

migration. The movement of radionuclides in the environment as a result of natural processes.

minimization, waste. The process of reducing the amount and activity of radioactive waste to a level as low as reasonably achievable, at all stages from the design of a facility or activity to decommissioning, by reducing waste generation and by means such as recycling and reuse, and treatment, with due consideration for secondary as well as primary waste. See also pretreatment; treatment; volume reduction.


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model. A representation of a system and the ways in which phenomena occur within that system, used to simulate or assess the behaviour of the system for a defined purpose.

model, computational. A calculation tool that implements a mathematical model.

model, conceptual. A set of qualitative assumptions used to describe a system.

model, mathematical. A set of mathematical equations designed to represent a conceptual model.

model, pathways. A mathematical representation used to simulate the transport of radionuclides from a source to a receptor.

model, transport. A mathematical representation of mechanisms controlling the movement of finely dispersed or dissolved substances in fluids.

monitoring. Continuous or periodic measurement of radiological and other parameters or determination of the status of a system.

monitoring, environmental. The measurement and evaluation of external dose rates due to sources in the environment or of radionuclide concentrations in the environmental media.

naturally occurring radioactive material (NORM). Material containing no significant amounts of radionuclides other than naturally occurring radionuclides. The exact definition of ‘significant amounts’ would be a regulatory decision. Materials in which the activity concentrations of the naturally occurring radionuclides have been changed by human made processes are included. These are sometimes referred to as technically enhanced NORM or TENORM.

naturally occurring radionuclides. Radionuclides that occur naturally in significant quantities on earth. The term is usually used to refer to the primordial radionuclides potassium-40, uranium-235, uranium-238 and thorium-232 (the decay product of primordial uranium-236), their radioactive decay products, and tritium and carbon-14 generated by natural activation processes.

NDA. Nuclear Decommissioning Authority. A public body that oversees nuclear decommissioning in the UK on designated sites such as Harwell.

near surface disposal. See repository, near surface.

nuclear facility. A facility and its associated land, buildings and equipment in which radioactive materials are produced, processed, used, handled, stored or disposed of on such a scale that consideration of safety is required.

nuclear material. Plutonium except that with isotopic concentration exceeding 80% in plutonium-238; uranium-233; uranium enriched in the isotope 235 or 233; uranium containing the mixture of isotopes occurring in nature other than in the form of ore or ore residue; any material containing one or more of the foregoing.

nuclear waste. See waste, radioactive.

NII. Nuclear Installations Inspectorate. Under UK law (the Health and Safety at Work etc. Act 1974) employers are responsible for ensuring the safety of their workers and the public, and this is just as true for a nuclear site as for any other. This responsibility is reinforced for nuclear installations by the Nuclear Installations Act 1965 (NIA), as amended. Under the relevant statutory provisions of the NIA, a site cannot have nuclear plant on it unless the user has been granted a site licence by the Health and Safety Executive (HSE). This licensing function is administered on HSE's behalf by its Nuclear Safety Directorate (NSD).

nuclear site licence. A licence issued under the Nuclear Installations Act (see NII).

off-site. Outside the physical boundary of a site.

on-site. Within the physical boundary of a site.


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on-site disposal. See disposal, on-site.

operation. All the activities performed to achieve the purpose for which a facility was constructed.

operational period. The period during which a nuclear facility (e.g. a repository) is being used for its intended purpose until it is decommissioned or is submitted for permanent closure.

optimization. The process of determining what level of protection and safety makes exposures, and the probability and magnitude of potential exposures, ‘as low as reasonably achievable, economic and social factors being taken into account’ (ALARA).

overpack. A secondary (or additional) outer container for one or more waste packages, used for handling, transport, storage or disposal.

package, waste. The product of conditioning that includes the waste form and any container(s) and internal barriers (e.g. absorbing materials and liners), prepared in accordance with the requirements for handling, transport, storage and/or disposal.

permeability, k. The ability of a porous medium to transmit fluid.

plume. The spatial distribution of a release of airborne or waterborne material as it disperses in the environment.

porosity. The ratio of the aggregate volume of interstices in rock, soil or other porous media to its total volume.

post-closure period. The period of time following the closure of a repository and decommissioning of related surface facilities. Some type of surveillance or control will probably be maintained in this period, particularly for near surface repositories. See also closure; preclosure period.

practice. Any human activity that introduces additional sources of exposure or exposure pathways or extends exposure to additional people or modifies the network of exposure pathways from existing sources, so as to increase the exposure or the likelihood of exposure of people or the number of people exposed.

preclosure period. The period of time spanning the construction and operation of a repository up to and including the closure and decommissioning of related surface facilities. See also closure; post-closure period.

predisposal. Any radioactive waste management steps carried out prior to disposal, such as pretreatment, treatment, conditioning, storage and transport activities. Decommissioning is considered to be a part of predisposal management of radioactive waste.

pretreatment. Any or all of the operations prior to waste treatment, such as collection, segregation, chemical adjustment and decontamination.

quality assurance (QA). Planned and systematic actions necessary to provide adequate confidence that an item, process or service will satisfy given requirements for quality, for example those specified in the licence.

quality control (QC). The part of quality assurance intended to verify that systems and components correspond to predetermined requirements.

radioactive contamination. See contamination.

radioactive discharges. See discharges, radioactive.

radioactive effluent. See effluent.

radioactive half-life. See half-life, radioactive.


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radioactive material. Material designated in national law or by a regulatory body as being subject to regulatory control because of its radioactivity.

radioactive waste. See waste, radioactive. Low activity solid radioactive wastes are taken to include all wastes with an activity level lying below the defined Low Level Waste (LLW) category upper limit, but above either the levels specified for exclusion from the provisions of the Radioactive Substances Act 1993 (RSA93) or for exemption from specific regulatory action under the Act as a result of the Substances of Low Activity (SoLA) Exemption Order. This range includes, at the lower end, an officially recognised waste category termed Very Low Level Waste (VLLW).

Low Level Waste (LLW) is a waste containing radioactive materials other than those suitable for disposal with ordinary refuse, but not exceeding 4GBq/te (gigabecquerels/tonne) of alpha or 12 GBq/te of beta/gamma activity; i.e., wastes that can normally be accepted for authorised disposal at Drigg, Dounreay or other engineered landfill sites.

Very Low Level Waste (VLLW) is a waste that can be disposed of with ordinary refuse, each 0.1 cubic metre (m3) of material containing less than 400kBq (kilobecquerels) of beta/gamma activity or single items containing less than 40kBq. In the application of the VLLW upper threshold, there are separate, complementary, restrictions on the permissible content of carbon-14 and tritium; these are a factor of ten greater. VLLW disposal was originally intended for small volumes and is also known as “dustbin” disposal.

In practice, there are other streams of low activity solid radioactive waste that are disposed of to routes other than Drigg and dustbin disposal. These waste streams are associated with landfill disposal, in-situ burial on licensed nuclear sites, and incineration. The waste streams deemed suitable for landfill or in-situ burial are generally characterised by radioactivity levels well below the defined LLW upper activity threshold, and by the fact that they may arise in large volumes. Incineration is essentially treatment of LLW and VLLW prior to landfill disposal of the secondary incineration products (hearth ash and gas cleaning residues) as VLLW dustbin disposal or exempt wastes.

Landfill disposal processes for LLW were developed for those wastes arising principally in the non-nuclear sector which were above the limits for dustbin disposal and unsuitable for incineration. The activity limit is typically above VLLW, but well below the LLW upper bound. The development of this route depended on the availability of suitable landfill sites with good containment characteristics that had been subject to an environmental assessment satisfying the regulators that public safety was assured, and to an ongoing leachate monitoring programme carried out by the regulators. Disposal of LLW is subject to issue of an authorization under RSA93 by the regulators.

radioactive waste management. See waste management, radioactive.

radioactivity. The phenomenon whereby atoms undergo spontaneous random disintegration, usually accompanied by the emission of radiation.

radiological survey. See survey, radiological.

radionuclide. A nucleus (of an atom) that possesses properties of spontaneous disintegration (radioactivity). Nuclei are distinguished by their mass and atomic number.

records. A set of documents, such as instrument charts, certificates, log books, computer printouts and magnetic tapes for each nuclear facility, organized in such a way that it provides past and present representations of facility operations and activities including all phases from design through closure and decommissioning (if the facility has been decommissioned). Records are an essential part of quality assurance.


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regulatory body. An authority or a system of authorities designated by the government of a State as having legal authority for conducting the regulatory process, including issuing authorizations, and thereby for regulating the siting, design, construction, commissioning, operation, closure, decommissioning and, if required, subsequent institutional control of the nuclear facilities (e.g. near surface repositories) or specific aspects thereof.

release. See discharge.

remedial action. Action taken when a specified action level is exceeded, to reduce a radiation dose that might otherwise be received, in an intervention situation involving chronic exposure. Examples are: (a) actions which include decontamination, waste removal and environmental restoration of a site during decommissioning and/or closure efforts; (b) actions taken beyond stabilization of tailings impoundments to allow for other uses of the area or to restore the area to near pristine conditions.

remediation. See cleanup.

repository. A nuclear facility where waste is emplaced for disposal.

repository, near surface. A facility for disposal of radioactive waste located at or within a few tens of metres from the earth’s surface.

restoration. See cleanup.

retardation. A reduction in the rate of radionuclide movement through the soil due to the interaction (e.g. by sorption) with an immobile matrix.

retardation coefficient, Rd. A measure of capability of porous media to impede the movement of a particular radionuclide being carried by fluid.

retrievability. The ability to remove waste from where it has been emplaced.

risk. A multiattribute quantity expressing hazard, danger or chance of harmful or injurious consequences associated with actual or potential exposures. It relates to quantities such as the probability that specific deleterious consequences may arise and the magnitude and character of such consequences. (2) The combination of the frequency, or probability, of occurrence and the consequence of a specified hazardous event. The concept of risk always has two elements: the frequency or probability with which a hazardous event occurs and the consequences of the hazardous event. Risk = Probability x Consequence.

risk assessment. See assessment, risk.

rock. In geology, any mass of mineral matter, whether consolidated or not, which forms part of the earth’s crust. Rocks may consist of only one mineral species, in which case they are called monomineralic but they usually consist of several mineral species.

RWMAC The Radioactive Waste Management Advisory Committee (RWMAC) was established in 1978 to offer independent advice to Ministers on radioactive waste management issues. Members of the Committee were drawn from a wide range of backgrounds and specialisms including radioactive waste management, radiological protection, earth sciences, environmental law & planning, medical physics and social sciences. Each year until 2004, RWMAC undertook a programme of work commissioned by Government Ministers.

safety case. An integrated collection of arguments and evidence to demonstrate the safety of a facility. This will normally include a safety assessment, but could also typically include information (including supporting evidence and reasoning) on the robustness and reliability of the safety assessment and the assumptions made therein.


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safety culture. The assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, protection and safety issues receive the attention warranted by their significance.

safety report. A document required from the operating organization by the regulatory body containing information concerning a nuclear facility (e.g. a repository), the site characteristics, design, operational procedures, etc., together with a safety analysis and details of any provisions needed to restrict risk to personnel and the public.

saturated zone. See zone, saturated.

scenario. A postulated or assumed set of conditions and/or events. They are most commonly used in analysis or assessment to represent possible future conditions and/or events to be modelled, such as possible accidents at a nuclear facility, or the possible future evolution of a repository and its surroundings.

screening. A type of analysis aimed at eliminating from further consideration factors that are less significant for the purpose of the analysis, in order to concentrate on the more significant factors. Screening is usually conducted at an early stage in order to narrow the range of factors needing detailed consideration in an analysis or assessment.

secondary waste. See waste, secondary.

segregation. An activity where waste or materials (radioactive and exempt) are separated or are kept separate according to radiological, chemical and/or physical properties which will facilitate waste handling and/or processing. For example, it may be possible to segregate radioactive from exempt material and thus reduce the waste volume.

sensitivity analysis. See analysis, sensitivity.

shielding. A material interposed between a source of radiation and persons, or equipment or other objects, in order to absorb radiation and thereby reduce radiation exposure.

short lived waste. See waste, short lived.

site. The area containing, or under investigation for its suitability for, a nuclear facility (e.g. a repository). It is defined by a boundary and is under effective control of the operating organization.

site characterization. See characterization, site.

solidification. Immobilization of gaseous, liquid or liquid-like materials by conversion into a solid waste form, usually with the intent of producing a physically stable material that is easier to handle and less dispersible. Calcination, drying, cementation, bituminization and vitrification are some of the typical ways of solidifying liquid waste. See also conditioning; immobilization.

solidified waste. See waste, solidified.

solubility. The amount of a substance that will dissolve in a given amount of another substance. The solubility of a waste form or a radionuclide is an important factor in determining the potential migration of radionuclides from a disposal area.

sorption. The interaction of an atom, molecule or particle with the surface of a solid. A general term including absorption (sorption taking place largely within the pores of a solid) and adsorption (surface sorption with a non-porous solid). The processes involved may also be divided into chemisorption (chemical bonding with the substrate) and physisorption (physical attraction, for example by weak electrostatic forces).

source. (1) Anything that may cause radiation exposure, such as by emitting ionizing radiation or by releasing radioactive substances or materials. (2) More specifically, radioactive material used as a source of radiation.


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source, natural. A naturally occurring source of radiation, such as the sun and stars (sources of cosmic radiation) and rocks and soil (terrestrial sources of radiation).

source term. A mathematical expression used to denote information about the actual or potential release of radiation or radioactive material from a given source, which may include further specifications, for example the composition, the initial amount, the rate and the mode of release of the material.

specific activity. See activity, specific.

storage. The holding of spent fuel or of radioactive waste in a facility that provides for its containment, with the intention of retrieval (3). Storage is by definition an interim measure, and the term interim storage would therefore be appropriate only to refer to short term temporary storage when contrasting this with the longer term fate of the waste. Storage as defined above should not be described as interim storage.

surface water. Water which fails to penetrate into the soil and flows along the surface of the ground, eventually entering a lake, a river or the sea.

survey, radiological. An evaluation of the radiological conditions and potential hazards associated with the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

sustainability The concept of meeting the needs of the present without compromising the ability of future generations to meet their needs. The term originally applied to natural resource situations, where the long term was the focus. Today, it applies to many disciplines, including economic development, environment, food production, energy, and social organization. Basically, sustainability/sustainable development refers to doing something with the long term in mind.

transmissivity, hydraulic. The rate at which water is transmitted through a unit width of a water conducting feature (e.g. an aquifer) under a unit hydraulic gradient.

transmutation. The conversion of one element into another. Transmutation is under study as a means of converting longer lived radionuclides into shorter lived or stable radionuclides. The term actinide burning is used in some countries.

transport, radionuclide. The movement (migration) of radionuclides in the environment, for example radionuclide transport by groundwater. This could include processes such as advection, diffusion, sorption and uptake. This usage does not include intentional transport of radioactive materials by humans (transport of radioactive wastes in casks, etc). See also migration.

treatment. Operations intended to benefit safety and/or economy by changing the characteristics of the waste. Three basic treatment objectives are: volume reduction, removal of radionuclides from the waste and change of composition. Treatment may result in an appropriate waste form.

UKAEA The United Kingdom Atomic Energy Authority (UKAEA) was incorporated as a statutory corporation in 1954 and pioneered the development of nuclear energy in the UK. Today we are responsible for managing the decommissioning of the nuclear reactors and other radioactive facilities used for the UK's nuclear research and development programme in a safe and environmentally sensitive manner. UKAEA is a non-departmental public body, funded mainly by its lead department the Department of Trade and Industry under contract to the NDA.

uptake. A general term for the processes by which radionuclides enter one part of a biological system from another. Used in a range of situations, particularly in describing the overall effect when there are a number of contributing processes, for example root uptake, the transfer of radionuclides from soil to plants through the plant roots.

uranium, depleted. Uranium containing a lesser mass percentage of uranium-235 than in natural uranium.


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uranium, enriched. Uranium containing a greater mass percentage of uranium-235 than 0.72%.

uranium, natural. Chemically separated uranium containing the naturally occurring distribution of uranium isotopes (approximately 99.28% uranium-238 and 0.72% uranium-235 by mass).

very low level waste (VLLW). See waste, very low level.

volume reduction. A treatment method that decreases the physical volume of a waste. Volume reduction is employed because it is economical and facilitates subsequent handling, storage, transport and disposal of the waste. Typical volume reduction methods are mechanical compaction, incineration and evaporation. Volume reduction of a given waste results in a corresponding increase in radionuclide concentration. The total volume of waste may also be reduced through decontamination (with subsequent exemption) or through the avoidance of waste generation. See also minimization, waste.

waste. Material in gaseous, liquid or solid form for which no further use is foreseen.

waste, alpha bearing. Radioactive waste containing one or more alpha emitting radionuclides. Alpha bearing waste can be short lived or long lived.

waste, exempt. Waste released from regulatory control in accordance with exemption principles. See also clearance levels; exemption.

waste, long lived. Radioactive waste that contains significant levels of radionuclides with half-lives greater than 30 years. Typical characteristics are long lived radionuclide concentrations exceeding limitations for short lived waste.

waste, low level (LLW). See waste, low and intermediate level.

waste, mixed. Radioactive waste that also contains non-radioactive toxic or hazardous substances.

waste, radioactive. For legal and regulatory purposes, waste that contains or is contaminated with radionuclides at concentrations or activities greater than clearance levels as established by the regulatory body. It should be recognized that this definition is purely for regulatory purposes and that material with activity concentrations equal to or less than clearance levels is radioactive from a physical viewpoint — although the associated radiological hazards are considered negligible.

waste, secondary. A form and quality of waste that results as a by-product from processing of waste.

waste, short lived. Radioactive waste that does not contain significant levels of radionuclides with half-lives greater than 30 years.

waste, very low level (VLLW). Radioactive waste considered suitable by the regulatory body for authorized disposal, subject to specified conditions, with ordinary waste in facilities not specifically designed for radioactive waste disposal.

waste acceptance requirements. Quantitative or qualitative criteria specified by the regulatory body, or specified by an operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. Waste acceptance requirements might include, for example, restrictions on the activity concentration or the total activity of particular radionuclides (or types of radionuclide) in the waste or requirements concerning the waste form or waste package.

waste characterization. See characterization, waste.

waste form. Waste in its physical and chemical form after treatment and/or conditioning (resulting in a solid product) prior to packaging. The waste form is a component of the waste package.


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waste generator. The operating organization of a facility or activity that generates waste. See also operator.

waste inventory. Quantity, radionuclides, activity and waste form characteristics of wastes for which an operator is responsible.

waste management, radioactive. All activities, administrative and operational, that are involved in the handling, pretreatment, treatment, conditioning, transport, storage and disposal of radioactive waste.

water table. The upper surface of a zone of groundwater saturation.

zone, saturated. A subsurface zone in which all the interstices are filled with water. This zone is separated from the unsaturated zone, i.e. the zone of aeration, by the water table. See also zone, unsaturated.

zone, unsaturated. A subsurface zone in which at least some interstices contain air or water vapour, rather than liquid water. Also referred to as the ‘zone of aeration’. See also zone, saturated.


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Annexes

A Radiation, People and the Environment (IAEA, 2004)

B Suitability Assessment – Galson Sciences

C ENRMF, IRRs 1999, Radiation Risk Assessment for Low Level Waste Disposal, HPA

D Dose Rate calculations in support of Low Level Waste disposal authorisation, TSG(09)0487

E SNIFFER Methodology Information

F Copy of Application Form

G Example Capacity Calculation Layout

H Calculation of dose rate at landfill, TSG(09)0488

I Baseline Groundwater and Leachate Sample Results

J Capability Statements

Annex A

Radiation, People and the Environment IAEA

This booklet is provided as a primer for readers who are new to the subject of radioactivity. The booklet is not specific to the application. The booklet has not been included as a hardcopy version and can be read on downloaded at: http://www.iaea.org/Publications/Booklets/RadPeopleEnv/radiation_booklet.html

Annex B

Suitability Assessment – Galson Sciences

0820-2 Version 2

Radiological Assessment for Disposal of Solid Low-level

Radioactive Waste at the Landfill at East Northants Resource Management Facility

R D Wilmot & D Reedha

July 2009

Galson S C I E N C E S L T D

5 Grosvenor House, Melton Road, Oakham, Rutland LE15 6AX, UK Tel: +44 (1572) 770649 Fax: +44 (1572) 770650 www.galson-sciences.co.uk

0820-2 Version 2

Radiological Assessment for Disposal of Solid Low-level

Radioactive Waste at the Landfill at East Northants Resource Management Facility

Report History

This document has been prepared by Galson Sciences Limited for UKAEA Harwell under the terms of Contract No. CF12/07.

Radiological Assessment for Disposal of Solid Low-level Radioactive Waste at the Landfill at East Northants Resource Management Facility

Version: Date: Principal Author: R D Wilmot

Reviewed by: D Reedha

Approved by: D A Galson

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0820-2 Version 2 14 July 2009

Date 14 July 2009

Date 14 July 2009

Date 14 July 2009

Radiological Assessment 0820-2 Version 2

Galson Sciences Limited i 14 July 2009

Executive Summary

The nuclear decommissioning industry has significant future arisings of decommissioning wastes that fall into the category of low level waste. In the recently published “Policy for the Long Term Management of Solid Low Level Radioactive Waste” (Defra 2007), the government has confirmed the acceptability of the disposal of LLW to landfill including a new subset of LLW classified as high-volume very low level waste.

Augean plc operates a hazardous waste disposal facility at the East Northants Resource Management Facility (ENRMF) in Northamptonshire and proposes that the site is used for the disposal of LLW with a specific activity up to 200 Bq / g. This report presents a radiological assessment for this site in order to assess the potential consequences of disposal and determine the radiological capacity of the site for different waste streams. In this report an example waste stream of LLW from UKAEA Harwell has been used for illustration.

This report investigates the suitability of the landfill as a disposal route for LLW with a specific activity up to 200 Bq / g, based on the established approaches used previously for assessing “Special Precautions Burial” (SPB) which has been used for wastes of comparable activity. The methodology has been modified to take account of the likely waste volumes and also for certain site-specific aspects that differ from the generic assumptions used in the available model.

The assessment model is a simplified and conservative model of the events and processes that will or might take place during and after operations. A number of simplifying assumptions are therefore required in order to represent the site and its surroundings. These assumptions are outlined in the report and the equations and parameter values used in the model are reported. Where there are significant uncertainties regarding aspects of the site, a range of assumptions have been used to test the sensitivity of the model.

The report presents specific dose calculations. These are the doses that would be received from a disposal of 1MBq of each radionuclide of interest. The specific dose depends on the pathway by which radionuclides are released and the time of the release. Specific doses have been calculated for the groundwater release, intrusion, irradiation and gas release pathways and for pathways relating to leachate management.

Specific doses are used to calculate the capacity of the site for individual radionuclides, based on a public dose criterion of 20 Sv / year for normal release scenarios and a 3 mSv / year dose criterion for pathways resulting from human intrusion. An illustrative overall site radiological capacity has also been calculated using preliminary data for the UKAEA Harwell Meashill Trenches waste stream.

In addition to specific dose calculations for the potential exposure of humans to releases from the site, assessments of dose rates have been made for wildlife in the vicinity of the site. All of the dose rates calculated, for both terrestrial and the


Galson Sciences Limited ii 14 July 2009

freshwater ecosystems, show that no organisms or wildlife groups are likely to receive dose rates in excess of the internationally recognised criterion of 10 μGy / hour.

The actual use of the ENRMF for disposal of LLW waste remains subject to discussions with the Environment Agency and other stakeholders.


Galson Sciences Limited iii 14 July 2009

Contents

Executive Summary..................................................................................................... i

1 Introduction............................................................................................................1 1.1 Project background..........................................................................................1 1.2 Approach .........................................................................................................1 1.3 Structure of report............................................................................................2

2 Background ............................................................................................................3 2.1 Site...................................................................................................................3

2.1.1 Design and operations .........................................................................3 2.1.2 Geology and hydrogeology .................................................................6 2.1.3 Biosphere and receptors ....................................................................10

2.2 Wastes............................................................................................................11

3 Assessment Methodology.....................................................................................14 3.1 Summary of SNIFFER methodology ............................................................14

3.1.1 Assessment framework......................................................................14 3.1.2 Scenarios............................................................................................16 3.1.3 Dose calculations...............................................................................18

3.2 Modifications to SNIFFER methodology .....................................................21 3.2.1 Dose criteria and compliance points..................................................21 3.2.2 Barrier design and performance ........................................................23 3.2.3 Distribution of waste .........................................................................24 3.2.4 Leachate concentration......................................................................25

3.3 Supplementary calculations...........................................................................26

4 Assessment Data and Assumptions ....................................................................28 4.1 Site characteristics .........................................................................................28

4.1.1 Size of site .........................................................................................28 4.1.2 Construction ......................................................................................28 4.1.3 Barrier................................................................................................28 4.1.4 Cap.....................................................................................................30 4.1.5 Operational period .............................................................................31 4.1.6 Leachate collection and management procedures .............................32 4.1.7 Leachate spillage ...............................................................................32 4.1.8 Control over future site use ...............................................................35

4.2 Hydrogeological setting.................................................................................35 4.2.1 Underlying geology ...........................................................................35 4.2.2 Unsaturated zone characteristics .......................................................36 4.2.3 Saturated zone characteristics............................................................36 4.2.4 Groundwater discharges ....................................................................37 4.2.5 Stream and river characteristics.........................................................37 4.2.6 Groundwater flow and radionuclide transport...................................38

4.3 Other scenarios and pathways .......................................................................40 4.3.1 Gas.....................................................................................................40


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4.3.2 Fire.....................................................................................................41 4.3.3 Barrier failure ....................................................................................41 4.3.4 Site remediation and re-engineering..................................................42 4.3.5 Bathtubbing .......................................................................................42

5 Dose Calculations .................................................................................................43 5.1 Groundwater pathway ...................................................................................43 5.2 Irradiation pathway........................................................................................45 5.3 Intrusion.........................................................................................................46 5.4 Leachate management and spillage ...............................................................48

5.4.1 Leachate management .......................................................................48 5.4.2 Leachate spillage ...............................................................................49 5.4.3 Aerosol pathway................................................................................51

5.5 Gas pathway ..................................................................................................52 5.7 Dose rates to wildlife.....................................................................................53

6 Radiological Capacity..........................................................................................59 6.1 Introduction ...................................................................................................59 6.2 Radionuclide-specific radiological capacities ...............................................60 6.3 Overall radiological capacity.........................................................................63

7 References.............................................................................................................66

Appendix A Dose calculations ............................................................................67 A.1 Doses during site operations..........................................................................67 A.2 Doses to site residents after closure...............................................................68 A.3 Doses during and after excavation of waste ..................................................69

A.3.1 Dose to the Excavator........................................................................69 A.3.2 Dose to Site Resident after Excavation .............................................71

A.4 Doses arising from use of contaminated groundwater ..................................74

Appendix B Radionuclide-specific data ............................................................76

Appendix C Sensitivity Studies ..........................................................................79 C.1 Groundwater Pathway ...................................................................................79 C.2 Leachate Spillage...........................................................................................87


Galson Sciences Limited 1 14 July 2009

Radiological Assessment for Disposal of Solid Low-level Radioactive Waste to the Landfill at East Northants

Resource Management Facility

1 Introduction

1.1 Project background

The nuclear decommissioning industry has significant future arisings of decommissioning wastes that fall into the category of low level waste. In the recently published “Policy for the Long Term Management of Solid Low Level Radioactive Waste” (Defra 2007), the government has confirmed the acceptability of the disposal of LLW to landfill including a new subset of LLW classified as high-volume very low level waste.

Augean plc operates a hazardous waste disposal facility at the East Northants Resource Management Facility (ENRMF) in Northamptonshire and proposes that the site is used for the disposal of LLW with a specific activity up to 200 Bq / g. This report presents a radiological assessment for this site in order to assess the potential consequences of disposal and determine the radiological capacity of the site for different waste streams. In this report an example waste stream of LLW from UKAEA Harwell has been used for illustration.

The radiological assessment presented in this report is based on the established approaches used previously for assessing “Special Precautions Burial” (SPB) which has been used for wastes of comparable activity. The methodology has been modified to take account of the likely waste volumes and also for certain site-specific aspects that differ from the generic assumptions used in the available model.

The actual use of the ENRMF for disposal of LLW remains subject to discussions with the Environment Agency and other stakeholders.

1.2 Approach

The assessment presented in this report comprises two stages:

Dose assessment

Radiological capacity assessment



The dose assessment is based on an established methodology developed for SNIFFER (SNIFFER 2006a). Because the inventory, or amount of each radionuclide, is not known at this stage, the dose calculations are based on unit disposals (1 MBq) of each radionuclide, and the results are expressed as specific doses (µSv y-1 per MBq).

Radiological capacity is the amount of radioactive material that can be consigned to a site without any of the potentially exposed groups considered receiving a dose above a specified criterion.

For a single radionuclide, the radiological capacity (in Bq) is calculated by dividing the dose criterion (expressed in µSv y-1) by the maximum specific dose for that radionuclide (expressed in µSv y-1 per MBq). In the case of waste streams, however, in which the proportions of different radionuclides are fixed, the calculation of capacity must consider both the specific dose and the ratio of radionuclides in the waste stream. This means that there is not a single radiological capacity for the site and this assessment provides only an illustrative overall site radiological capacity based on preliminary data for the UKAEA Harwell Meashill Trenches waste stream.

The radiological capacity determines how much radioactivity can be disposed of to the site without causing significant doses to workers or members of the public. The actual amount of waste that is disposed depends upon the specific activity of the waste (Bq / g or MBq / te). For very low activity wastes, the physical capacity of the site will impose a limit, and the radiological capacity may not be reached. For higher activity wastes, there are other constraints relating to waste handling and transport that impose an effective limit of 200 Bq / g. Because of the heterogeneous nature of the wastes envisaged for disposal at the ENRMF, the average activity would be significantly below 200 Bq / g but this value can be used to estimate the volume of the site that could be used for LLW disposals.

1.3 Structure of report

Following this Introduction, Section 2 of the report summarises available information concerning the proposed disposal site and also summarises information on the illustrative waste stream used in the radiological capacity calculations. Section 3 summarises the SNIFFER methodology used for the radiological assessment, including a review of where the assumptions in the SNIFFER assessment model have been modified for this site-specific radiological assessment. Section 4 sets out the key assumptions, equations and parameter values for the scenarios adopted for the calculations of potential radionuclide releases from the site. Section 5 presents specific doses calculated for these release scenarios and also presents calculated dose rates for wildlife around the site. Section 6 presents radiological capacities based on the principal exposure pathways. Appendix A presents details of the exposure models used to calculate specific doses, including the equations and parameter values used. Appendix B presents the radionuclide-specific data used in the dose calculations. Appendix C presents results from sensitivity studies undertaken as part of the assessment.



2 Background

This section provides background information for the radiological assessment.

Section 2.1 presents information on the operation and construction of the ENRMF, together with a summary of the geological and hydrogeological setting. A brief description of the environmental setting of the site and the populations that might be exposed to any release from the site is also presented.

For the purpose of illustrating the overall radiological capacity of the site, and for estimating the volumes of waste that could be disposed, information on one waste stream currently being considered for disposal via the landfill route is presented in Section 2.2.

2.1 Site

This section provides a brief description of the ENRMF site and its surroundings. This information provides the basis for the assessment of potential doses from disposals of radioactive waste, and is derived principally from the Environmental Statement (Bullen Consultants Ltd, 2005) and Hydrogeological Risk Assessment (HRA) (ESI, 2004) made available by Augean.

The ENRMF (formerly known as King’s Cliffe or Slipe Clay Pit) landfill site is about 6 km from Stamford and began as a landfill site in 2002. Prior to that date it had been a clay pit used for the extraction of refractory clays. The landfill was initially used for the co-disposal of hazardous and non-hazardous wastes but, following a change in legislation, became a hazardous waste disposal site in 2004.

2.1.1 Design and operations

The site is divided into a series of cells, separated by clay bunds (Figure 2.1). These cells are progressively, constructed, infilled with waste and temporarily capped.

The overall volume of the site is in the order of 1.8 x 106 m3 (Table 2.1). At the time that the site became a hazardous waste site, Cells 1 and 2 were already full and had temporary caps. Hazardous waste only disposals started in Cell 3 which is now full. Cells 1, 2 and 3 have been permanently capped and partially restored. At the time the radiological assessment was initiated (October 2007), there was about 700,000 m3 of remaining capacity. Planning permission for the site imposes a limit of 249,999 m3 on annual disposals and completion of inputs by 2013.



Figure 2.1 Disposal cells at the ENRMF landfill site. Cells 1, 2 and 3 have been capped.



Cell Basal area

(m2) Surface area

(m2) Void volume

(m3)

1A 12,866 15,195 268,072

Western extension

2,000 2,396 13,080

1B 11,934 11,934 193,865

2A 12,449 12,449 187,900

2B 10,464 11,822 143,010

3A 8,412 13,922 160,877

3B 9,090 11,624 174,685

4A 11,144 14,097 174,564

4B 12,552 12,552 206,102

5A 6,929 11,669 144,211

5B 8,294 9,887 147,221

Table 2.1 Cell areas and volumes at the ENRMF landfill site.

Cell construction comprises:

Leachate drainage layer. This varies between cells, but in Cells 3, 4 and 5 it will be 500 mm of crushed granite.

Artificial sealing liner. All cells include a 2 mm thick high density polyethylene (HDPE) geomembrane.

Artificial mineral layer. All cells (except the western extension and Cell 2b) include at least 1.5m thickness of artificially emplaced geological barrier (Upper Lias clay sourced from the Slipe Clay Pit). This clay will be placed with a maximum design permeability of 3x10-10 m/s and thickness of 1.5m or equivalent to meet the Environmental Permit requirement for permeability of less than or equal to 1.0 x l0-9 m/s with a thickness of more than or equal to 5m.

The geological barrier also includes between 3 and 8 m of natural geological barrier (unsaturated zone) above the water table.

Waste is examined on receipt at the site and after checking for compliance with the waste acceptance criteria is deposited within the current filling area. Temporary haul roads are formed within the fill area ensuring the de1ivery trucks do not traffic, unnecessarily, on the waste surface.



Following deposit of the waste the material is levelled and compacted and intermediate inert cover materials placed over the top. Cover materials are currently sourced on site and are primarily made up of waste clays from the mineral excavation operations. Alternative cover materials are being investigated. Wastes are deposited in controlled layers to ensure adequate compaction and minimise settlement.

Leachate forms in both open and capped cells as rainwater infiltrates through the cap, if present, and the waste. Leachate levels are monitored through a series of boreholes across the site, and excess leachate is removed by pumping. The leachate can be managed by recirculation in Cells 1 and 2 but is otherwise removed by tanker for treatment off site. The amount of leachate allowed to accumulate is regulated, with trigger levels of 2 m head in the sumps and 1 m in the monitoring wells. The Annual Monitoring Report for 2007 shows that these levels are maintained except during periods of unusually high rainfall.

The “Kings Cliffe Landfill Site Annual Monitoring Report 2007” (ENRMF was formerly known as Kingscliffe or Slipe Clay pit) reports that approximately 5000 tonnes of leachate were abstracted from the site during 2007 and transferred to a disposal facility. The “Pollution Inventory reporting form” submitted to the Environment Agency for the site reports 5402 tonnes of landfill leachate (EWC code 19 07 03) being discharged in 2007.

Leachate abstracted from the site is transferred by tanker to an off-site facility for treatment and discharge. The current arrangement is for transfer to a biological treatment plant at Avonmouth, discharge to trade effluent sewer and further treatment at a water treatment plant. There are feasibility studies underway for use of an alternative off-site facility and for construction of an on-site leachate treatment plant.

The closed cells are capped with a composite cap consisting of a gas drainage layer, clay regulating layer, geotextile protector, geosynthetic clay liner, LDPE geomembrane liner and soil cover.

There are lagoons on site that receive surface water that does not enter the disposal cells. Water from one lagoon is used for dust suppression. After capping, surface water will be directed to settling ponds and then discharged to a swallow hole (northern slopes) or surface water (southern areas).

Following capping, the landfill site will enter a post closure managed stage. This stage will include the maintenance of leachate levels and gas abstraction, and will continue until it can be confirmed that the site no longer represents a significant risk of pollution of the environment or harm to human health. Leachate, surface water and groundwater quality will be monitored throughout the post closure managed stage.

2.1.2 Geology and hydrogeology

The regional geology comprises a sequence of Jurassic sedimentary rocks, including limestones, clays and mudstones (Table 2.2). On higher ground, the Jurassic rocks are overlain by Pleistocene glacial clays.



ENRMF was originally a clay pit, which exploited the refractory clays at the base of the Upper Estuarine Series. The base of the pit is therefore effectively the top of the Lincolnshire Limestone, necessitating the need for an artificial geological barrier over most of the site, as described above.

Group Formation Thickness Lithology Notes

Glacial Till (Pleistocene)

0 – 7m Yellow-brown clay with chalk and limestone

fragments

Patchy distribution; not present in the south east

comer of the site

Great Oolite Limestone (formerly Blisworth

limestone)

0 – 1.9 m Yellow micritic limestone

Gre

at O

olite

Gro

up

Upper Estuarine Series

9 – 12 m Grey-brown firm silty mudstone

Locally fissured. Has been excavated to win

clay from the base of the unit.

Upper and Lower Lincolnshire Limestone

15 – 20 m Oolitic, pisolitic and massive limestones

interbedded with sandy limestones

The only formation remaining beneath the excavation. Fractured, with some small voids

and fissures.

Grantham Formation 0 – 2m Fine sands, sills, silty clays and mudstones.

Noted in most boreholes drilled at King’s Cliffe.

Infe

rior

Ool

ite

Gro

up

Northampton Sand ~2m Sands and sandstones with siderite nodules,

some subordinate limestones and silts.

Sometimes present at the base of the Lower

Lincolnshire Limestone.

Lias Group

Upper Lias >2m Grey mudstones and clays with subordinate thin

limestone bands

Table 2.2 Outline geological succession in the region of the ENRMF landfill site.

Drilling near the site confirmed the presence of the Blisworth Limestone (Great Oolite Limestone) to the east of the site. The underlying Upper Estuarine Series (corresponding to the material exploited at the clay pit) ranges in thickness from 4.2 m to 12.9 m, with a typical thickness of 11.5m.

The Upper Estuarine Series is mainly argillaceous which has been divided into two parts. The lower part is the Lower Freshwater Sequence which appears to be devoid of marine fossils and is composed of dark or brown grey mudstones and seatearths with abundant rootlets and listric surfaces. Bioturbated laminae, load casts and sand filled cracks are common sedimentary features. This lower sequence is around 5 m thick. The upper division of the Series is composed of a cyclical sequence of marine and brackish/freshwater sediments. The marine beds are composed of shelly



limestone and mudstones and the brackish sediments constitute mudstones and siltstones. This upper division ranges between 1 and 8 m in thickness.

The Upper Estuarine Series is recorded in most borehole logs around the site. The interface between this unit and the overlying Glacial Boulder Clay and weathered clayey soils is difficult to discern.

The upper part of the Lincolnshire Limestones underlying the Upper Estuarine Series and forming the base of the clay pit comprises a sequence of oolitic limestones. The lower part of the Lincolnshire Limestones is composed of fine grained sandy limestones. The Lincolnshire Limestone has been proven in the majority of the boreholes drilled within the site boundaries and ranges from 9 to 21 m in thickness.

Discrete horizontal fissuring associated with bedding is present within the limestones. Fissures are generally clean and smooth with infilling material composed of fragments of limestone, crystalline calcite sand, silt and clay. Extensive fissuring and fragmented limestone has been seen in cores at certain elevations with decalcification producing cavities within the limestone. During preparation of the formation base of Cell 3A, two fissures were observed on the exposed surface of the limestone. These fissures were up to 1 m in length, 8 cm wide and estimated to be 30 cm deep. These fissures were infilled with a gravel concrete mix before emplacement of the basa1 barrier.

Beneath the Lincolnshire Limestones, the Grantham Formation is somewhat discontinuous around King’s Cliffe, and often the Lower Lincolnshire Limestone is in direct contact with the Northampton Sands. Below the Northampton Sand is the Upper Lias.

Glacial Till deposits are relatively extensive, especially on higher ground to the southwest, east and southeast of the site. Glacial Till was encountered to the southwest of the quarry, where it has been described as firm to stiff, dark brown and grey, slightly sandy clay with limestone gravels, with a thickness of 8.6 m.

The principal hydrogeological units in the Jurassic rocks of the area are listed in Table 2.3. Although groundwater vulnerability maps show the ENRMF landfill site to be in a non-aquifer area (Upper Estuarine Series), the removal of clay during quarrying means that the base of the landfill situated on the Lincolnshire Limestone, classified as a Major Aquifer by the Environment Agency.

The hydraulic properties of the Lincolnshire Limestone in the East Midlands area are summarised in Table 2.4. Because of the removal of the overlying Upper Estuarine Series, the Limestone in the region of the site is unconfined. Regional groundwater flow is down dip towards the confined area in the east (Allen et al., 1997).

The Lincolnshire Limestone aquifer is characterised by fracture flow and there are also swallow holes in the vicinity which allow rapid flow of water from surface to the water table. The depth to the water table is estimated to be between 4m to 8m, and the thickness of the aquifer at the site is 15m to 22m. Permeability is lower than would be



expected in other regions and the aquifer, although used for agricultural supply, is not used for public water supply in this area.

Formation Hydrogeological classification

Comment

Great Oolite Limestone Aquifer Minor aquifer. Unsaturated

Upper Estuarine Series (Silty Mudstone)

Aquitard Potentially semi-confines the underlying Lincolnshire Limestone aquifer in the local area and to the south and southeast. Elsewhere may be absent.

Lincolnshire Limestone Aquifer Major aquifer (EA classification). Locally has limited thickness. Dominated by fracture flow. Recharge through swallow holes extending through Upper Estuarine Series. The semi-confining clays of the Upper Estuarine Series were removed at quarry site resulting in water tab1e conditions. Elsewhere to the east conditions are confined/artesian.

Grantham Formation Aquitard May be absent.

Northampton Sand Aquifer Minor aquifer. It is semi-confined by the silts/clays of the Lower Estuarine Series. Where this is absent, it is in hydraulic continuity with the Lower Lincolnshire Limestone Aquifer.

Upper Lias Aquiclude Basal Aquiclude to the Oolite Series aquifer/aquitard system.

Table 2.3 Principal hydrogeological units in the region of the ENRMF landfill site.



Hydraulic Property Sample method Samples Range (Mean)

Transmissivity (m /day)

Pumping tests 59

1 – 14 000* (665) a

Porosity Core data 415 0.13 – 0.22 (0.18) b

Fracture porosity Estimates 0.004 – 0.01

Storage Pumping tests 37 2x10-7 – 6x10-1

Hydraulic conductivity (m/day)

Core data 415 <1.9x10-6 – 0.17 (1.33x10-4) a

a = Geometric mean, b = Arithmetic mean. *Confidence in values above 10 000 m2/day is low (Allen et al., 1997).

Table 2.4 Hydraulic data for the Lincolnshire Limestone (Allen et al., 1997)

There are several springs at the base of the Great Oolite Limestone 1 - 3 km to the south of the site that drain into the Willow Brook, and at the base of the Lincolnshire Limestone that drain towards the River Nene. There are also springs to the north-east that drain towards the River Welland, but these are up gradient of the site. There is a swallow hole north of the site that will be used after site closure as a discharge point for surface water draining from the northern part of the site.

2.1.3 Biosphere and receptors

The site is located on relatively high ground (85 m AOD) between the valleys of the River Nene to the south and the River Welland to the north. The general topography around the site is relatively flat, becoming more rolling towards the main valleys.

Locally, the site lies between woodland to the north and arable farmland to the south and west. Immediately to the east of the site there is a road, with a few cottages, farm buildings and light industrial operations on the opposite side of the road. The nearest villages are Duddington (1.5 km to the north-west) and King’s Cliffe, some 2.5 km to the south.

There are no significant natural water courses within 300 m of the site, although there is a pond close to the southern boundary of the site. Surface water drains will be constructed around the site perimeter after capping. The nearest significant natural water course is the River Welland, some 2.5 km to the west.

There are eleven licensed surface water abstraction points within 5 km of the site, but only one of these is within 3 km of the site, where a nursery is licensed to abstract about 1,100 m3 per year for agricultural and private supply purposes.

There are eight licensed abstraction boreholes within 3 km of the site, the closest that is in use being at Law’s Lawn, some 1,487 m from the site. The majority of licenses, including that for Law’s Lawn, are for agricultural usage. There are no public supply boreholes in the vicinity of the site.



2.2 Wastes

The nuclear industry is generating significant quantities of radioactive waste from decommissioning, and these types of waste will continue to arise as decommissioning continues. Some of these wastes are potentially suitable for disposal at landfill sites. Wastes from other activities and industries, such as hospitals, universities and radiochemical manufacture, could also be considered for disposal at such sites.

To ensure that the potential radiological consequences of the disposal of a representative range of LLW can be assessed, the radionuclides listed in Table 2.5 have been considered in the radiological assessment. Radionuclides with half-lives less than one year have not been explicitly assessed. Where such radionuclides arise from ingrowth, they are included through the assumption that they will be in secular equilibrium with the parent radionuclide, and the dose coefficients used are adjusted accordingly.

Radionuclide Half-life (years)

Daughters assumed to be in secular equilibrium

3H 12.3 14C 5,730 36Cl 3.01E+05 55Fe 2.73 60Co 5.27 63Ni 96.0 90Sr 28.8 90Y 94Nb 2.00E+04 99Tc 2.11E+05 106Ru 1.02 106Rh 108mAg 418 125Sb 2.80 126Sn 2.07E+05 126Sb 129I 1.57E+07 133Ba 10.7 134Cs 2.10 137Cs 30.0 137mBa 147Pm 2.60 152Eu 13.3 154Eu 8.80 155Eu 4.96 210Pb 22.3 210Bi, 210Po




Daughters assumed to be in secular equilibrium

226Ra 1,600 222Rn, 218Po, 218At, 214Pb, 214Bi, 214Po, 210Tl, 210Pb

227Ac 21.7 227Th, 223Fr, 223Ra, 219Rn, 215Po, 211Pb, 211Bi, 207Tl

229Th 7,340 225Ra, 225Ac, 221Fr, 221Ra, 217Rn, 217At, 213Bi, 213Po, 209Tl, 209Pb

230Th 7.54E+04

232Th 1.40E+10 228Ra, 228Ac, 228Th, 224Ra, 220Rn, 216Po, 212Pb, 212Bi, 212Po, 208Tl

231Pa 3.27E+04 232U 68.9 233U 1.59E+05 234U 2.45E+05 235U 7.04E+08 231Th 236U 2.34E+07 238U 4.47E+09 234Th, 234mPa, 234Pa 237Np 2.14E+06 233Pa 238Pu 87.7 239Pu 2.41E+04 235mU 240Pu 6,540 241Pu 14.4 242Pu 3.76E+05 241Am 432 243Cm 29.1 244Cm 18.1

Table 2.5 List of radionuclides for the radiological assessment. Radionuclides with half-lives of <1year have been excluded, except where in secular equilibrium with parent radionuclides.

The radionuclide capacity of the site for each of the individual radionuclides in Table 2.5 can be determined, but the overall capacity of the site for LLW depends on the ratio between radionuclides in the waste streams consigned to the site. There is, therefore, no single radiological capacity for the site and, without knowledge of the consigned wastes, it is only possible to illustrate the radiological capacity by reference to particular waste streams.

For illustration, a decommissioning waste stream from UKAEA Harwell’s ongoing decommissioning programme has been used in this report. The radionuclide fingerprint of this waste stream is expected to be broadly applicable to other nuclear



industry decommissioning wastes that might be considered for disposal at the ENRMF.

The waste stream used to illustrate the potential radiological capacity of the ENRMF has been compiled from information on material in the Meashill Trenches at Harwell. This waste stream comprises a mixture of activated synchrotron components, reactor components and decommissioning/land remediation wastes, but is dominated by Co-60 through the presence of activated steel. Other waste streams from Harwell are more typically dominated by Cs-137. To partly reduce this dominance by Co-60, the 2000 inventory provided has been decayed to 2010 (Table 2.6).

The inventory presented in Table 2.6 does not represent the complete inventory for the Meashill Trenches. There will be small quantities of long-lived daughter radionuclides from the plutonium, uranium and thorium decay series. These become of increasing importance as the inventory decays, and are considered in the long-term assessments, but are not significant after only 10 years of decay. There are likely to be other radionuclides present, but these would contribute much less to any potential dose than the radionuclides listed.


2000 inventory 2010 inventory

MBq % MBq %

H3 12.3 5.70E+00 0.02 3.25E+00 0.03

Co60 5.27 3.00E+04 89.90 8.05E+03 72.25

Cs137 30 1.20E+03 3.60 9.52E+02 8.55

Ra226 1600 1.00E+02 0.30 9.96E+01 0.89

Th232 1.4E+10 4.00E+01 0.12 4.00E+01 0.36

U234 2.44E+05 5.00E+02 1.50 5.00E+02 4.49

U235 7.04E+08 2.40E+01 0.07 2.40E+01 0.22

U238 4.47E+09 5.00E+02 1.50 5.00E+02 4.49

Pu238 87.7 4.00E+01 0.12 3.70E+01 0.33

Pu239 2.41E+04 4.00E+02 1.20 4.00E+02 3.59

Pu240 6540 4.00E+02 1.20 4.00E+02 3.59

Pu241 14.4 6.18E+01 0.19 3.82E+01 0.34

Am241 432 1.00E+02 0.30 9.92E+01 0.89

Total 3.34E+04 1.11E+04

Table 2.6 Illustrative inventory for wastes from the Meashill Trenches, Harwell. The 2000 inventory provided by UKAEA Harwell has been decayed to 2010 to provide an input to radiological capacity calculations.



3 Assessment Methodology

This section describes the overall assessment methodology used to calculate potential doses from disposals of LLW at the ENRMF and to determine radiological capacities.

Section 3.1 summarises the SNIFFER methodology, on which the assessment is based. Section 3.2 describes the changes made to the SNIFFER methodology to take account of specific features of the ENRMF and the proposed disposals.

3.1 Summary of SNIFFER methodology

3.1.1 Assessment framework

The SNIFFER methodology was developed so as to provide the regulators, and other stakeholders, with a consistent approach to assessing the potential for landfill sites to accept the category of LLW known as Special Precautions Burial (SPB). The overall assessment approach is illustrated in Figure 3.1 (SNIFFER 2006a).

It was originally envisaged that a screening stage would be useful if large numbers of sites were examined (SNIFFER 2006a). This might be done by site owners, seeking to put forward a few sites as potential disposal sites, or by planners, seeking to assess the overall availability of disposal capacity for LLW. The principal application of the methodology, however, would be for the assessment of particular sites and this screening stage would not be required.

An important aim of the SNIFFER methodology was to provide regulators with a means of assessing radiological capacity for a landfill site and updating this capacity as more information becomes available and the available capacity is reduced though disposals. To ensure that the assessment is robust and fit for purpose, the approach developed by the IAEA and others of defining an assessment context forms an important part of the SNIFFER methodology. To provide as much consistency and flexibility as possible, the elements comprising the assessment context were incorporated as generic elements (providing consistency) or site-specific elements (providing flexibility) as considered most appropriate.

Using the assessment terminology established by the IAEA, the generic aspects of the assessment context are the assessment purpose, endpoints, basis, and assumptions regarding future society. The site-specific aspects of the assessment context are the environmental system of interest, site context, nature of the wastes, and assessment timescales.

The assessment endpoint and basis are established as generic factors so as to ensure as much consistency as possible between sites. The assessment end-point is dose, so that the results can be compared with an effective dose criterion. The SNIFFER methodology was based on a criterion of 20 μSv/year, representing the point at which doses arising from disposals can be regarded as being below regulatory concern (SNIFFER 2006a).



Figure 3.1: The overall SNIFFER assessment approach (SNIFFER 2006a).

ScreeningProtocol

DoseCalculations

AuthorisationConditions

DevelopAssessment

Context

AssessMitigationMeasures

RadiologicalCapacity

Calculations

GeneralSite

Information

Site-Specific

Data

GenericData

ExistingInventory

DoseConstraint

SiteUnacceptable

Fail

Pass

The assessment basis, also established as a generic element, includes all of the scenarios (describing ways in which doses could be received) that should be considered in an assessment. Some scenarios may be excluded from particular assessments, but only if there is a documented reason for doing so. Scenarios are discussed in more detail below.

The site context includes a range of features of the site and its surroundings that help to define the source-pathway-receptor system(s) used in the assessment calculations. These features include:



The identity and proximity of potentially affected populations or other environmental receptors.

Potential exposure pathways associated with the potentially affected populations, such as stream and groundwater discharge points, drinking water wells and irrigation practices, and atmospheric pathways for gas and dust, including point source emissions from combustion of landfill gas.

Site management practices, such as waste segregation, coverage of waste, liner type, permitted leachate head, and leachate management.

Past disposals of radioactive wastes and other wastes that might interact with radioactive wastes (e.g., organic materials).

3.1.2 Scenarios

As noted above, the selection of applicable scenarios is a site-specific aspect of the assessment context. As an aid to uniformity of approach, and to make possible the development of a useable assessment model, a set of potential scenarios is defined within the SNIFFER methodology (SNIFFER 2006a).

Scenarios are divided into operational and post-closure scenarios. Four exposed groups are considered.

Site workers. At the type of facility considered using the SNIFFER methodology, site workers are not considered as radiation workers, and may have no specific information about the types of material being consigned. In terms of dose constraints, therefore, they are considered in the same way as members of the public.

Members of the public living near the site.

Members of the public exploiting potentially contaminated groundwater or surface water resources. Depending on the hydrological setting of the site, this group may be the same as the local resident group.

Members of the public living on the site after closure and the withdrawal of controls.

Potential operational scenarios are presented in Table 3.1 and post-closure scenarios are presented in Table 3.2.



Scenario name Description Hazards

Gas Release

Liquid release (leachate)

Aerosols (leachate) Normal operations

Expected operation of the landfill up to capping and closure, as approved by the relevant Agency. Doses to site workers and to the public are considered.

Direct irradiation

Barrier failure

Failure of the artificial sealing liner and geological barrier during operations. Doses to the public are considered.


Leachate spillage

Unintentional release of leachate to surface water. Doses to the public are considered.


Solid release (dust while uncovered) Site remediation or

re-engineering

Workers expose waste during operations to remediate containment failure or to enlarge or otherwise re-engineer site.

Direct irradiation

Fire Fire releases radioactivity. Doses to site workers and to the public are considered.

Solid release (dust), gases and vapour

Table 3.1: Operational scenarios included in the SNIFFER methodology and the associated hazards (SNIFFER 2006a).

The last two of the scenarios in Table 3.1 are considered to encompass the range of other events that may result in a site worker being exposed, such as short-term contact with leachate.



Scenario name Description Hazards

Gas Release

Liquid release (leachate) Normal post-closure evolution

During this time, the landfill engineering is assumed to gradually degrade. Doses to the public are considered. Direct irradiation (through

cover)

Bathtubbing

Blockage of the drainage system causes overflow of leachate laterally from the landfill onto the soil. Doses to the public are considered.


Direct irradiation

Solid release (dust) Inadvertent excavation

Waste is inadvertently excavated and re-distributed, e.g., during building or farming. Doses to the intruder and the subsequent user of the site are considered.

Solid release (waste)

Table 3.2: Post-closure scenarios included in the SNIFFER methodology and the associated hazards (SNIFFER 2006a).

3.1.3 Dose calculations

This section describes the potential pathways identified within the SNIFFER methodology. Not all of these pathways will be necessarily be relevant to the assessment of a specific site, and the methodology requires both the identification and characterisation of the exposure pathways associated with a particular landfill. The pathways considered in the assessment of the ENRMF are discussed in Section 3.2 and Section 4.

External irradiation from standing near radioactively-contaminated waste. This pathway will be minimised when the waste is covered, and will then only apply to gamma-emitting wastes.

Inhalation of contaminated dust. Because the waste will be emplaced in sacks/drums and be buried on emplacement, creation of contaminated dust is not considered as an exposure pathway during the normal operation of the landfill. However, deliberate intervention to maintain, remediate or re-engineer the site (including the drilling of boreholes for landfill gas abstraction), or inadvertent excavation during unrelated development of the site after closure, could lead to the creation of contaminated dust.

Inhalation of aerosols from leachate. Leachate treatment potentially generates aerosols that could be inhaled by workers or members of the public near the site or any off-site treatment facility. The spraying of leachate back onto the



surface of the landfill is a practice that should be prevented through the Environmental Permitting process. Aerosols from leachate may, however, be generated during other types of leachate treatment either on or off-site, particularly if this involves aeration. Leachate treatment may continue after closure, but will end at the end of the control period. Use of leachate following the loss of control may also lead to aerosol formation but concentrations are likely to be lower than during leachate treatment.

Inhalation of dust, particles and gases from fires. Accidental fires in the waste are a potential hazard at landfill sites with combustible wastes. A fire at the site could lead to the release of radioactive particles and dust that could be inhaled by workers and members of the public downwind of the site, and could also lead to some gaseous releases. Waste fires may be associated with the collection and utilisation of landfill gas at sites which accept biodegradable wastes. Gaseous releases of radioactive material from flaring or other use are included in the following pathway.

Inhalation of radioactive gas, i.e., 14CO2, 14CH4,

3H, and radon. The first three may be generated through microbial degradation or corrosion of the radioactive waste. Landfill sites which accept biodegradable wastes are required to collect and flare or utilise the gas, and this could disperse radioactive gases that could be inhaled by workers and members of the public downwind of the site. Radon is generated through the decay of Ra-226, which in turn is a decay product of Th-230. Radon could be inhaled by workers, members of the public downwind of the site, and occupants working or living on the site after loss of control.

Ingestion of contaminated water. This pathway arises mainly through the leakage of leachate through the engineering and into groundwater (Figure 5). Once groundwater is contaminated, ingestion can occur through:

- extraction of contaminated groundwater via a well for drinking; and

- discharge of contaminated groundwater to surface water used for drinking.

Surface water may also be contaminated by the unintentional release of contaminated leachate. Once surface water is contaminated, ingestion can occur through:

- extraction of water for drinking.

Spillage of leachate may also contaminate groundwater used for drinking water supply, but retardation and dilution are likely to mean that potential doses through this pathway are less than those from surface water.

Ingestion of contaminated food. This pathway arises mainly through the leakage of leachate through the engineering and into groundwater. Once in the groundwater, radioactivity can contaminate food supplies through:



- extraction of groundwater for irrigation, thereby contaminating soil used for farming, or for stock watering;

- discharge of contaminated groundwater to surface water used for irrigation, thereby contaminating soil used for farming, or for stock watering; and

- discharge of contaminated groundwater to surface water or marine water that is used for fishing.

Surface water may also be contaminated by the unintentional release of contaminated leachate. Once surface water is contaminated, radioactivity can contaminate food supplies through:

- use of surface water for irrigation, thereby contaminating soil used for farming, or for stock watering;

- use of surface waters for fishing.

Spillage of leachate may contaminate groundwater used for irrigation, but retardation and dilution are likely to mean that potential doses through this pathway are less than those from use of contaminated surface water.

Soil may be contaminated by the lateral discharge of leachate directly from the site after blockage of the drainage system (bathtubbing).

Inhalation of dust from contaminated soil. This pathway mainly arises indirectly through the leakage of leachate through the engineering and into groundwater. Once in the groundwater, radioactivity can contaminate soil through:

- capillary rise of contaminated groundwater into the soil;

- discharge of contaminated groundwater to surface water and subsequent flooding;

- extraction of groundwater for irrigation, thereby contaminating soil; and

- discharge of contaminated groundwater to surface water used for irrigation, thereby contaminating soil.

Soil may also be contaminated indirectly through spillage or inadvertent discharge of leachate to surface water and subsequent irrigation.

Soil may be contaminated directly by the lateral discharge of leachate from the site after blockage of the drainage system (bathtubbing).

Details of the models and equations used to calculate doses via these pathways are given in the Technical Reference Manual for the SNIFFER methodology (SNIFFER 2006b). For the radiological assessment of the disposal of LLW with an activity of



up to 200Bq/g at the ENRMF, the models and assumptions underlying all of the scenarios described above have been re-examined and a number of changes made. These are described in Section 3.2 and Section 4.

3.2 Modifications to SNIFFER methodology

The assessment of a hazardous waste site for the disposal of significant volumes of LLW is sufficiently different to the original application of the SNIFFER methodology outlined above to require a re-examination of the key assumptions.

This re-examination identified several aspects of the overall methodology and the assessment model where different assumptions are required:

Dose criteria and compliance points

Barrier design and performance

Distribution of waste

Leachate concentration

Several aspects of the assessment model were also identified where the default parameter values included in the SNIFFER model required change to take account of site-specific features. These are highlighted in Section 4 and Appendices 1 and 2, where the assessment data are presented.

It should also be noted that there were errors in the dose coefficients included in the original SNIFFER assessment model, which did not account for the contribution to dose from short-lived daughter radionuclides in secular equilibrium with the parent radionuclides. These dose coefficients have been corrected and updated for the radiological assessment of the ENRMF.

3.2.1 Dose criteria and compliance points

In the original application of the SNIFFER methodology, a single dose criterion of 20 Sv/year was used as the basis for radiological capacity calculations and applied to all scenarios. More recently, the environment agencies have recognised that human intrusion into disposal facilities represents a different class of uncertainties about system behaviour than barrier degradation and natural processes. An alternative dose criterion for human intrusion has been specified in guidance for near-surface facilities intended solely for radioactive wastes (Environment Agency et al. 2009). This revised guidance states that the assessed effective dose to any person during and after an intrusion should not exceed a dose guidance level in the range of around 3 mSv/year to around 20 mSv/year. Values towards the lower end of this range are applicable to assessed exposures continuing over a period of years (prolonged exposures), while values towards the upper end of the range are applicable to assessed exposures that are only short term (transitory exposures).



The regulatory guidance notes that the following are events for which the dose guidance levels for human intrusion events apply:

human intrusion directly into a disposal facility;

other human actions that damage barriers or degrade their functions, such as removing material from a disposal facility cap. Barriers considered to be affected by these human actions may be engineered, natural or a combination of both.

The dose criteria used do not affect the way in which the dose assessments are conducted, but they are important for assessing the radiological capacity of a disposal facility. For the assessment of the ENRMF, three potential events are assessed that are considered to fall within these definitions:

Direct excavation of waste.

Occupation and subsequent use of the site following removal of the cap or excavation and re-distribution of waste.

Use of a borehole at the site boundary as a source of drinking water.

The first two of these events are derived from the inadvertent intrusion scenario in the SNIFFER methodology. Although doses to those excavating the waste could be regarded as transitory according to the regulatory guidance, and therefore subject to a dose guidance level of up to 20 mSv/year, the lower dose criterion of 3 mSv/year has been used in the calculation of radiological capacities.

The third event is included to provide a comparison with assessments of potential releases of non-radiological hazardous substances. Radiological assessments are based on calculating releases to the accessible environment and then determining doses to members of the critical group. For future releases, the same approach is used but a range of potentially exposed groups are considered at different release points where contaminated resources might be exploited in the future. In order to show compliance with the Groundwater Directive, assessments of potential releases of non-radiological hazardous substances must show that a site does not allow the discharge of List I substances into groundwater or the pollution of groundwater by List II substances. Such assessments therefore use compliance points at the water table, regardless of whether the groundwater is actually exploited at that point.

To provide a comparison between the two types of assessment, the radiological assessment has been extended to include use of a borehole at the site boundary for drinking water. This provides a compliance point for groundwater, although boreholes for drinking water would not normally be permitted in such locations as they would degrade the function of the natural barriers that are a key part of providing long-term safety for radioactive waste disposal. In calculating the radiological capacity of the site, it is therefore appropriate to regard such a borehole as an intrusion event and to use the dose guidance level of 3 mSv/year.



The radiological criteria in the GRA (Environment Agency et al. 2009) apply to a representative of the critical group or those at greatest risk. Although the criteria are expressed as annual doses, they are established on the basis that exposures may be prolonged (several years or lifetime exposures) rather than transitory (occurring in one specific year). Combined with the uncertainties inherent in these types of assessments, this means that the representative person is assumed to be an adult, and consumption rates and dose coefficients are set accordingly.

In the case of accidental releases, particularly during the operational phase, exposures may be for shorter periods than from post-closure, normal evolution releases. In these cases, it may be appropriate to use alternative assumptions about consumption rates and the corresponding dose coefficients to determine whether infants or children receive significantly greater doses than adults.

In the case of foetuses, the Health Protection Agency (HPA 2008) notes that for most radionuclides doses to the foetus are lower than to the mother, and that:

… for solid waste disposals it will generally be unnecessary to consider the ernbryo/fetus/breastfed infant as any increases in doses over those to other age groups will be small compared to the overall uncertainty in the assessed doses.

The radionuclides identified in this guidance as giving higher doses to the foetus than to the mother do not generally occur in decommissioning or similar wastes and are not included in the set of radionuclides considered in this assessment (Table 2.5).

3.2.2 Barrier design and performance

The principal differences between different types of landfill are the requirements relating to the barrier at the base of the landfill. Schedule 2 of the Landfill (England and Wales) Regulations 2002 states:

(4) The landfill base and sides shall consist of a mineral layer which provides protection of soil, groundwater and surface water at least equivalent to that resulting from the following permeability and thickness requirements -

(a) in a landfill for hazardous waste: k <= 1.0 × 10-9 metre/second: thickness ≥ 5 metres;

(b) in a landfill for non-hazardous waste: k <= 1.0 × 10-9 metre/second: thickness ≥ 1 metres;

(c) in a landfill for inert waste: k <= 1.0 × 10-7 metre/second: thickness ≥ 1 metres.

(5) Where the geological barrier does not meet the requirements of sub-paragraph (4) naturally, it may be completed artificially and reinforced by other means providing equivalent protection; but in any such case a geological barrier established by artificial means must be at least 0.5 metres thick.



The assessment model developed as part of the SNIFFER methodology was based on the disposal of LLW at non-hazardous waste sites (SNIFFER 2006a,b), and the default values for an effective geological barrier were set to a thickness of 1 m and a hydraulic conductivity of 1 x 10-9 m s-1. The assumption was that this barrier would be provided by an artificial mineral layer between a basal liner and the natural geological barrier. The presence of a natural geological barrier (unsaturated zone) is allowed for in the assessment model, but typical characteristics of this zone mean that it is likely to be less resistant to leachate transport than the artificial geological barrier and effectively redundant in terms of assessing potential doses via the groundwater pathway.

In the case of the landfill cap, the default values used in the SNIFFER methodology assume that the cap is initially 95% efficient in terms of preventing infiltration into the landfill, and that the cap will gradually degrade over a period of 60 years from the time of emplacement until it is no more effective than a soil layer. Also, the SNIFFER methodology assumes that it is only the effectiveness of the cap that limits infiltration of leachate into groundwater after the end of the operational period – the liner is assumed to become ineffective at the time the cap is emplaced.

At the ENRMF, the basal liner in the area considered for disposal of radioactive wastes is constructed with a 2 mm thick high density polyethylene (HDPE) geomembrane, and at least 1.5m thickness of artificially emplaced geological barrier (Upper Lias clay sourced locally). This clay is placed with a maximum design permeability of 3x10-10 m/s.

The final cap for the ENRMF comprises a gas drainage layer, clay regulating layer, geotextile protector, geosynthetic clay liner, LDPE geomembrane liner and soil cover. The cap design will aim for a minimum effectiveness of 99%.

Following capping the assessment model assumes that there will be a minimum of 60 years management of the site, which will include monitoring of leachate levels within the waste. In practice the management period will be considerably longer. This will enable the effectiveness of the cap and the bottom liner to be assessed and for mitigation measures to be taken if there is evidence of damage or deterioration. It is therefore reasonable to assume that the design performance of the cap and liner will be maintained during the management phase and that degradation will not take place until after the withdrawal of control.

3.2.3 Distribution of waste

The SNIFFER methodology does not require the actual volume of waste to be specified, because the dose calculations are based on the disposal of 1 MBq of each radionuclide. However, to determine the concentration of waste that might be excavated after site closure, the methodology assumes that all of the disposals at a particular site could be in part of a cell as small as 10 m3.

The proposed disposals of LLW at the ENRMF could form a significant proportion of the material disposed of to the selected cell. The SNIFFER methodology has



therefore been revised to account for waste deposited throughout the cell rather than in a particular volume. For the radiological assessment, it is again not necessary to assume the volume of waste, because the calculations are based on a unit disposal of 1 MBq, homogeneously dispersed throughout the cell. Assumptions regarding waste activity are required to convert the calculated radiological capacity into waste volumes and to determine whether the assumption concerning homogeneity is reasonable.

3.2.4 Leachate concentration

There are significant uncertainties associated with modelling the release of radionuclides from radioactive waste and into leachate. The mechanisms by which this release would occur depend on the type of waste (e.g., waste composition, how it is contaminated and how it is packaged), on the conditions within the landfill (e.g., pH, Eh, degree of saturation) and on the radionuclides concerned (e.g., whether they are readily sorbed). Even with detailed mechanistic models of waste behaviour, significant variability (due to heterogeneities in the wastes and landfill conditions) and uncertainties (due to lack of information about the processes involved) would remain.

In the SNIFFER methodology, these uncertainties are treated by means of conservative assumptions:

For scenarios involving waste excavation, it is assumed that the entire radionuclide inventory remains in the solid waste and that there are no losses to leachate.

For scenarios involving leakage of leachate, it is assumed that the entire radionuclide inventory is available for dissolution into leachate at site closure, with the concentration in leachate determined by the appropriate sorption coefficient (Kd).

The assumption regarding the partitioning of radionuclides between waste and leachate would be conservative even if sorption coefficients could be determined for the actual wastes and conditions within the landfill, because not all of the radioactive contamination would be on the surface of the waste and available for immediate dissolution. Furthermore, because of the difficulties in determining sorption coefficients, the default values are set to zero, effectively meaning that in the SNIFFER methodology the entire radionuclide inventory enters the leachate at site closure.

For the radiological assessment of the ENRMF, alternative assumptions have been made for the scenarios involving leakage of leachate:

For pathways involving contamination of soil (including irrigation using contaminated groundwater), the assumption that the entire radionuclide inventory is available for dissolution into leachate at site closure is retained.



For pathways involving drinking water or other transitory exposures such as aerosols, it is assumed that dissolution is gradual and an annual release to leachate is used to calculate releases and doses.

These pathways are distinguished because in the former case radionuclides will accumulate in the soil and an assessment effectively based on a single year would not be demonstrably conservative.

The closest analogue for landfill disposal is the trench disposals at the LLWR near Drigg. A comparison of the annual discharges through the marine pipeline (BNFL 2002a) with estimates of the disposed inventory (BNFL 2002b) indicates that a factor of at least 1x10-3 year-1 should be applied to determining what fraction of the inventory might be in leachate. Initial concentrations, and concentrations of more insoluble radioelements, would probably be lower than this, but this factor has been used in this assessment for all radionuclides as a conservative assumption.

3.3 Supplementary calculations

In addition to a radiological assessment based on the pathways and scenarios included within the SNIFFER methodology, two supplementary calculations have been undertaken. These relate to potential doses from the treatment and discharge of leachate at an off-site water treatment plant, and to possible radiological effects on wildlife.

The SNIFFER methodology includes a leachate spillage scenario, which is modelled as a release of leachate to a water body (e.g., river, lake) that is then exploited as a water resource (e.g., drinking, fishing). This scenario and modelling treatment is intended to address accidental releases and not routine discharges. At the ENRMF, leachate is collected and sent by tanker to a water treatment plant at Avonmouth. Following treatment, water is then discharged to the Severn Estuary. Potential doses arising from the leachate treatment and discharge are not included within the SNIFFER methodology and have been separately assessed using the Environment Agency’s Initial Radiological Assessment - Sewer methodology (Environment Agency 2006a; 2006b).

Discharges and migration of radionuclides from a disposal facility might have a detrimental effect on non-human species or more general environmental effects such as damaging habitat quality. The guidance from the Environment Agencies includes a requirement to ensure that all aspects of the accessible environment are protected:

The developer/operator should carry out an assessment to investigate the radiological effects of a disposal facility on the accessible environment both during the period of authorisation and afterwards with a view to showing that all aspects of the accessible environment are adequately protected.

Although there is no specific evidence that there might be a threat to populations of non-human species from the authorised release of radioactive substances if people are protected, environmental damage might occur to areas and habitats that are not extensively exploited by people. Furthermore, there is a specific need to be able to



demonstrate that non-human species are protected under legislation related to conservation, for example that derived from the EC Habitats Directive (EC 1992).

There are currently (June 2009) no internationally established criteria for determining radiological protection of the environment. However, a number of research studies and regulatory guidance documents have proposed that an incremental dose rate value of 10 Gyh-1 is appropriate as a screening criterion, although dose rates less than 40 Gyh-1 are unlikely to exert any effect on the reproductive capacity of mammals and chronic effects for other organisms are unlikely at even greater dose rates (Copplestone et al. 2002).

An assessment tool developed as part of the ERICA project (Environmental Risk from Ionising Contaminants: Assessment and Management) has been used to calculate potential dose rate values. The ERICA assessment tool allows three tiers of assessment. A Tier 1 assessment has been undertaken and the calculated incremental dose rate values are below the screening value indicated above. More detailed assessments (Tier 2 and Tier 3) are therefore not required.



4 Assessment Data and Assumptions

The assessment model is a simplified model of the events and processes that will or might take place during and after operations. A number of simplifying assumptions are therefore required in order to represent the site and its surroundings, as summarised in Section 2 of this report, in the model. These assumptions are outlined in this Section. Where there are significant uncertainties regarding aspects of the site, alternative sets of assumptions have been made and assessment calculations carried out.

This section includes the equations used in modelling the release of radionuclides from the site according to the different scenarios and assumptions. Equations for calculating the potential doses from these releases are presented in Appendix A. Parameter values used in the calculations are included in this section and in Appendices A and B (radionuclide-specific data). Unless site-specific parameter values have been identified, the parameter values used in the modelling are those used in SNIFFER (2006b), which are derived in large part from IAEA (2003).

4.1 Site characteristics

4.1.1 Size of site

The ENRMF landfill site has an overall disposal volume of 1,800,000 m3, with a surface area of about 125,000 m2. About 700,000 m3 remains available for disposal, with a surface area of about 50,000 m2.

For the purpose of the radiological assessment reported here, it is assumed that disposal of LLW will be restricted to Cells 4B, 5A and 5B with a volume of 497,534 m3 and a surface area of 34,108 m2.

4.1.2 Construction

It is assumed that Cells 4B, 5A and 5B are hydrologically isolated from the remainder of site.

4.1.3 Barrier

All cells include a 2 mm thick high density polyethylene (HDPE) geomembrane. Seepage through a geomembrane sealing layer is dominated by leaks through flaws (holes) in the liner. The number of holes will depend on the effectiveness of the quality control during emplacement, but some holes will occur in all cases. Large holes will generally be detected, and so smaller holes or pinholes will be most common. For a geomembrane liner underlain by a mineral layer or host geology, the flow, qliner (m

3 year-1), through holes in the liner is given by:

07E16.374.09.01.0 barrierholesliner Khacq



where c is a constant depending on the contact between the liner and the material beneath (0.21 for good contact, 1.15 for poor contact) (dimensionless).

aholes is the area of the holes (m2). h is the head of leachate (m). Kbarrier is the hydraulic conductivity of the barrier (material beneath the

liner) (m s-1). 3.16E+07 is the number of seconds in a year (s year–1).

The parameter values assumed for the assessment are:

c 0.5 aholes 4.2 x 10-4 m2 h 1 m Kbarrier 3 x 10-10 m s-1 Other assumed characteristics of the engineered clay barrier are presented in Table 4.1.

Parameter Value Rationale Reference

Thickness 1.5 m Representative value based on design parameter

Clay liner thickness is in the range 0.5 to 2.5 m (≥ 1.5 m for future

cells) – HRA, pp 62-70

Hydraulic conductivity 3×10-10 m/s Mean value Minimum (1×10-11 m/s), mean (3×10-10 m/s) and maximum (6.6×10-10 m/s) permeability

values for 115 clay samples from Cells 1A, 1B and 3A – HRA, p 107

Porosity 0.05 Mean value Minimum (0.01), mean (0.05) and maximum (0.1) estimates based on

specific yield values for clay – HRA, p 107

Density 1560 kg/m3 Calculated mean value Minimum (1260 kg/m3) and maximum (1860 kg/m3) values for 445 clay samples from Cells 1A,

1B and 3A – HRA, p 107

Table 4.1 Properties of artificial clay barrier.

The geological barrier and unsaturated zone (Section 4.2 below) are treated as a single unit of thickness, D (m), and the advective transfer of radionuclide Rn through the unit, λbarrier (year-1), is given by:

Rnbarrierdbarrierbarrierlandfill

barrierRnbarrier KaD

q

,

where qbarrier is the volume of the water flowing through the barrier (m3 year-1).



φbarrier is the porosity of the barrier (dimensionless). ε is the degree of saturation of the barrier (dimensionless). Kd,barrier is the distribution coefficient for radionuclide Rn in the barrier

(m3 kg-1). ρbarrier is the bulk density of the barrier (kg m-3). alandfill is the area of the landfill (m2). D is the depth of the barrier (m). The maximum value of qbarrier is determined by the product of the area of the landfill, alandfill (m

2) and the hydraulic conductivity of the unit, Kbarrier (m year-1). The actual value of qbarrier varies over the assessment period:

For the period of operation of the landfill, qbarrier is set to qliner.

After emplacement of the cap and while the site is still managed and monitored, qbarrier is set to the minimum of qliner and infiltration through the intact cap (see below).

After the management phase, and before complete degradation of the cap, qbarrier is set to the minimum of the infiltration through the degraded cap (see below) and the value determined by the barrier properties.

After degradation of the cap, qbarrier is determined by the barrier properties.

The release of radioactivity over time into the geological barrier and radioactive decay result in a change to the inventory remaining in the landfill:

Rnwastedwastewastelandfill

barrierRnwaste KV

q

,

t

initialRnRn

RnwasteRneAtA ,)(

where Rnwaste is the rate constant for radionuclide Rn from loss of leachate (year-1).

qbarrier is the volume of the water flowing through the geological barrier (m3 year-1).

t is the time (years). λRn is the radioactive decay constant of radionuclide Rn (year-1). ARn,initial is the initial inventory of each radionuclide (Bq).

4.1.4 Cap

The assessment model does not require explicit details of cap construction. The volume of water available to infiltrate the landfill is assumed to be a function of the annual precipitation and the efficiency of the cap in diverting this precipitation.

landfilleff aPq inf

)1( 0ErunoffAEPP totaleff for tc < t ≤ te



ef

etotaleff tt

ttErunoffAEPP 11 0 for te < t ≤ tf

runoffAEPP totaleff for t > tf

where qinf is the volume of water entering the landfill through the cap (m3 year-1).

alandfill is the area of the landfill (m2). Peff is the potential rate of water infiltration through the cap of the

landfill (m year-1). Ptotal is the total precipitation (m year-1). AE is the amount of precipitation that is lost by evapotranspiration

(m year-1). runoff is the amount of precipitation lost by runoff (m year-1). E0 is the initial cap efficiency (a dimensionless fraction of the

infiltration water initially deflected by the cap). t is the time after closure (years). tc is the time cap is emplaced (years). te is the time cap starts to degrade (years). tf is the time of cap failure (years).

The HRA (p 31) provides a table of average monthly effective rainfall (rainfall minus potential evapotranspiration) for the period 1961 to 1990, indicating a mean annual value of ~ 0.072 m. In comparison with the annual rainfall for the area (~ 0.6 m/year), this value is low for effective rainfall but is considered appropriate for net infiltration (effective rainfall minus runoff).

Based on the assumptions for infiltration made in the HRA, the initial cap efficiency is set at 99%. The period of cap effectiveness is assumed to be 60 years and the period of cap degradation is assumed to be 100 years.

In addition to providing protection to the landfill against infiltration, the cap also reduces potential doses from external radiation to members of the public living and working on the cap after closure. For these calculations, the cap is assumed to have a minimum thickness of 1.5 m.

4.1.5 Operational period

For the purpose of the assessment, it is assumed that the facility will continue in operation for a further 5 years, and that LLW disposals will take place throughout this period.

It is proposed that waste will be transported to the site in suitable transport packages and disposed directly, and that loose waste will not be handled at the site. It is also proposed that waste will be covered by 0.3 m of soil material. These practices will ensure that members of the public off-site will not be exposed to groundshine. On-



site workers, using mechanical handling to emplace the waste packages and soil cover or working near-by, may be exposed. Occupational dose assessments will be made separately from this report.

4.1.6 Leachate collection and management procedures

Leachate levels at the ENRMF are maintained by pumping excess leachate to tankers and transporting this leachate to a water treatment plant at Avonmouth. There may be periods when this route is unavailable – leachate could be re-circulated to the upper parts of the site but leachate would not be used for dust suppression or other processes that could lead to aerosol formation at the site.

The radiological assessment includes two scenarios that could result in doses to off-site exposed groups:

Tanker accident resulting in spillage of leachate and contamination of a water resource.

Routine treatment of leachate and discharge of treated water to an estuary.

The first of these scenarios has been considered using the SNIFFER methodology as discussed below. The second of the leachate scenarios has been considered by means of supplementary calculations described in Section 4.4.

4.1.7 Leachate spillage

Notwithstanding any radioactive components, landfill leachate poses a hazard to the environment if spilt and any road accident involving loss of an entire load would be subject to mitigation measures. Leachate that did enter water resources would also become diluted. For this assessment, it is conservatively assumed that an entire tanker load of leachate (30 m3 of leachate) reaches a small reservoir (2 x 106 m3) that is used for drinking water, irrigation and fishing.

The dissolved radionuclide concentration, CRn,leachate (Bq m-3) in the leachate associated with an inventory ARn (Bq), is given by:

wastelandfill

RnleachateRn V

DfAC ,

where Vlandfill is the volume of the waste (m3). Df is the dissolution factor (-). φwaste is the porosity of the waste (dimensionless). ε is the degree of saturation of the waste (dimensionless).

Parameter values used in the calculation of radionuclide concentrations in leachate are listed in Table 4.2.



Parameter Description Value Units Vlandfill volume of the waste 672,098 m3 Df dissolution factor 0.001 - φwaste porosity of the waste 0.5 - ε degree of saturation of the waste 0.5 -

Table 4.2 Parameter values used in calculating leachate concentrations.

The contamination is assumed to relate to a one-off event, but the resulting radioactive contamination, CRn,water,spill (Bq m-3), is assumed to remain constant for one year (i.e., no dilution by throughflow):

water

spillleachateRnspillwaterRn V

VtCC

)(,,,

where CRn,leachate(t) is the concentration of radionuclide in the leachate at the time of the spill, t (Bq m-3).

Vspill is the volume of leachate in the spill (m3). Vwater is the volume of the surface water body (m3). Dose calculations for drinking contaminated water, or ingesting fish taken from contaminated water are described in Appendix A.

If the contaminated water body is used for irrigation, then a one-off soil concentration, CRn,soil,spill (Bq kg-1), is calculated from:

soilsoil

ratespillwaterRnspillsoilRn d

IrrigCC

,,,,

where Irrigrate is the amount of irrigation in one year (m). dsoil is the depth of the soil layer being irrigated (m). ρsoil is the density of the soil (kg m-3). Dose calculations for ingestion of crops grown on irrigated soil and ingestion of contaminated soil are described in Appendix A. Decay constants and other radionuclide-specific parameter values are presented in Appendix B. Other parameter values used in the calculations of specific doses for the ENRMF are listed in Table 4.3.

Parameter Description Value Units Vspill volume of leachate in the spill 30 m3 Vwater volume of the surface water body 2 x 106 m3 Irrigrate amount of irrigation in one year 0.3 m dsoil depth of the soil layer being irrigated 1 m ρsoil density of the soil 1300 kg m-3

Table 4.3 Parameter values used in the calculation of the effects of leachate spillage.



There is potential, during leachate management or spillage, for the production of aerosols which could lead to doses via the inhalation pathway. Other pathways, such as external irradiation from deposited aerosols or ingestion of foodstuffs contaminated by aerosols would give specific doses that are comparable to or less than the inhalation pathway.

The concentration of aerosols at time t, CRn,air,aero(t) (Bq m-3), created during leachate management or spillage is assumed to be equivalent to the concentration of the leachate diluted by the aerosol load:

)(1000

)( ,,, tCaerosol

tC leachateRnaeroairRn

where aerosol is the aerosol concentration (kg m-3 of air). CRn,leachate is the activity of radionuclide, Rn, in the leachate at time t (Bq m-3). 1000 is the density of water (kg m-3).

The above equation cautiously assumes that the aerosols are non-depleting during passage towards the exposed individual. An aerosol concentration of 0.001 kg m-3 of air is assumed.

Exposure to aerosols at the ENRMF or during a tanker accident will be abnormal and short-lived. An initial assessment of the potential impacts from routine, off-site leachate management has been made using the Environment Agency’s methodology and the assumption that doses from water treatment would be similar to doses from sewage treatment.

The Environment Agency’s methodology allows for a range of exposure groups affected by releases to a public sewer, depending on the discharge route for treated effluent. For this assessment, only the groups associated directly with operation of the treatment plant, farming of land conditioned by sludge or using the estuary are considered. These groups and the relevant exposure pathways are:

Sewage treatment workers (adults only)

External irradiation from radionuclides in raw sewage and sludge

Inadvertent inhalation and ingestion of raw sewage and sludge containing radionuclides

Farming family living on land conditioned with sewage sludge

Consumption of food produced on land conditioned with sludge and incorporating radionuclides

External irradiation from radionuclides in sludge conditioned soil

Inadvertent inhalation and ingestion of sludge conditioned soil

Fisherman family (estuary/coastal water receives treated effluent from sewage works, typically via a river)



External irradiation from radionuclides deposited in sediments

Consumption of fish incorporating radionuclides

A key assumption in assessing potential doses from off-site leachate management is the extent of dilution with other inputs to the water treatment plant. For this assessment, it is assumed that the Avonmouth facility treats some 1,080 m3 per day, based on the use of six anaerobic digesters, with a daily feed of about 180 m3 per day each.

4.1.8 Control over future site use

It is intended that the future use of the ENRMF site, after closure, would be for agriculture, and that normal agricultural practices, combined with knowledge of the previous site use, would prevent intrusion into the waste or the excavation of radioactive material. However, for the purpose of the assessment, it is assumed that knowledge of the site will be lost and there will be no control over use of the site at some time after closure.

Loss of control means that there is a potential for the site to be disturbed and for radioactive material to be incorporated into soil used to grow crops and graze animals. The calculations of effective doses to workers engaged in excavation activities and to members of the public residing on the disturbed facility are described in Appendix A. The principal assumption is that control over the site will be lost 60 years after closure. To illustrate the sensitivity to this assumption, alternative cases in which it is assumed that control over the site will be lost 20 years and 100 years after closure have also been considered.

4.2 Hydrogeological setting

Hydrogeology data were derived from the HRA (ESI, 2004) and the Environmental Statement (Bullen Consultants Ltd, 2005) prepared for assessments of hazardous waste disposal at the ENRMF.

4.2.1 Underlying geology

ENRMF landfill site was originally a clay pit, which has been largely quarried out, exposing the top of the underlying Lincolnshire Limestone, the thickness of which is between 15 m to 20 m. The upper part of the formation comprises a sequence of oolitic limestones, whereas the lower part is composed of fine grained sandy limestones. The Lincolnshire Limestone has been classified as a Major Aquifer by the Environment Agency. It is characterised by fracture flow and there are also swallow holes in the vicinity which allow rapid flow of water from the surface to the water table.



4.2.2 Unsaturated zone characteristics

Water levels records from monitoring boreholes at the site indicate fluctuation in groundwater levels in the range of approximately 3 m to 7 m below the base of Cells 1 and 2 (HRA, p. 40). Although, given the uncertainties in the dataset, it is possible for the unsaturated zone to be no more than 3 m thick across the entire site, it is considered that assuming this minimum thickness would be overly conservative. For the purpose of the assessment the mean value, 5.5 m, of the range of values reported has been used.

The unsaturated zone is modelled together with the geological barrier as a single unit (see Section 4.1). The approach adopted ignores any dispersion effects in the unsaturated zone.

4.2.3 Saturated zone characteristics

The thickness of the saturated zone across the base of the landfill site can be derived from the estimates for the thickness of the Lincolnshire Limestone and that of its unsaturated zone. As mentioned above, the latter is in the range of 3-7 m and the former in the range of 15-20 m, suggesting a thickness for the saturated zone in the range of 7-17 m, with a mean value of 12 m. (Note that the Environmental Statement suggests a thickness for this layer of approximately 7-18 m.)

There is uncertainty on the range of hydraulic conductivity values for the Lower Lincolnshire Limestone formation (saturated zone) underneath the landfill site. Values reported in 1998 suggest a hydraulic conductivity of 0.01 to 0.1 m/day (Environmental Statement, p. 44), whereas the results of the most recent slug tests indicate hydraulic conductivity in the range of 1 to 7 m/day (HRA, p. 38). The assessment uses a mean value derived from the more recent findings.

The permeability and transmissivity of the Lincolnshire Limestone are mainly due to fractures and, as such, the fracture porosity of this formation is used as the effective porosity in the assessment.

Parameter values assumed for the saturated zone are presented in Table 4.4.




Thickness 12 m Calculated mean value The Lincolnshire Limestone formation is 15-20 thick, with an

unsaturated zone thickness of 3-7 m – HRA, pp 34 and 40

Hydraulic conductivity 4.63×10-5 m/s Calculated mean value Minimum (1 m/day) and maximum (7 m/day) values derived from slug

tests – HRA, p 38

Hydraulic gradient 0.0025 Calculated mean value 0.002 - 0.003 – HRA, p 41

Porosity 0.007 Calculated mean value 0.004 – 0.01 – HRA, p 107

Density 2000 kg/m3 Estimated density for sedimentary rock

Estimated density for sedimentary rock – HRA, p 108

Table 4.4 Properties of saturated zone.

4.2.4 Groundwater discharges

Local groundwater level contours at the landfill site indicate that groundwater flows approximately southwards to south-eastwards (HRA, p. 41). Hence, for assessment purposes, only groundwater abstractions south to south-easterly of the landfill site are considered. The nearest such active licensed groundwater abstraction point is at Law’s Lawn, 1487 m south-east of the site – the water is abstracted from the confined aquifer under artesian conditions and is used solely for agricultural purposes (Environmental Statement, p 48) - Table 4.5.


Distance to nearest groundwater abstraction

point

1487 m Exact value Groundwater is abstracted at Law's Lawn (1,487 m south-east of the

site) for agricultural usage

Irrigation Assumed sub-activity under agricultural usage

As above

Abstracted water usage Livestock Assumed sub-activity under

agricultural usage As above

Table 4.5 Groundwater abstraction.

4.2.5 Stream and river characteristics

There are no natural surface water features on, and no known springs in the vicinity of, the site. The nearest spring is about 1 km south-east of the site, but no further data are available. The nearest natural surface water course is the River Welland, approximately 2.5 km to the west of the site – only data on water quality are available



for this stream. Hence, for the assessment, it is assumed that groundwater does not discharge to a stream or river, but is instead abstracted from a borehole (see above).

4.2.6 Groundwater flow and radionuclide transport

The groundwater flow and radionuclide transport model is based on a series of units or compartments, with the transfer of radionuclide Rn from each unit to the next downstream, λgw (year-1), being given by:

Rnrockdgwgw

gwRngw KL

HK

,

where Kgw is the hydraulic conductivity of the rock in which the groundwater flow is occurring (m year-1).

ΔH is the hydraulic gradient (dimensionless). φgw is the porosity of the groundwater pathway (dimensionless). Kd,rock is the distribution coefficient for radionuclide Rn in the rock

(m3 kg-1). ρgw is the bulk density of the groundwater pathway (kg m-3). L is the length of each groundwater compartment (m). Longitudinal dispersion is approximated implicitly by dividing the path length into ten units. Transverse dispersion is approximated by successively increasing the width (and, thereby, the volume) of each downstream unit to account for spreading of the plume of contaminated groundwater. The width, W (m), at a distance, Δx (m), downstream is given by:

xWW T 2422

0

where Wo is the initial width of the unit in which the groundwater flow is occurring (m).

Δx is the distance downstream (m). αT is the transverse dispersion length (m), assumed to be one tenth of the

initial width. For the calculation of radionuclide concentrations in a borehole at the site boundary, the overall groundwater path length is assumed to be 100 m, representing flow from a point below the centre of the site to the site boundary.

The concentration of a radionuclide in water abstracted from groundwater at time t, CRn,water(t) (Bq m-3), is given by:

Rnrockdgwgwgw

gwRnwaterRn KV

tAtC

,

,,

)()(

where ARn,gw(t) is the activity in the groundwater compartment at time t (Bq). Vgw is the volume of the groundwater compartment (m3). φgw is the porosity of the groundwater pathway (dimensionless).



Kd is the distribution coefficient for radionuclide Rn in the rock (m3 kg-1).

ρgw is the bulk density of the rock for the groundwater pathway (kg m-3).

The change in concentration of radionuclides in soil, CRn,soil (Bq kg-1), that is irrigated with contaminated water is given by:

)()( ,,, tC

d

IrrigtC

dt

dCsoilRneff

soilsoil

ratewaterRn

soilRn

where CRn,water(t) is the concentration of radionuclide in the water used for irrigation at time t (Bq m-3).

Irrigrate is the rate of irrigation (m year-1). dsoil is the depth of the soil layer being irrigated (m). ρsoil is the density of the soil (kg m-3). λeff is an effective decay coefficient that considers radioactive decay,

leaching from the soil, uptake by plants, and erosion (year-1), given by:

erosionsoilsoil

plantRn

plant

Rnsoildsoilsoilsoil

totalRneff d

YieldTF

Kd

runoffAEP

,

where Ptotal is the total precipitation (m year-1). AE is the amount of precipitation that is lost by evapotranspiration

(m year-1). runoff is the amount of precipitation lost by runoff (m year-1). Yieldplant is the plant yield (kg m-2 year-1). TFplant is the soil to plant transfer factor for radionuclide, Rn (Bq kg-1 fresh

weight of crop per Bq kg-1 of soil). dsoil is the depth of the soil layer being irrigated (m). ρsoil is the bulk density of the soil (kg m-3). φsoil is the porosity of the soil (dimensionless). ε is the degree of saturation (dimensionless). Kd,soil is the distribution coefficient for radionuclide Rn in the soil (m3 kg-1). λRn is the decay constant of radionuclide Rn (year-1). λerosion is the loss of radioactivity owing to erosion of the soil (year-1).

The removal of activity from the groundwater through the irrigation process is not tracked.

Dose calculations for the groundwater pathway are described in Appendix A. Decay constants and other radionuclide-specific parameter values are presented in Appendix B. Other parameter values used in the calculations of specific doses for the ENRMF are listed in Table 4.6.



Parameter Description Value Unit Irrigrate rate of irrigation 0.3 m year-1 dsoil depth of the soil layer being irrigated 1.0 m ρsoil density of the soil 1300 kg m-3

Pasture 1.7 Grain 0.4 Green veg 3.0

Yieldplant plant yield

Root veg 3.5

kg m-2 year-1

φsoil porosity of the soil 0.3 dimensionless ε degree of saturation 0.5 dimensionless λerosion loss of radioactivity owing to erosion

of the soil 2.0x10-4 year-1

Table 4.6 Parameter values used in the calculation of radionuclide concentrations in groundwater used for irrigation and in irrigated soil.

4.3 Other scenarios and pathways

4.3.1 Gas

The gas pathway has been considered only for tritium and radon. It is not envisaged that there would be sufficient organic waste material in the LLW to generate radiogenic CO2 or CH4. Potentially exposed groups for the gas pathway are site workers, members of the public spending time immediately downwind of the site during the operational period, and members of the public living in a house built on the site after closure.

Radioactive Gas Release

For H-3 (in hydrogen, water, or methane) and C-14 (in carbon dioxide or methane), the release rate of radioactive gas, RRn,gas (Bq year-1), at time t (years) is given by:

gas

gast

wasteRngasRn

feAtR

Rn

,

, )(

where: ARn,waste is the initial activity of radionuclide Rn in the waste (Bq). λRn is the decay constant of radionuclide Rn (year-1). fgas is the fraction of the activity associated with each gas

(dimensionless). τgas is the average timescale of generation of each gas (years).

For radon (Rn-222), the release rate at time t is given by:

2

2

226

-h

1wastewaste226,-Ra222 eH)( HtRnradon

RaeCatR

where: λ is the decay constant of the indicated radionuclide(year–1). a is the surface area of the disposal unit (m2). CRa-226,waste is the initial Ra-226 concentration in the waste (Bq kg–1).



ρwaste is the bulk density of the waste (kg m-3). τ is the emanation factor, defined as the fraction of the radon atoms

produced which escape from the solid phase of the waste into the pore spaces (dimensionless).

H1 is the effective diffusion relaxation length for the waste (m). h2 is the thickness of the cover (m). H2 is the effective relaxation length of the cover (m).

Dose calculations for the gas pathway are described in Appendix A. Decay constants and other radionuclide-specific parameter values are presented in Appendix B. Other parameter values used in the calculations of specific doses for the ENRMF landfill are listed in Table 4.7.

Parameter Description Value Units fgas fraction of the activity associated with

tritium 3.9x10-2 dimensionless

τgas average timescale of generation of tritium

50 years

a surface area of the disposal unit 34,108 m2 ρwaste bulk density of the waste 700 kg m-3 τ emanation factor, defined as the

fraction of the radon atoms produced which escape from the solid phase of the waste into the pore spaces

0.1 dimensionless

H1 effective diffusion relaxation length for the waste

0.2 m

h2 thickness of the cover 1.5 m H2 effective relaxation length of the cover 0.2 m

Table 4.7 Parameter values used in the calculation of gas release rates during operations and after closure.

4.3.2 Fire

Fire is a potential issue at landfill sites where LLW is disposed of alongside municipal and other wastes with large amounts of combustible material. It is not envisaged that there would be significant amounts of combustible material amongst the LLW or the hazardous waste, and fires within existing disposal cells would not affect the cells containing LLW. The consequences of a fire starting within or affecting the LLW have therefore not been assessed. There is a potential for accidents such as aircraft impact to release material in a similar manner to a fire, but the scale and non-radiological consequences of this type of accident means that they are more appropriately discussed in qualitative terms in the overall safety case rather than modelled within the radiological assessment.

4.3.3 Barrier failure

This scenario was included in the SNIFFER methodology to account for the possibility of damage or defects in the lining and a damaged or inadequate geological barrier could lead to leachate release during operations. This is a conservative



scenario even for a non-hazardous waste site with LLW disposals. It is considered unreasonable for a hazardous waste site receiving LLW where the construction, operation and monitoring will all reduce the possibility of the barrier failing in a manner that allows the release of large amounts of leachate. Even if damage did occur, the potential for environmental damage from leachate from such a site would ensure that remediation would occur before members of the public were exposed. The barrier failure scenario has therefore not been assessed.

4.3.4 Site remediation and re-engineering

This scenario was included in the SNIFFER methodology because it was possible that a site operator would have no records of radioactive waste disposals or their location. In the case of comparatively large volumes of LLW disposed of to a hazardous waste landfill, records would be maintained. Any remediation work would be done with the knowledge that there was radioactive material on the site and it can be assumed that appropriate precautions against exposure would be adopted.

4.3.5 Bathtubbing

This scenario was included in the SNIFFER methodology to account for the possibility of excessive infiltration through the cap at a time when the barrier still prevents leakage to the underlying formation. For a hazardous waste site, it is envisaged that controls on cap construction and leachate monitoring would prevent or identify releases through this pathway. For the purpose of the assessment, it has been assumed that remediation, cognisant of radioactive material, would occur before members of the public were exposed via this pathway.



5 Dose Calculations

This section presents results from the calculations of specific dose arising from the different scenarios and pathways described in Section 4 and the dose calculations described in Appendix A.

Radiological capacity calculations for the ENRMF, based on these dose calculations, are presented in Section 6.

5.1 Groundwater pathway

Specific doses calculated for members of the public via the groundwater pathway are presented in Table 5.1. These calculations are based on assumptions described in Sections 4.1 and 4.2 and in Appendix A (Section A.4). Sensitivity studies showing the effect of variations in leachate head, cap lifetime, cap efficiency, the length of the assessment period and the exposed individual on the results for the groundwater pathway are presented in Appendix C.

Specific dose (µSv y-1 per MBq)


Radionuclide Borehole

1500m Irrigation

Site boundary Drinking


1500m Irrigation


H-3 3.66E-30 6.89E-23 Ra-226 1.56E-08 1.36E-06

C-14 2.06E-09 1.39E-07 Ac-227 5.66E-26 6.70E-19

Cl-36 6.52E-08 1.69E-06 Th-229 1.44E-08 1.29E-06

Fe-55 1.04E-43 4.45E-36 Th-230 8.24E-09 7.35E-07

Co-60 1.21E-39 3.79E-32 Th-232 4.04E-08 3.59E-06

Ni-63 7.94E-21 3.47E-15 Pa-231 3.60E-08 3.08E-06

Sr-90 3.04E-24 2.19E-17 U-232 5.07E-20 6.19E-14

Nb-94 3.76E-10 9.07E-09 U-233 4.62E-09 4.13E-07

Tc-99 2.12E-09 1.52E-07 U-234 4.35E-09 3.89E-07

Ru-106 3.44E-46 1.67E-38 U-235 4.35E-09 3.87E-07

Ag-108m 1.68E-17 1.51E-12 U-236 4.22E-09 3.78E-07

Sb-125 4.81E-42 2.03E-34 U-238 4.35E-09 3.89E-07

Sn-126 1.20E-08 4.87E-07 Np-237 1.22E-06 9.52E-05

I-129 1.38E-05 3.02E-04 Pu-238 1.86E-12 1.67E-10

Ba-133 1.44E-31 3.25E-24 Pu-239 1.54E-08 1.37E-06

Cs-134 1.82E-43 8.12E-36 Pu-240 1.04E-08 9.33E-07

Cs-137 1.28E-25 9.32E-19 Pu-241 1.59E-10 3.62E-08

Pm-147 3.41E-44 1.47E-36 Pu-242 1.69E-08 1.51E-06






1500m Irrigation



1500m Irrigation


Eu-152 2.77E-32 4.84E-25 Am-241 5.37E-11 1.08E-08

Eu-154 3.28E-35 7.98E-28 Cm-243 1.82E-10 1.63E-08

Eu-155 3.63E-41 1.29E-33 Cm-244 1.29E-09 1.15E-07

Pb-210 1.25E-25 1.43E-18

Table 5.1 Specific doses to members of the public via the groundwater pathway. Results include doses arising from ingrowth of daughter radionuclides for 100 years.

The results from the sensitivity studies show that the calculated doses for most radionuclides are not sensitive to leachate head. This is because the most significant releases take place after the engineered barriers (cap and liner) have degraded and radionuclide transport into groundwater is governed by the infiltration rate through soil and the properties of the geological barrier. Radionuclides with short half-lives do show some sensitivity to leachate head, because they have largely decayed by the time the barriers degrade. Calculated doses for these radionuclides are determined by the relatively small releases while the barriers are effective, and the magnitude of these releases is governed by the leachate head. The small changes in calculated doses for longer-lived radionuclides arise because some of the inventory is lost from the site while the barriers are effective, and this reduces the inventory available for later release.

Results from the sensitivity studies for cap lifetime again show a dependency on half-life. Calculated doses for long-lived radionuclides show little variation with cap lifetime because they are not released in significant amounts during the period the cap is effective. Radionuclides with shorter half-lives show a sensitivity to cap lifetime; this affects whether a significant inventory is still available for release once the engineered barriers have degraded.

Sensitivity studies for cap efficiency show that this has relatively little effect on calculated doses. This is because it is the barriers at the base of the landfill that have most effect on the release of radionuclides during the period when the engineered barriers are effective. In practice, a less effective cap would allow more infiltration which would lead to an increase in leachate head and potentially to bath-tubbing if the site was not monitored and managed. The assessment methodology used does not explicitly model these links and so the secondary effects of changes in cap efficiency are not apparent in the calculated doses.

In the case of the sensitivity studies on the effects of varying the assessment period, the reverse of the effects discussed above is apparent – calculated doses for short-lived radionuclides show little or no sensitivity and those for long-lived radionuclides are very sensitive. This is because radionuclides are released only slowly to the



groundwater system, even after the engineered barriers become degraded. The geological barrier in particular has an important role in retarding transport of radionuclides. For short-lived radionuclides, the barriers are effective enough that radioactive decay reduces the inventory to less significant levels during the assessment period. For long-lived radionuclides, even a 5,000 year assessment period does not lead to significant radioactive decay and there is no peak in the calculated dose. For these radionuclides, the assessment model is effectively assuming that the inventory is transferred to a part of the accessible environment where it leads to increasing calculated doses with time. In practice, there will be more dispersion of the radionuclides over long periods, reducing calculated doses.

A comparison of calculated specific doses for exposed individuals in different age groups shows that for the majority of the radionuclides assessed, specific doses to adults are higher than those to infants or children. This arises because the adult rates of consumption for foodstuffs grown on contaminated soil are sufficiently higher then those for infants and children to off-set the higher dose coefficients for these age groups. In the case of Cl-36, specific doses to children and infants are higher than those to adults, but the difference is less than a factor of 10.

5.2 Irradiation pathway

Specific doses through external irradiation to members of the public living on the site after closure, are presented in Table 5.2. These calculations are based on the assumptions described in Section 4.1 and Appendix A (Section A.2).

Radionuclide Specific dose

(µSv y-1 per MBq) Radionuclide


H-3 0.00E+00 Ra-226 3.14E-18

C-14 1.29E-45 Ac-227 7.80E-23

Cl-36 3.09E-20 Th-229 2.24E-21

Fe-55 0.00E+00 Th-230 8.06E-20

Co-60 1.17E-14 Th-232 3.41E-22

Ni-63 0.00E+00 Pa-231 1.70E-17

Sr-90 2.22E-25 U-232 9.70E-47

Nb-94 2.49E-12 U-233 1.27E-23

Tc-99 2.61E-33 U-234 1.39E-22

Ru-106 1.05E-32 U-235 2.15E-20

Ag-108m 3.23E-13 U-236 4.15E-23

Sb-125 1.07E-20 U-238 2.96E-24

Sn-126 1.76E-26 Np-237 1.65E-27

I-129 1.96E-91 Pu-238 1.42E-22

Ba-133 5.59E-15 Pu-239 6.36E-28




(µSv y-1 per MBq) Radionuclide


Cs-134 8.72E-22 Pu-240 2.04E-27

Cs-137 1.97E-21 Pu-241 2.55E-30

Pm-147 2.92E-34 Pu-242 8.14E-33

Eu-152 1.90E-13 Am-241 4.84E-30

Eu-154 4.30E-14 Cm-243 4.80E-16

Eu-155 5.99E-31 Cm-244 4.43E-28

Pb-210 6.65E-62

Table 5.2 Specific doses to workers and members of the public via the external irradiation pathway. Results for doses to the public include doses arising from ingrowth of daughter radionuclides for 60 years.

5.3 Intrusion

Specific doses to workers intruding into the waste and to members of the public living on excavated waste after intrusion are presented in Table 5.3. These calculations assume that the LLW is disposed of to all of the remaining cells at the site. Other assumptions are described in Appendix A (Section A.3).


Radionuclide Intruder 20 years

Intruder 60 years

Resident 60 years

Resident 100 years

H-3 4.31E-11 4.52E-12 1.28E-10 1.35E-11

C-14 7.09E-09 7.05E-09 8.68E-09 8.63E-09

Cl-36 3.02E-08 3.02E-08 4.28E-07 4.28E-07

Fe-55 1.40E-12 4.85E-17 8.13E-18 2.82E-22

Co-60 2.44E-06 1.27E-08 4.19E-11 2.17E-13

Ni-63 1.16E-09 8.68E-10 2.60E-10 1.95E-10

Sr-90 2.49E-07 9.59E-08 3.31E-07 1.28E-07

Nb-94 7.53E-05 7.52E-05 2.44E-07 2.44E-07

Tc-99 1.09E-08 1.09E-08 3.68E-07 3.68E-07

Ru-106 1.16E-14 1.39E-26 7.44E-29 8.90E-41

Ag-108m 6.38E-05 5.13E-05 1.68E-07 1.35E-07

Sb-125 1.13E-08 5.68E-13 1.89E-15 9.47E-20

Sn-126 1.34E-04 1.34E-04 4.90E-07 4.90E-07

I-129 1.06E-06 1.06E-06 1.36E-06 1.36E-06

Ba-133 2.20E-06 1.65E-07 5.63E-10 4.22E-11




Radionuclide Intruder 20 years

Intruder 60 years

Resident 60 years

Resident 100 years

Cs-134 3.70E-09 6.83E-15 2.76E-17 5.09E-23

Cs-137 1.41E-05 5.60E-06 2.61E-08 1.04E-08

Pm-147 1.52E-12 3.54E-17 4.55E-19 1.06E-23

Eu-152 1.14E-05 1.42E-06 4.63E-09 5.76E-10

Eu-154 5.63E-06 2.41E-07 7.87E-10 3.37E-11

Eu-155 2.15E-08 8.04E-11 2.74E-13 1.03E-15

Pb-210 7.61E-06 2.19E-06 1.28E-07 3.70E-08

Ra-226 1.10E-04 1.08E-04 5.40E-06 5.31E-06

Ac-227 7.65E-05 2.14E-05 1.81E-08 5.07E-09

Th-229 9.52E-05 9.48E-05 8.84E-08 8.80E-08

Th-230 3.31E-05 3.49E-05 1.56E-07 2.43E-07

Th-232 1.87E-04 1.87E-04 4.75E-07 4.75E-07

Pa-231 8.60E-05 6.82E-05 1.66E-06 1.65E-06

U-232 1.34E-05 8.99E-06 1.57E-08 1.05E-08

U-233 3.54E-06 3.89E-06 4.32E-09 4.65E-09

U-234 3.28E-06 3.29E-06 3.71E-09 3.78E-09

U-235 8.92E-06 8.96E-06 2.47E-08 2.61E-08

U-236 1.38E-06 1.38E-06 3.14E-09 3.14E-09

U-238 3.88E-06 3.88E-06 6.87E-09 6.88E-09

Np-237 2.47E-05 2.47E-05 4.65E-08 4.65E-08

Pu-238 2.79E-05 2.03E-05 9.86E-09 7.19E-09

Pu-239 3.85E-05 3.84E-05 1.86E-08 1.86E-08

Pu-240 3.84E-05 3.82E-05 1.85E-08 1.84E-08

Pu-241 1.74E-06 1.47E-05 8.29E-09 5.27E-08

Pu-242 3.54E-05 3.54E-05 1.76E-08 1.76E-08

Am-241 2.97E-05 2.79E-05 1.57E-08 1.47E-08

Cm-243 7.50E-06 3.03E-06 4.49E-09 1.92E-09

Cm-244 6.43E-06 1.57E-06 9.45E-10 2.38E-09

Table 5.3 Specific doses to intruders into the landfill and to members of the public living on excavated waste. Results include doses arising from ingrowth of daughter radionuclides over 20, 60 and 100 years as appropriate.



5.4 Leachate management and spillage

5.4.1 Leachate management

Specific doses to members of three exposed groups at a sewage treatment works receiving leachate from the ENRMF are presented in Table 5.4. These results are based on the Environment Agency’s Initial Assessment methodology (Environment Agency 2006a; 2006b), which does not include the same range of radionuclides as used in the remainder of the radiological assessment.

These results are based on the assumption that the sewage treatment works has an overall throughput of 1,080 m3 of effluent per day and that the average exchange rate in the estuary is 30 m3 / s. The specific doses for the STW worker and farming family are sensitive to the throughput (decreasing as overall throughput increases), and the specific doses for the fisherman are sensitive to the exchange rate (decreasing as the exchange rate increases).

Specific dose (microSv / y per MBq / y)

Radionuclide STW worker Farming family Fisherman

H-3 2.11E-09 2.83E-06 2.52E-09

C-14 7.78E-08 4.72E-03 1.30E-03

Cl-36 1.33E-06 7.78E-02 4.80E-09

Fe-55 2.00E-07 1.33E-03 1.00E-07

Co-60 4.94E-02 7.78E-01 1.87E-03

Sr-90 2.28E-05 2.17E-02 1.83E-05

Tc-99 1.17E-07 2.83E-01 2.10E-05

Ru-106 6.11E-04 3.06E-03 1.44E-04

I-129 2.44E-05 6.11E-02 6.67E-05

Cs-134 1.11E-02 1.17E-01 2.80E-04

Cs-137 4.11E-03 1.00E-01 3.50E-04

Pm-147 2.11E-07 1.67E-05 6.50E-07

Eu-152 1.39E-02 2.67E-01 3.67E-03

Eu-154 1.50E-02 2.72E-01 3.33E-03

Eu-155 3.33E-04 5.06E-03 6.17E-05

Pb-210 4.44E-04 5.33E-01 6.33E-02

Ra-226 2.22E-02 5.56E-01 1.83E-03

Th-230 3.22E-04 1.28E-02 3.67E-05

Th-232 4.89E-04 1.39E+00 2.23E-03



U-234 1.11E-05 1.17E-03 3.90E-05

U-235 3.61E-04 7.78E-03 6.60E-05

U-238 8.33E-05 2.06E-03 4.20E-05

Np-237 3.67E-04 7.22E-02 6.00E-04

Pu-238 4.44E-04 2.00E-02 2.67E-03

Pu-239 4.83E-04 2.28E-02 2.83E-03

Pu-240 4.83E-04 2.28E-02 2.83E-03

Pu-241 8.89E-06 3.39E-04 5.33E-05

Pu-242 4.67E-04 2.22E-02 2.67E-03

Am-241 8.33E-04 3.94E-02 2.37E-05

Cm-243 2.44E-03 6.67E-02 1.00E-04

Cm-244 4.50E-04 1.78E-02 9.00E-06

Table 5.4 Specific doses to workers at a sewage treatment works (STW) receiving leachate from the ENRMF and to members of the public using resources contaminated by effluent from the STW.

5.4.2 Leachate spillage

Specific doses calculated for members of the public via the leachate spillage pathway are presented in Table 5.5. These calculations are based on the assumptions described in Section 4.1.8 and in Appendix A.


Pathways associated with water contaminated

by leachate

Pathways associated with soil contaminated by irrigation with

contaminated water Radionuclide

Drinking water

Fish Crops Livestock and

associated products

Soil

H-3 1.58E-12 4.34E-15 3.53E-11 5.17E-13 1.50E-17

C-14 5.11E-11 1.26E-09 1.12E-09 2.08E-11 4.84E-16

Cl-36 8.19E-11 1.12E-11 1.82E-09 4.26E-11 7.76E-16

Fe-55 2.91E-11 7.96E-12 6.39E-10 1.54E-12 2.76E-16

Co-60 2.99E-10 2.46E-10 6.59E-09 8.64E-12 2.84E-15

Ni-63 1.32E-11 3.62E-12 2.91E-10 1.92E-12 1.25E-16

Sr-90 2.70E-09 4.44E-10 5.96E-08 2.02E-10 2.56E-14

Nb-94 1.50E-10 1.23E-10 3.29E-09 6.25E-16 1.42E-15

Tc-99 5.63E-11 3.09E-12 1.28E-09 7.94E-14 5.34E-16

Ru-106 6.16E-10 1.69E-11 1.36E-08 8.19E-11 5.84E-15





by leachate



Drinking water


associated products

Soil

Ag-108m 2.02E-10 2.77E-12 4.46E-09 1.03E-13 1.92E-15

Sb-125 2.74E-10 7.50E-11 6.03E-09 8.54E-14 2.60E-15

Sn-126 6.25E-10 1.71E-09 1.38E-08 8.63E-12 5.93E-15

I-129 9.68E-09 7.96E-10 2.13E-07 1.86E-09 9.18E-14

Ba-133 1.32E-10 1.45E-12 2.91E-09 7.24E-13 1.25E-15

Cs-134 1.67E-09 9.17E-09 3.68E-08 3.31E-10 1.59E-14

Cs-137 1.14E-09 6.27E-09 2.52E-08 2.27E-10 1.09E-14

Pm-147 2.29E-11 1.88E-12 5.04E-10 3.06E-13 2.17E-16

Eu-152 1.23E-10 1.01E-11 2.71E-09 2.04E-13 1.17E-15

Eu-154 1.76E-10 1.45E-11 3.88E-09 2.91E-13 1.67E-15

Eu-155 2.82E-11 2.32E-12 6.20E-10 4.65E-14 2.67E-16

Pb-210 1.66E-07 1.37E-07 3.66E-06 5.88E-10 1.58E-12

Ra-226 1.91E-07 2.62E-08 4.21E-06 2.52E-09 1.81E-12

Ac-227 1.06E-07 2.33E-07 2.34E-06 4.52E-11 1.01E-12

Th-229 5.40E-08 4.44E-09 1.19E-06 3.87E-10 5.12E-13

Th-230 1.85E-08 1.52E-09 4.07E-07 1.32E-10 1.75E-13

Th-232 9.36E-08 7.70E-09 2.06E-06 6.70E-10 8.88E-13

Pa-231 6.25E-08 1.71E-09 1.38E-06 1.09E-11 5.93E-13

U-232 4.05E-08 1.11E-09 8.91E-07 1.66E-10 3.84E-13

U-233 4.49E-09 1.23E-10 9.88E-08 1.83E-11 4.26E-14

U-234 4.31E-09 1.18E-10 9.49E-08 1.76E-11 4.09E-14

U-235 4.17E-09 1.14E-10 9.17E-08 1.70E-11 3.95E-14

U-236 4.14E-09 1.13E-10 9.11E-08 1.69E-11 3.92E-14

U-238 4.26E-09 1.17E-10 9.38E-08 1.74E-11 4.04E-14

Np-237 9.76E-09 2.67E-10 2.15E-07 2.62E-11 9.26E-14

Pu-238 2.02E-08 2.22E-10 4.46E-07 7.17E-13 1.92E-13

Pu-239 2.20E-08 2.41E-10 4.84E-07 7.80E-13 2.09E-13

Pu-240 2.20E-08 2.41E-10 4.84E-07 7.80E-13 2.09E-13

Pu-241 4.23E-10 4.63E-12 9.30E-09 1.50E-14 4.01E-15

Pu-242 2.11E-08 2.32E-10 4.65E-07 7.49E-13 2.00E-13





by leachate



Drinking water


associated products

Soil

Am-241 1.76E-08 1.45E-09 3.88E-07 2.08E-12 1.67E-13

Cm-243 1.32E-08 1.09E-09 2.91E-07 3.59E-12 1.25E-13

Cm-244 1.07E-08 8.76E-10 2.34E-07 2.90E-12 1.01E-13

Table 5.5 Specific doses via exposure pathways associated with spillage of leachate into a surface water resource. Results do not include the effects of ingrowth of long-lived daughter radionuclides.

Sensitivity studies to show the effect of different assumptions about the exposed individual are presented in Appendix C, where specific doses for an adult, an infant (1 year old) and child (10 year old) are given for the different potential exposure pathways following a leachate spill to a surface water body. A comparison of the calculated specific doses for different age groups shows that for pathways associated with the consumption of fish and crops the greater consumption rates for adults outweigh the higher dose coefficients for infants and children. For the pathways associated with livestock and drinking water, the age group with the highest specific dose depends on the radionuclide, but in all cases the ratio of specific doses between age groups is less than a factor of 10. Only in the case of exposure through the consumption of contaminated soil are specific doses to infants and children higher than those to adults by a factor of more than 10, because of the greater consumption rates assumed for these age groups.

5.4.3 Aerosol pathway

Specific doses calculated for workers and members of the public via the aerosol pathway are presented in Table 5.6. These calculations are based on the assumptions described in Section 4.1.8 and in Appendix A, and are presented as specific doses per hour of exposure to aerosols.

Specific dose (µSv y-1 per MBq per

hour)


hour) Radionuclide

Workers Public

Radionuclide

Workers Public

H-3 2.51E-12 2.09E-12 Ra-226 1.88E-07 1.57E-07

C-14 5.6E-11 4.66E-11 Ac-227 5.49E-06 4.57E-06

Cl-36 7.04E-11 5.87E-11 Th-229 2.47E-06 2.06E-06

Fe-55 7.43E-12 6.19E-12 Th-230 9.65E-07 8.04E-07

Co-60 2.99E-10 2.49E-10 Th-232 1.64E-06 1.36E-06




hour)


hour) Radionuclide

Workers Public

Radionuclide

Workers Public

Ni-63 4.63E-12 3.86E-12 Pa-231 1.35E-06 1.13E-06

Sr-90 1.56E-09 1.30E-09 U-232 4.52E-07 3.77E-07

Nb-94 1.06E-10 8.84E-11 U-233 9.26E-08 7.72E-08

Tc-99 1.25E-10 1.05E-10 U-234 9.07E-08 7.56E-08

Ru-106 6.37E-10 5.31E-10 U-235 8.20E-08 6.83E-08

Ag-108m 3.57E-10 2.97E-10 U-236 3.09E-08 2.57E-08

Sb-125 5.27E-11 4.39E-11 U-238 7.73E-08 6.44E-08

Sn-126 3.01E-10 2.51E-10 Np-237 4.82E-07 4.02E-07

I-129 3.47E-10 2.89E-10 Pu-238 1.06E-06 8.84E-07

Ba-133 2.99E-11 2.49E-11 Pu-239 1.16E-06 9.65E-07

Cs-134 6.56E-11 5.47E-11 Pu-240 1.16E-06 9.65E-07

Cs-137 3.76E-10 3.14E-10 Pu-241 2.22E-08 1.85E-08

Pm-147 4.82E-11 4.02E-11 Pu-242 1.06E-06 8.84E-07

Eu-152 4.05E-10 3.38E-10 Am-241 9.26E-07 7.72E-07

Eu-154 5.11E-10 4.26E-10 Cm-243 3.00E-07 2.50E-07

Eu-155 6.66E-11 5.55E-11 Cm-244 2.61E-07 2.18E-07

Pb-210 9.64E-08 8.03E-08

Table 5.6 Specific doses to workers and members of the public through exposure to aerosols during the operational phase. Results do not include the effects of ingrowth of long-lived daughter radionuclides.

5.5 Gas pathway

Specific doses via the gas pathway to workers, members of the public living near the site and members of the public living on the site after closure are presented in Table 5.7. These calculations are based on the assumptions described in Appendix A (Sections A.1 and A.2).

Although carbon-based gases (e.g., CO, CO2, CH4) are likely to be present within the landfill, and may be collected and flared, it is unlikely that the processes generating such gases would take place within cells dominated by LLW. C-14 would therefore not be released as a gas and is excluded from this assessment.



Specific dose (microSv y-1 per MBq)

Radionuclide Worker

Public near site

Resident

after closure

H-3 5.44E-08 1.13E-07 3.43E-08

Ra-226 1.19E-07 2.48E-07 2.85E-06

Th-230 - - 7.3E-08

U-234 - - 2.02E-11

U-238 - - 2.57E-12

Pu-238 - - 1.64E-15

Pu-242 - - 6.46E-21

Table 5.7 Specific doses to workers and to members of the public via the gas pathway. Doses to residents after closure include the effects of ingrowth of daughter radionuclides for 60 years.

5.7 Dose rates to wildlife

As noted in Section 3.3, a radiological assessment of the potential effects of LLW disposal at the ENRMF on wildlife has been undertaken. This assessment has been undertaken using the Tier 1 approach within the assessment tool developed as part of the ERICA project (Environmental Risk from Ionising Contaminants: Assessment and Management).

The ERICA toolkit allows for consideration of three ecosystems: terrestrial, freshwater and marine. Only the first two of these have been considered for the ENRMF. Within these ecosystems, the ERICA tool considers a range of organisms and wildlife groups (Table 5.9).

Terrestrial Freshwater

Bird Amphibian

Bird egg Benthic fish

Detritivorous invertebrate Bird

Flying insects Bivalve mollusc

Gastropod Crustacean

Grasses & Herbs Gastropod

Lichen & Bryophytes Insect larvae

Mammal (Deer) Mammal

Mammal (Rat) Pelagic fish

Reptile Phytoplankton

Shrub Vascular plant



Soil Invertebrate (worm) Zooplankton

Tree

Table 5.9 Organisms and wildlife groups included in the two ecosystems considered in the assessment of impacts on wildlife.

Within the Tier 1 assessment, the ERICA tool compares environmental concentrations for individual radionuclides with limiting concentrations calculated using generic assumptions about the ecosystems. The limiting concentrations are based on a screening dose rate of 10 μGy h-1. Table 5.10 presents the limiting concentrations for the radionuclides considered in the wildlife assessment1. Note that the calculated dose rate for the same environmental concentration differs between organisms and therefore the limiting concentration does not necessarily apply to the same organism or wildlife group for each radionuclide. Within the Tier 1 assessment, it is assumed that all of the organisms and groups listed in Table 5.9 are present, so that the limiting concentration is the lowest calculated for any organism.

Freshwater ecosystem Terrestrial ecosystem Radio-nuclide

Bq l-1 Limiting organism Bq kg-1 Limiting organism

Am-241 2.63E-03 Phytoplankton 6.25E+02 Flying insects

C-14 1.56E+01 Bird 8.33E+01 Mammal (Deer)

Cl-36 1.06E+02 Vascular plant 1.47E+03 Grasses & Herbs

Cm-243 5.13E-03 Zooplankton 7.19E+02 Flying insects, Gastropod

Cm-244 5.18E-03 Zooplankton 7.30E+02 Flying insects, Gastropod

Co-60 1.87E-02 Insect larvae 7.35E+03 Mammal (Rat)

Cs-134 2.06E-02 Insect larvae 1.67E+03 Mammal (Deer)

Cs-137 5.10E-02 Insect larvae 3.13E+03 Mammal (Deer)

Eu-152 7.19E+00 Vascular plant 1.72E+04 Soil Invertebrate (worm), Detritivorous invertebrate

Eu-154 7.14E+00 Vascular plant 1.56E+04 Detritivorous invertebrate

H-3 3.45E+05 Phytoplankton 2.60E+03 Detritivorous invertebrate

I-129 2.75E+01 Phytoplankton 4.26E+02 Bird egg

Ni-63 4.17E+01 Gastropod 1.08E+06 Grasses & Herbs

Np-237 3.05E-03 Phytoplankton 3.77E+02 Shrub

Pb-210 7.87E-02 Phytoplankton 3.88E+03 Lichen & bryophytes

Pu-238 2.14E-02 Phytoplankton 1.02E+03 Lichen & bryophytes


1 The ERICA tool does not include data for the complete range of radionuclides considered in the other parts of the radiological assessment.




Pu-241 8.47E+01 Phytoplankton 4.05E+06 Lichen & bryophytes

Ra-226 1.40E-02 Vascular plant 2.27E+02 Lichen & bryophytes

Ru-106 1.28E-01 Insect larvae 8.20E+02 Lichen & bryophytes

Sb-125 8.40E-01 Insect larvae 3.73E+04 Detritivorous invertebrate

Sr-90 3.51E+00 Insect larvae 3.76E+02 Reptile

Tc-99 5.05E+01 Vascular plant 2.11E+03 Bird egg

Th-230 3.10E-02 Phytoplankton 1.63E+03 Lichen & bryophytes

Th-232 3.64E-02 Phytoplankton 1.90E+03 Lichen & bryophytes

U-234 4.22E-02 Vascular plant 1.67E+03 Lichen & bryophytes



Table 5.10 Limiting concentrations in freshwater and terrestrial ecosystems, based on a dose rate of 10 μGy h-1 to the limiting organism or wildlife group.

For the purposes of the wildlife assessment, the modified SNIFFER model described in Section 4.2 has been used to calculate radionuclide concentrations in a hypothetical stream close to the site boundary. This stream is assumed to receive baseflow from groundwater using the same assumptions as used for the drinking water pathway (Section 5.1). For the terrestrial ecosystem, it is assumed that this stream periodically floods an adjacent area of land. The SNIFFER model does not explicitly model flooding, and the irrigation model is therefore used to determine radionuclide concentrations in soil.

Radionuclide concentrations in the freshwater and terrestrial ecosystems calculated using the SNIFFER model do not reflect actual concentrations as they are based on unit disposal (1 MBq) of each radionuclide. To allow comparison with the limiting concentrations presented in Table 5.10, it is conservatively assumed that the site receives the maximum amount of each radionuclide that keeps the site below the radiological capacity (see Section 6.2). Environmental concentrations based on disposal of radionuclides at capacity are presented in Tables 5.11 and 5.12, together with calculated risk quotients. The risk quotient for a radionuclide is the highest value of the ratio between calculated dose rate and the 10 μGy h-1 criterion (i.e., a risk quotient of 1 or greater would indicate that the screening criterion was exceeded).

Radionulide Soil concentration (Bq kg-1)

Risk quotient

Limiting reference organism

Am-241 4.92E-03 7.87E-06 Flying insects

C-14 2.22E+00 2.66E-02 Mammal (Deer)

Cl-36 5.28E-02 3.59E-05 Grasses & Herbs

Cm-243 1.60E-13 2.22E-16 Flying insects, Gastropod



Radionulide Soil concentration (Bq kg-1)

Risk quotient


Cm-244 3.41E-17 4.67E-20 Flying insects, Gastropod

Co-60 1.12E-23 1.52E-27 Mammal (Rat)

Cs-134 1.41E-22 8.45E-26 Mammal (Deer)

Cs-137 5.11E-13 1.64E-16 Mammal (Deer)

Eu-152 6.76E-18 3.92E-22 Soil Invertebrate (worm), Detritivorous invertebrate

Eu-154 3.25E-20 2.08E-24 Detritivorous invertebrate

H-3 8.14E-13 3.13E-16 Detritivorous invertebrate

I-129 1.27E-04 2.98E-07 Bird egg

Ni-63 4.79E-05 4.42E-11 Grasses & Herbs

Np-237 1.52E-03 4.03E-06 Shrub

Pb-210 8.45E-16 2.18E-19 Lichen & bryophytes

Pu-238 4.42E-08 4.34E-11 Lichen & bryophytes




Ra-226 5.04E-04 2.22E-06 Lichen & bryophytes

Ru-106 1.62E-13 1.98E-16 Lichen & bryophytes

Sb-125 4.03E-22 1.08E-26 Detritivorous invertebrate

Sr-90 2.83E-13 7.53E-16 Reptile

Tc-99 2.06E-02 9.74E-06 Bird egg

Th-230 1.29E-02 7.92E-06 Lichen & bryophytes

Th-232 2.76E-03 1.45E-06 Lichen & bryophytes

U-234 2.25E-02 1.35E-05 Lichen & bryophytes



Table 5.11 Calculated soil concentrations based on disposal of radionuclides at the individual radiological capacity. Risk quotients based on the 10 μGy h-1 dose rate criterion.



Radionulide Water concentration (Bq m-3)

Risk quotient


Am-241 4.50E-02 1.71E-02 Phytoplankton

C-14 1.41E+01 9.07E-04 Bird

Cl-36 3.43E+00 3.23E-05 Vascular plant

Cm-243 1.66E-11 3.24E-12 Zooplankton

Cm-244 5.45E-15 1.05E-15 Zooplankton

Co-60 5.29E-21 2.83E-22 Insect larvae

Cs-134 1.74E-19 8.44E-21 Insect larvae

Cs-137 5.20E-11 1.02E-12 Insect larvae

Eu-152 1.42E-15 1.97E-19 Vascular plant

Eu-154 9.62E-18 1.35E-21 Vascular plant

H-3 5.67E-10 1.64E-18 Phytoplankton

I-129 2.43E-02 8.82E-07 Phytoplankton

Ni-63 1.67E-03 4.01E-08 Gastropod

Np-237 8.70E-02 2.85E-02 Phytoplankton

Pb-210 1.12E-13 1.42E-15 Phytoplankton

Pu-238 1.66E-06 7.75E-08 Phytoplankton




Ra-226 2.23E-03 1.59E-04 Vascular plant

Ru-106 3.42E-10 2.67E-12 Insect larvae

Sb-125 3.91E-19 4.65E-22 Insect larvae

Sr-90 4.07E-11 1.16E-14 Insect larvae

Tc-99 1.25E+01 2.48E-04 Vascular plant

Th-230 4.81E-02 1.55E-03 Phytoplankton

Th-232 1.03E-02 2.83E-04 Phytoplankton

U-234 2.24E-01 5.31E-03 Vascular plant



Table 5.12 Calculated water concentrations based on disposal of radionuclides at the individual radiological capacity. Risk quotients based on the 10 μGy h-1 dose rate criterion.



The ERICA assessment tool allows three tiers of assessment. A Tier 1 assessment has been undertaken and the calculated incremental dose rate values are all below the recommended screening value. More detailed assessments (Tier 2 and Tier 3) are therefore not required.



6 Radiological Capacity

6.1 Introduction

Section 5 presents the calculated doses for a range of scenarios and exposure pathways. These are presented as specific doses, the annual dose calculated to arise from a disposal of 1 MBq of the radionuclide concerned. The actual doses will depend on how much radioactive waste is actually disposed of to the site. Because of the conservative assumptions involved in the assessment model any doses received in the future are likely to be significantly less that these calculated doses. These specific doses do, however, provide a basis for decisions about the amount of waste that can be disposed.

The radiological capacity is the amount that can be disposed without the calculated doses exceeding the appropriate dose criterion. Two types of radiological capacity can be calculated. The first is the capacity for individual radionuclides, which represents how much of any one radionuclide could be disposed of. If radionuclides are completely independent, then these capacities can be apportioned directly to the radionuclides – e.g., 50% of the capacity to radionuclide A and 50% to radionuclide B.

In practice, radionuclides are not independent and are present in waste streams in certain ratios. In this case, the capacity cannot be directly apportioned to the radionuclides and must take account of both the specific dose for each radionuclide and the proportion of the radionuclide in the waste stream.


DCfSD

fRC

ii

ii

where:

fi is the fraction of the overall activity arising from Rni (such that fi=1)

SDi is the specific dose from Rni

DC is the dose constraint

Radiological capacities for mixtures of waste streams can be calculated by apportioning part of the overall capacity to each waste stream.

The following sections present radiological capacities at the ENRMF for individual radionuclides (Section 6.2) and overall capacities based on an illustrative waste stream from the UKAEA decommissioning programme at Harwell (see Section 2.2).

Section 5 presents specific doses for a range of scenarios and exposure pathways, but radiological capacities have only been calculated for the principal exposure routes:



groundwater and intrusion. Specific doses through at least one of these routes are higher than the corresponding doses through irradiation, and irradiation will not therefore determine the overall capacity. The aerosol and leachate spillage pathways are all highly uncertain, both in terms of the possibility of occurring and duration. The specific doses presented in Section 5 are illustrative, and might be considered in establishing mitigation measures, but should not be used to determine overall capacities.

The specific doses calculated as a result of off-site leachate management are based on a generic model for a sewage treatment plant, with significant uncertainties regarding the extent of dilution by other waste streams and the type and fat of effluents. This model again calculates specific doses based on unit radioactivity inputs. As discussed elsewhere, there are large uncertainties about the rate at which radionuclides would enter the leachate at the ENRMF. Conservative assumptions have been made so that future doses arising from releases to groundwater or accidental spillage of leachate can be calculated. It would, however, be unreasonable to apply these same assumptions to the routine management of leachate during the operational period. Even if the same generic model for the sewage treatment plant is used to estimate specific doses, it would be more appropriate to determine a permissible level of radioactivity in leachate and then to develop authorisation conditions from these. Monitoring of leachate would ensure compliance with these conditions and give more control than applying conservative assumptions.

As discussed in Section 3.2.1, two dose criteria have been used to determine radiological capacity. For exposures arising from releases to groundwater and subsequent use of an existing borehole for irrigation purposes, a dose criterion of 20 μSv / year has been used. For exposures resulting from intrusion into the waste, from excavation and subsequent use of waste as soil, and from consumption of groundwater extracted from close to the site boundary, a dose criterion of 3 mSv / year has been used.

6.2 Radionuclide-specific radiological capacities

Tables 6.1 to 6.3 present specific doses and the corresponding radionuclide-specific capacities for the two groundwater pathways and the principal intrusion pathway described in Sections 4 and 5.

Radionuclide

Specific dose

(µSv y-1 per MBq)

Radiological capacity

(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

H-3 3.66E-30 5.47E+30 Ra-226 1.56E-08 1.28E+09

C-14 2.06E-09 9.69E+09 Ac-227 5.66E-26 3.53E+26

Cl-36 6.52E-08 3.07E+08 Th-229 1.44E-08 1.39E+09

Fe-55 1.04E-43 1.92E+44 Th-230 8.24E-09 2.43E+09

Co-60 1.21E-39 1.65E+40 Th-232 4.04E-08 4.95E+08



Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

Ni-63 7.94E-21 2.52E+21 Pa-231 3.60E-08 5.56E+08

Sr-90 3.04E-24 6.59E+24 U-232 5.07E-20 3.94E+20

Nb-94 3.76E-10 5.33E+10 U-233 4.62E-09 4.33E+09

Tc-99 2.12E-09 9.45E+09 U-234 4.35E-09 4.60E+09

Ru-106 3.44E-46 5.81E+46 U-235 4.35E-09 4.60E+09

Ag-108m 1.68E-17 1.19E+18 U-236 4.22E-09 4.74E+09

Sb-125 4.81E-42 4.16E+42 U-238 4.35E-09 4.59E+09

Sn-126 1.20E-08 1.67E+09 Np-237 1.22E-06 1.64E+07

I-129 1.38E-05 1.45E+06 Pu-238 1.86E-12 1.07E+13

Ba-133 1.44E-31 1.39E+32 Pu-239 1.54E-08 1.30E+09

Cs-134 1.82E-43 1.10E+44 Pu-240 1.04E-08 1.92E+09

Cs-137 1.28E-25 1.57E+26 Pu-241 1.59E-10 1.26E+11

Pm-147 3.41E-44 5.87E+44 Pu-242 1.69E-08 1.19E+09

Eu-152 2.77E-32 7.21E+32 Am-241 5.37E-11 3.73E+11

Eu-154 3.28E-35 6.09E+35 Cm-243 1.82E-10 1.10E+11

Eu-155 3.63E-41 5.50E+41 Cm-244 1.29E-09 1.55E+10

Pb-210 1.25E-25 1.60E+26

Table 6.1 Specific doses to members of the public via use of water from a borehole 1500 m from the site boundary for irrigation, and corresponding radiological capacities. Radiological capacities are based on a 20 μSv / year dose criterion. Results include doses arising from ingrowth of daughter radionuclides for 100 years.

Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

H-3 6.89E-23 4.36E+25 Ra-226 1.36E-06 2.20E+09

C-14 1.39E-07 2.16E+10 Ac-227 6.70E-19 4.48E+21

Cl-36 1.69E-06 1.78E+09 Th-229 1.29E-06 2.33E+09

Fe-55 4.45E-36 6.74E+38 Th-230 7.35E-07 4.08E+09

Co-60 3.79E-32 7.91E+34 Th-232 3.59E-06 8.36E+08

Ni-63 3.47E-15 8.65E+17 Pa-231 3.08E-06 9.74E+08

Sr-90 2.19E-17 1.37E+20 U-232 6.19E-14 4.85E+16



Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

Nb-94 9.07E-09 3.31E+11 U-233 4.13E-07 7.26E+09

Tc-99 1.52E-07 1.97E+10 U-234 3.89E-07 7.72E+09

Ru-106 1.67E-38 1.80E+41 U-235 3.87E-07 7.75E+09

Ag-108m 1.51E-12 1.98E+15 U-236 3.78E-07 7.94E+09

Sb-125 2.03E-34 1.48E+37 U-238 3.89E-07 7.71E+09

Sn-126 4.87E-07 6.16E+09 Np-237 9.52E-05 3.15E+07

I-129 3.02E-04 9.94E+06 Pu-238 1.67E-10 1.80E+13

Ba-133 3.25E-24 9.22E+26 Pu-239 1.37E-06 2.18E+09

Cs-134 8.12E-36 3.69E+38 Pu-240 9.33E-07 3.21E+09

Cs-137 9.32E-19 3.22E+21 Pu-241 3.62E-08 8.29E+10

Pm-147 1.47E-36 2.04E+39 Pu-242 1.51E-06 1.99E+09

Eu-152 4.84E-25 6.20E+27 Am-241 1.08E-08 2.77E+11

Eu-154 7.98E-28 3.76E+30 Cm-243 1.63E-08 1.84E+11

Eu-155 1.29E-33 2.32E+36 Cm-244 1.15E-07 2.61E+10

Pb-210 1.43E-18 2.09E+21

Table 6.2 Specific doses to members of the public via use of water from a borehole at the site boundary for drinking, and corresponding radiological capacities. Radiological capacities are based on a 3 mSv / year dose criterion. Results include doses arising from ingrowth of daughter radionuclides for 100 years.

Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

H-3 1.28E-10 2.34E+13 Ra-226 5.40E-06 5.56E+08

C-14 8.68E-09 3.46E+11 Ac-227 1.81E-08 1.66E+11

Cl-36 4.28E-07 7.01E+09 Th-229 8.84E-08 3.39E+10

Fe-55 8.13E-18 3.69E+20 Th-230 1.56E-07 1.92E+10

Co-60 4.19E-11 7.16E+13 Th-232 4.75E-07 6.32E+09

Ni-63 2.60E-10 1.15E+13 Pa-231 1.66E-06 1.80E+09

Sr-90 3.31E-07 9.06E+09 U-232 1.57E-08 1.91E+11

Nb-94 2.44E-07 1.23E+10 U-233 4.32E-09 6.94E+11

Tc-99 3.68E-07 8.15E+09 U-234 3.71E-09 8.09E+11



Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq) Radionuclide

Specific dose

(µSv y-1 per MBq)


(MBq)

Ru-106 7.44E-29 4.03E+31 U-235 2.47E-08 1.21E+11

Ag-108m 1.68E-07 1.79E+10 U-236 3.14E-09 9.55E+11

Sb-125 1.89E-15 1.59E+18 U-238 6.87E-09 4.37E+11

Sn-126 4.90E-07 6.12E+09 Np-237 4.65E-08 6.45E+10

I-129 1.36E-06 2.21E+09 Pu-238 9.86E-09 3.04E+11

Ba-133 5.63E-10 5.33E+12 Pu-239 1.86E-08 1.61E+11

Cs-134 2.76E-17 1.09E+20 Pu-240 1.85E-08 1.62E+11

Cs-137 2.61E-08 1.15E+11 Pu-241 8.29E-09 3.62E+11

Pm-147 4.55E-19 6.60E+21 Pu-242 1.76E-08 1.71E+11

Eu-152 4.63E-09 6.48E+11 Am-241 1.57E-08 1.91E+11

Eu-154 7.87E-10 3.81E+12 Cm-243 4.49E-09 6.68E+11

Eu-155 2.74E-13 1.09E+16 Cm-244 9.45E-10 3.17E+12

Pb-210 1.28E-07 2.34E+10

Table 6.3 Specific doses to members of the public living on excavated waste after 60 years, and corresponding radiological capacities. Radiological capacities are based on a 3 mSv / year dose criterion. Results include doses arising from ingrowth of daughter radionuclides for 60 years.

6.3 Overall radiological capacity

To illustrate the potential overall capacity of the ENRMF, waste stream data for the UKAEA Harwell Meashill Trenches (Table 2.6) has been used to calculate overall radiological capacities based on the groundwater and human intrusion pathways using the assumptions and dose criteria described above. The results presented in Tables 6.4 to 6.6 show that the calculated capacities for the groundwater and human intrusion pathways are very similar, in large part because of the different dose criteria applied. Table 6.6 also shows the contributions to the overall dose of the individual radionuclides. For the groundwater pathway, Pu-239 and Pu-240 are the key contributors to dose. In the case of human intrusion, Ra-226 is the principal contributor to dose.


(µSv y-1 per MBq)

2010 Inventory

(MBq)


(MBq)

Contribution to dose (µSv)

H-3 5.47E+30 3.25 3.62E+06 0.00

Co-60 1.65E+40 8090 9.01E+09 0.00

Cs-137 1.57E+26 952 1.06E+09 0.00

Ra-226 1.28E+09 99.6 1.11E+08 1.73




(µSv y-1 per MBq)

2010 Inventory

(MBq)


(MBq)


Th-232 4.95E+08 40 4.46E+07 1.80

U-234 4.60E+09 500 5.57E+08 2.42

U-235 4.60E+09 24 2.67E+07 0.12

U-238 4.59E+09 500 5.57E+08 2.42

Pu-238 1.07E+13 37 4.12E+07 0.00

Pu-239 1.30E+09 400 4.46E+08 6.84

Pu-240 1.92E+09 400 4.46E+08 4.65

Pu-241 1.26E+11 38.2 4.25E+07 0.01

Am-241 3.73E+11 98.4 1.10E+08 0.01

Total 1.25E+10 20

Table 6.4 Radiological capacity of the site based on specific doses to members of the public via use of water from a borehole 1500 m from the site boundary for irrigation. Radiological capacity of the site is based on an illustrative waste inventory for the Meashill Trenches. Results include doses arising from ingrowth of daughter radionuclides for 100 years.


(µSv y-1 per MBq)

2010 Inventory

(MBq)


(MBq)


H-3 4.36E+25 3.25 6.08E+06 0.00

Co-60 7.91E+34 8090 1.51E+10 0.00

Cs-137 3.22E+21 952 1.78E+09 0.00

Ra-226 2.20E+09 99.6 1.86E+08 254.30

Th-232 8.36E+08 40 7.49E+07 268.72

U-234 7.72E+09 500 9.36E+08 363.85

U-235 7.75E+09 24 4.49E+07 17.38

U-238 7.71E+09 500 9.36E+08 364.15

Pu-238 1.80E+13 37 6.92E+07 0.01

Pu-239 2.18E+09 400 7.49E+08 1028.38

Pu-240 3.21E+09 400 7.49E+08 698.63

Pu-241 8.29E+10 38.2 7.15E+07 2.59

Am-241 2.77E+11 98.4 1.84E+08 2.00

Total 2.09E+10 3000

Table 6.5 Radiological capacity of the site based on specific doses to members of the public via use of water from a borehole at the site boundary for



drinking. Radiological capacity of the site is based on an illustrative waste inventory for the Meashill Trenches. Results include doses arising from ingrowth of daughter radionuclides for 100 years.


(µSv y-1 per MBq)

2010 Inventory

(MBq)


(MBq)


H-3 2.34E+13 3.25 1.61E+07 0.00

Co-60 7.16E+13 8090 4.01E+10 1.68

Cs-137 1.15E+11 952 4.72E+09 123.31

Ra-226 5.56E+08 99.6 4.94E+08 2666.92

Th-232 6.32E+09 40 1.98E+08 94.20

U-234 8.09E+11 500 2.48E+09 9.19

U-235 1.21E+11 24 1.19E+08 2.94

U-238 4.37E+11 500 2.48E+09 17.03

Pu-238 3.04E+11 37 1.83E+08 1.81

Pu-239 1.61E+11 400 1.98E+09 36.95

Pu-240 1.62E+11 400 1.98E+09 36.74

Pu-241 3.62E+11 38.2 1.89E+08 1.57

Am-241 1.91E+11 98.4 4.88E+08 7.67

Total 5.54E+10 3000

Table 6.6 Radiological capacity of the site based on specific doses to members of the public living on excavated waste after 60 years. Radiological capacity of the site is based on an illustrative waste inventory for the Meashill Trenches. Results include doses arising from ingrowth of daughter radionuclides for 60 years.

As noted above, the radiological capacities based on the Meashill Trenches data are illustrative. They do demonstrate that the ENRMF could use the whole of Cells 4B, 5A and 5B for the disposal of LLW at up to 200 Bq / g and remain, subject to an appropriate mixture of radionuclides, within an acceptable radiological capacity.



7 References

Allen, D.J., Brewerton, L.J., Coleby, L.M., Gibbs, B.R., Lewis, M.A., MacDonald, A.M., Wagstaff, S.J. and Williams, A.T., (1997). The physical properties of major aquifers in England and Wales. British Geological Survey Technical Report WD/97/34. Environment Agency R&D Publication 8.

Bullen Consultants Ltd (2005). Environmental Statement: Slipe Clay Pit Landfill Site. Report no. 104b028/Re01-Rev A/ arb.

Copplestone, D., Bielby, S., Jones, S.R., Patton, D., Daniel, P., and Gize, I. 2001. Impact assessment of ionising radiation on wildlife. Environment Agency R&D Publication 128. Environment Agency, Bristol.

Defra (2007) Policy for the Long Term Management of Solid Low Level Radioactive Waste in the United Kingdom. Department for Environment, Food and Rural Affairs, London.

Environment Agency (2006a) Initial Radiological Assessment Methodology – Part 1 User Report. Environment Agency Science Report, SC030162/SR Part 1.

Environment Agency (2006b) Initial Radiological Assessment Methodology – Part 2 Methods and Input Data. Environment Agency Science Report, SC030162/SR Part 2.

Environment Agency, Scottish Environment Protection Agency and Northern Ireland Environment Agency (2009). Near-Surface Disposal Facilities on Land for Solid Radioactive Wastes: Guidance on Requirements for Authorisation. Environment Agency, Bristol.

Environmental Simulations International Ltd. (ESI) (2004). Hydrogeological Risk Assessment and risk based monitoring scheme: King’s Cliffe Landfill. Report reference: 6490R3rev1.

HPA (Health Protection Agency) (2008). Guidance on the Application of Dose Coefficients for the Embryo, Fetus and Breastfed Infant in Dose Assessments for members of the Public. Health Protection Agency, Didcot.

IAEA (International Atomic Energy Agency). (2003). Derivation of Activity Limits for the Disposal of Radioactive Waste in Near Surface Disposal Facilities. IAEA-TECDOC-1380. ISBN 92-0-113003-1.

SNIFFER (2006a) Development of a Framework for Assessing the Suitability of Controlled Landfills to Accept Disposals of Solid Low-Level Radioactive Waste: Principles Document. SNIFFER, Edinburgh.

SNIFFER (2006b) Development of a Framework for Assessing the Suitability of Controlled Landfills to Accept Disposals of Solid Low-Level Radioactive Waste: Technical Reference Manual. SNIFFER, Edinburgh.



Appendix A Dose calculations

A.1 Doses during site operations

The air concentration of a radionuclide, CRn,gas,outdoors (Bq m-3), can be approximated by dividing by the air volume into which the activity released per year is diluted:

073.16E,

,,

huW

RC gasRn

outdoorsgasRn

where: RRn,gas is the release rate of radionuclide Rn in gas (Bq year–1) at the time of interest.

W is the width of the source perpendicular to the wind direction (m). u is the mean wind speed (m s–1). h is the height for vertical mixing (m). 3.16E+07 is the number of seconds in a year (s year–1).

The dose from gases other than radon is given by:

RninhoutoutdoorsgasRnoutdoorsgas DOBCDose ,,,

where: Oout is the time spent in the gas plume (years year–1). B is the breathing rate (m3 year-1). Dinh is the dose coefficient for inhalation of radionuclide Rn (Sv Bq-1).

The dose calculation for radon must account for the effect of the daughters of Rn-222 in the body, and has several additional terms:

21,, KOBKCDose outoutdoorsradonoutdoorsradon

where: K1 is the effective dose equivalent corresponding to an absorbed energy of 1 joule (Sv J-1).

ψ is the equilibrium factor (dimensionless). K2 is the potential α-energy of Rn-222 in equilibrium with its daughters

(J Bq-1).

Radionuclide-specific data are presented in Appendix 2. Other parameter values used in the calculation of specific doses for the ENRMF site are presented in Table A.1.



Parameter Description Value Units W width of the source perpendicular to

the wind direction 10 m

u mean wind speed 6.2 m s–1 h height for vertical mixing 2.0 m

worker 880 Oout time spent in the gas plume public 2192

hours year–1

worker 1.2 B breathing rate public 1.0

m3 hour-1

K1 effective dose equivalent corresponding to an absorbed energy of 1 joule

2.0 Sv J-1

ψ equilibrium factor 0.8 dimensionless K2 potential α-energy of Rn-222 in

equilibrium with its daughters 5.5x10-9 J Bq-1

Table A.1 Parameter values used in calculations of doses through the gas pathway during site operations.

A.2 Doses to site residents after closure

For calculation of peak dose, it is assumed that a house is constructed on top of the landfill cap immediately after closure. Irradiation doses are calculated for a resident spending 75% of the time indoors and 25% outdoors. Doses from gas inhalation are calculated for indoor exposure of the house resident to gas accumulating in the dwelling.

)t(x

wastewaste

waste,Rninout

Rnslab,irrirr

Rn

eV

)t(AsfOODDose

where: Oout is the time spent outside exposed to the waste (years year–1). Oin is the time spent inside (years year–1). sf is the shielding factor from the ground when indoors (dimensionless). ARn, waste(t) is the activity of the radionuclide, Rn (Bq), in the waste at time t Dirr,slab The dose conversion factor for irradiation from radionuclide Rn (Sv

year-1 Bq-1 kg), based on the receptor being 1 m from the ground and the contamination being spread out so as to approximate a semi-infinite slab.

Vwaste is the volume of material in which radioactivity is present (m3). ρwaste is the bulk density of the waste (kg m-3). Rn is the attenuation coefficient for radionuclide Rn (m-1). x is the thickness of the cap (m).

Doses from inhalation of radioactive gases (excluding radon) are calculated from:

Vka

atROBDDose H

gasRninRninhindoorsgas

1)(,,

where: B is the breathing rate (m3 year-1).



Oin is the occupancy of the house (years year-1). RRn,gas(t) is the release rate of gas at time t (Bq year-1). aH/a is the horizontal area of a dwelling divided by the area over which

the radioactive gas is being released (i.e., the facility footprint) (dimensionless).

k is a turnover rate to account for release of the gas by ventilation (year-1).

V is the volume of the house (m3). Dinh is the dose coefficient for inhalation of radionuclide Rn (Sv Bq-1).

As for the outdoor calculation, the dose calculation for radon must account for the effect of the daughters of Rn-222 in the body:

Vka

atROBKKDose H

radoninindoorsradon

1)(21,

where the terms are the same as those in the equations above.

Radionuclide-specific data are presented in Appendix 2. Other parameter values used in the calculation of specific doses for the ENRMF site are presented in Table A.3.

Parameter Description Value Units Vwaste volume of material in which

radioactivity is present 497,534 m3

ρwaste bulk density of the waste 700 kg m-3 x thickness of the cap 1.5 m B breathing rate 1.0 m3 hour-1 Oin occupancy of the house 6575 hours year-1 Oout time spent outside exposed to the waste 2192 hours year-1 sf shielding factor from the ground when

indoors 0.1 dimensionless

house 50 m2

aH/a horizontal area of a dwelling divided by the area over which the radioactive gas is being released (i.e., the facility footprint)

facility 3.4x104 m2

1.04x10-3 dimensionless

k is a turnover rate to account for release of the gas by ventilation

8.8x103 year-1

V volume of the house 130 m3

Table A.3 Parameter values used in calculations of doses to site residents after site closure.

A.3 Doses during and after excavation of waste

A.3.1 Dose to the Excavator

The excavator may receive a dose from irradiation, inhalation, and ingestion:



)()()( ,,,,excavator tCTMDtCTBMDtTCDDose wasteRningRningwasteRninh

RninhwasteRn

Rnslabirr

where Minh is the dust load of contaminated waste inhaled by the excavator (kg m-3).

Ming is the rate of ingestion of dust from the material (kg hour-1). T is the time the excavator is exposed to the material (hours year-1). B is the breathing rate (m3 hour-1). Dirr,slab, Dinh, and Ding are the dose coefficients for radionuclide Rn (Sv hour-1 Bq-1 kg;

Sv Bq-1; and Sv Bq-1, respectively). CRn,waste(t) is the concentration of radionuclide Rn (Bq kg-1) in the excavated

material at the time of excavation, t:

wastelandfill

RnwasteRn V

tAtC

)()(,

where ARn(t) is the activity of radionuclide Rn in the landfill at the time of excavation, t (Bq).

Vlandfill is the volume of landfill in which the activity ARn(t) is homogeneously distributed (m3).

ρwaste is the density of the waste (kg m-3).

The exposure from external irradiation is assumed to come from proximity to contaminated material, approximated by a semi-infinite slab.

The excavator might also receive a dose through direct contact with contaminated waste dust on hands and face:

TArea

AreaWDD

dtCDose

body

handsskin

Rnbeta

Rngamma

wastehandswasteRn

4074

,handsskin, 10

)(

where CRn,waste(t) is the concentration of radionuclide Rn (Bq kg-1) in the waste at the time of excavation, t.

Dgamma7 is the skin equivalent dose rate for radionuclide Rn to the basal layer of the skin epidermis for gamma irradiation (Sv h-1 per Bq cm-2).

Dbeta40 is the skin equivalent dose rate for radionuclide Rn to the basal layer of the skin epidermis for hands for beta irradiation, skin thickness 400 μm (40mg cm-2), (Sv h-1 per Bq cm-2).

104 converts Bq m-2 to Bq cm-2. dhands is the thickness of the contaminated layer on the hands (m). ρwaste is the density of the waste (kg m-3). Wskin is the tissue weighting factor for skin (dimensionless). Areahands is the area of skin in contact with the contaminated dust (cm2). Areabody is the total exposed skin area of the adult body (cm2). T is the time the worker is exposed to the material (hours year-1).

TArea

AreaWDD

dtCDose

body

faceskin

Rnbeta

Rngamma

wastefacewasteRn

474

,faceskin, 10

)(

where CRn,waste(t) is the concentration of radionuclide Rn (Bq kg-1) in the waste at the time of excavation, t.



Dgamma7 is the skin equivalent dose rate for radionuclide Rn to the basal layer of the skin epidermis for gamma irradiation (Sv h-1 per Bq cm-2).

Dbeta4 is the skin equivalent dose rate for radionuclide Rn to the basal layer of the skin epidermis for face for beta irradiation, skin thickness 40 μm (4 mg cm-2), (Sv h-1 per Bq cm-2).

104 converts Bq m-2 to Bq cm-2. dface is the thickness of the contaminated layer on the face (m). ρwaste is the density of the waste (kg m-3). Wskin is the tissue weighting factor for skin (dimensionless). Areaface is the area of skin in contact with the contaminated dust (cm2). Areabody is the total exposed skin area of the adult body (cm2). T is the time the worker is exposed to the material (hours year-1).

A.3.2 Dose to Site Resident after Excavation

It is assumed that following, or as part of the reason for, the excavation, the waste and the cover are mixed together and re-laid, creating a soil layer partly contaminated with the radioactivity that was in the waste. The initial concentration of radionuclide Rn in the material, CRn,soil,excavate (Bq kg-1), immediately after the excavation event is calculated by:

soillandfill

RnexcavatesoilRn V

DiltAC

)(

,,

where ARn(t) is the activity of radionuclide Rn in the landfill at the time of excavation, t (Bq).

Dil is the dilution factor given by the ratio of the volume of contaminated landfill waste to the volume of other material that is mixed in to form the soil (dimensionless).

Vlandfill is the volume of the landfill (m3). ρsoil is the density of the soil (kg m-3).

Dose from ingesting contaminated soil that may be attached to crops is given by:

RningexcavatesoilRnsoilsoiling DCQDose ,,,

where Qsoil is the soil consumption rate (kg year–1). Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).

Dose from crops grown on contaminated soil is given by:

crop

Rning

RncropexcavatesoilRncropcropsing DTFCQDose ,,,

where Qcrop is the crop consumption rate (kg year–1). TFcrop is the soil to crop transfer factor for radionuclide, Rn (Bq kg-1 fresh

weight of crop per Bq kg-1 of soil). Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).



Dose from livestock and associated products (e.g., milk) raised on contaminated ground is given by:

Rning

animal

Rnanimal

crop

RncropexcavatesoilRncropexcavatesoilRnsoilanimalanimaling DTFTFCqCqQDose

,,,,,

where Qanimal is the animal foodstuff consumption rate (kg year–1). qsoil is the soil consumption rate by the animal (kg day–1). qcrop is the crop consumption rate by the animal (kg day–1). TFcrop is the soil to crop transfer factor for radionuclide, Rn (Bq kg-1 fresh

weight of crop per Bq kg-1 of soil). TFanimal is the animal product transfer factor for radionuclide, Rn (days kg-1). Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).

Dose from external irradiation while living or working on contaminated soil is given by:

RnslabirrexcavatesoilRninoutsoilirr DCsfOODose ,,,,

where: Oout is the time spent outside exposed to the soil (years year–1). Oin is the time spent inside (years year–1). sf is the shielding factor from the ground when indoors (dimensionless). Dirr,slab is the dose conversion factor for irradiation from radionuclide Rn

(Sv year-1 Bq-1 kg), based on the receptor being 1 m from the ground and assuming a semi-infinite slab of contamination.

Dose from inhaling dust derived from contaminated soil is given by:

RninhexcavatesoilRndustsoilinh DdustloadCOBDose ,,,

where: Odust is the time spent exposed to dust from the soil (hours year–1). B is the breathing rate (m3 hour-1). dustload is the dust concentration (kg m-3 of air). Dinh is the dose coefficient for inhalation of radionuclide Rn (Sv Bq-1).

Radionuclide-specific data are presented in Appendix 2. Other parameter values used in the calculation of specific doses arising from excavation and intrusion into the ENRMF site are presented in Table A.4.



Parameter Description Value Units Minh dust load of contaminated waste

inhaled by the excavator 1.0x10-6 kg m-3

Ming rate of ingestion of dust from excavated material

3.45x10-5 kg hour-1

T time the excavator is exposed to excavated material

88 hours year-1

B breathing rate (worker) 1.2 m3 hour-1 Vlandfill volume of landfill in which the

activity is homogeneously distributed 497,534 m3

ρwaste density of the waste 1700 kg m-3 dhands thickness of the contaminated layer

on the hands 1.0x10-4 m

Wskin tissue weighting factor for skin 1x10-2 dimensionless Areahands area of skin in contact with the

contaminated dust 2x102 cm2

Areabody total exposed skin area of the adult body

3x103 cm2

dface thickness of the contaminated layer on the face

5x10-5 m

Areaface area of skin in contact with the contaminated dust

1x102 cm2

Dil dilution factor given by the ratio of the volume of contaminated landfill waste to the volume of other material that is mixed in to form the soil

0.3 dimensionless

Qsoil soil consumption rate 3.0x10-2 kg year–1 Grain 50 Green veg 30

Qcrop crop consumption rate

Root veg 120

kg year–1

Meat 32 Qanimal animal foodstuff consumption rate

Milk 100

kg year–1

qsoil soil consumption rate by the animal 0.6 kg day–1 qcrop crop consumption rate by the animal 55 kg day–1 Oout time spent outside exposed to the soil 0.25 years year–1 Oin time spent inside 0.75 years year–1 sf shielding factor from the ground

when indoors 0.1 dimensionless

Odust time spent exposed to dust from soil 2.2x103 hours year–1 B Breathing rate (public) 1.0 m3 hour-1 dustload dust concentration 1x10-7 kg m-3 of air

Table A.4 Parameter values used in calculations of doses during and after excavation of waste. Consumption rates for animal foodstuffs and grain are about one-third of the average rates cited in IAEA (2003). Consumption rates for vegetables are about one-half of the average rates cited in IAEA (2003).



A.4 Doses arising from use of contaminated groundwater

If a well or river is used for irrigation, then doses can result from ingestion of foodstuffs raised on contaminated soil, inhalation of dust from the soil, and external exposure to the soil. Drinking of contaminated water from a well or river is also a potential exposure pathway. If contaminated groundwater discharges to surface water (spring, river, sea), then ingestion of foodstuffs from the surface water is a potential exposure pathway.

Dose from ingesting contaminated soil that may be attached to crops is given by:

RningsoilRnsoilsoiling DtCQDose )(,,

where Qsoil is the soil consumption rate (kg year–1). CRn,soil(t) is the concentration of radionuclide in the soil at time t (Bq kg-1). Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).

Dose from crops grown on contaminated soil is given by:

crop

Rning

RncropsoilRn

crop

cropcropratewaterRncropcropsing DTFtC

Yield

FIntIrrigtCQDose )()( ,,,

where Qcrop is the crop consumption rate (kg year–1). Irrigrate is the rate of irrigation (m year-1). Intcrop is the effective interception factor (dimensionless). Fcrop is the fraction remaining after processing (dimensionless). Yieldcrop is the crop yield (kg m-2). TFcrop is the soil to crop transfer factor for radionuclide, Rn (Bq kg-1 fresh

weight of crop per Bq kg-1 of soil). CRn,water(t) is the concentration of radionuclide in the water used for irrigation at

time t (Bq m-3). CRn,soil(t) is the concentration of radionuclide in the crop soil at time t

(Bq kg-1). Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).

Dose from livestock and associated products (e.g., milk) raised on contaminated ground and fed with contaminated crops is given by:

Rning

animal

Rnanimal

crop

RncropsoilRncropsoilRnsoilanimalanimaling DTFTFtCqtCqQDose

)()( ,,,

where Qanimal is the animal foodstuff consumption rate (kg year–1). qwater is the water consumption rate by the animal (m3 day–1). qsoil is the soil consumption rate by the animal (kg day–1). qcrop is the crop consumption rate by the animal (kg day–1). TFcrop is the soil to crop transfer factor for radionuclide Rn (Bq kg-1 fresh

weight of crop per Bq kg-1 of soil). TFanimal is the animal product transfer factor for radionuclide Rn (days kg-1).



CRn,soil(t) is the concentration of radionuclide Rn in the pasture and crop soil at time t (Bq kg-1).

Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq-1).

Dose from external irradiation while living or working on contaminated soil is given by:

RnslabirrsoilRninoutsoilirr DtCsfOODose ,,, )(

where: Oout is the time spent outside exposed to the soil (years year–1). Oin is the time spent inside (years year–1). sf is the shielding factor from the ground when indoors (dimensionless). CRn,soil(t) is the concentration of radionuclide in the soil at time t (Bq kg-1). Dirr,slab is the dose conversion factor for irradiation from radionuclide Rn

(Sv year-1 Bq-1 kg), based on the receptor being 1 m from the ground and assuming a semi-infinite slab of contamination.

Dose from inhaling dust derived from contaminated soil is given by:

RninhsoilRndustsoilinh DdustloadtCOBDose )(,,

where: Odust is the time spent exposed to dust from the soil (years year–1). B is the breathing rate (m3 year-1). CRn,soil(t) is the concentration of radionuclide in the soil at time t (Bq kg-1). dustload is the dust concentration (kg m-3 of air). Dinh is the dose coefficient for inhalation of radionuclide Rn (Sv Bq-1).

Parameter Description Value Units Irrigrate rate of irrigation 0.3 m year-1 Intcrop effective interception factor 0.33 dimensionless

Grain 1.0

Green veg 0.3

Fcrop fraction remaining after processing

Root veg 1.0

dimensionless

Pasture 1.7 Grain 0.4 Green veg 3.0

Yieldcrop crop yield

Root veg 3.5

kg m-2

Table A.5 Parameter values used in calculations of doses arising from use of contaminated groundwater for irrigation.



Appendix B Radionuclide-specific data

Radionuclide Half-life DC inhalation

DC ingestion

Irradiation slab

Dgamma7 Dbeta4 Dbeta40

name y (Sv Bq-1) (Sv Bq-1) (Sv y-1 Bq-1 kg) (Sv h-1 Bq-1 cm2) (Sv h-1 Bq-1 cm2)

(Sv h-1 Bq-1 cm2)

H-3 1.23E+01 2.60E-10 1.80E-11 0.00E+00 0.00E+00 0.00E+00 0.00E+00

C-14 5.73E+03 5.80E-09 5.80E-10 3.64E-12 0.00E+00 9.02E-07 0.00E+00

Cl-36 3.01E+05 7.30E-09 9.30E-10 6.46E-10 1.10E-11 2.51E-06 5.37E-07

Fe-55 2.70E+00 7.70E-10 3.30E-10 0.00E+00 1.60E-08 0.00E+00 0.00E+00

Co-60 5.27E+00 3.10E-08 3.40E-09 4.38E-06 1.30E-07 1.83E-06 2.85E-08

Ni-63 9.60E+01 4.80E-10 1.50E-10 0.00E+00 0.00E+00 1.83E-08 0.00E+00

Sr-90 2.91E+01 1.62E-07 3.07E-08 6.65E-09 2.40E-12 5.14E-06 1.76E-06

Nb-94 2.00E+04 1.10E-08 1.70E-09 2.62E-06 9.47E-08 2.17E-06 1.83E-07

Tc-99 2.13E+05 1.30E-08 6.40E-10 3.39E-11 3.49E-14 1.60E-06 1.37E-08

Ru-106 1.01E+00 6.60E-08 7.00E-09 3.49E-07 1.20E-08 2.85E-06 1.60E-06

Ag-108m 1.27E+02 3.70E-08 2.30E-09 2.61E-06 1.28E-07 2.76E-07 1.15E-07

Sb-125 2.80E+00 5.46E-09 3.11E-09 6.61E-07 3.54E-08 2.05E-06 8.45E-08

Sn-126 1.00E+05 3.12E-08 7.10E-09 4.66E-06 1.33E-07 4.54E-06 1.43E-06

I-129 1.57E+07 3.60E-08 1.10E-07 3.50E-09 9.70E-09 6.51E-07 0.00E+00

Ba-133 1.07E+01 3.10E-09 1.50E-09 5.32E-07 0.00E+00 0.00E+00 0.00E+00

Cs-134 2.10E+00 6.80E-09 1.90E-08 2.56E-06 8.79E-08 1.83E-06 3.08E-07

Cs-137 3.00E+01 3.90E-08 1.30E-08 9.75E-07 3.30E-08 2.54E-06 3.90E-07

Pm-147 2.60E+00 5.00E-09 2.60E-10 1.35E-11 4.91E-13 1.26E-06 4.11E-10

Eu-152 1.33E+01 4.20E-08 1.40E-09 1.89E-06 1.18E-07 1.60E-06 1.71E-07

Eu-154 8.80E+00 5.30E-08 2.00E-09 2.08E-06 9.02E-08 3.42E-06 3.77E-07

Eu-155 4.96E+00 6.90E-09 3.20E-10 4.93E-08 1.77E-08 8.68E-07 3.20E-10

Pb-210 2.23E+01 9.99E-06 1.89E-06 1.65E-09 8.30E-09 2.63E-06 8.45E-07

Ra-226 1.60E+03 1.95E-05 2.17E-06 3.02E-06 1.64E-07 5.89E-06 1.64E-06

Ac-227 2.18E+01 5.69E-04 1.21E-06 5.43E-07 3.81E-08 6.59E-06 2.00E-06

Th-229 7.34E+03 2.56E-04 6.13E-07 4.33E-07 7.31E-08 8.56E-06 1.36E-06

Th-230 7.70E+04 1.00E-04 2.10E-07 3.27E-10 3.83E-09 1.04E-07 0.00E+00

Th-232 1.40E+10 1.70E-04 1.06E-06 4.37E-06 2.20E-09 3.08E-08 0.00E+00

Pa-231 3.27E+04 1.40E-04 7.10E-07 5.15E-08 6.27E-08 1.48E-07 5.14E-09

U-232 6.89E+01 4.69E-05 4.60E-07 2.44E-10 9.36E-08 3.20E-08 0.00E+00

U-233 1.58E+05 9.60E-06 5.10E-08 3.78E-10 1.70E-09 5.25E-07 0.00E+00

U-234 2.44E+05 9.40E-06 4.90E-08 1.09E-10 2.70E-09 7.42E-09 0.00E+00

U-235 7.04E+08 8.50E-06 4.73E-08 2.05E-07 5.31E-08 2.52E-06 1.09E-08

U-236 2.34E+07 3.20E-06 4.70E-08 5.78E-11 3.55E-09 4.57E-09 0.00E+00

U-238 4.47E+09 8.01E-06 4.84E-08 3.58E-08 9.23E-09 3.82E-06 1.26E-06

Np-237 2.14E+06 5.00E-05 1.11E-07 2.97E-07 3.20E-08 3.46E-06 9.93E-08

Pu-238 8.77E+01 1.10E-04 2.30E-07 4.09E-11 2.70E-09 1.06E-07 0.00E+00

Pu-239 2.41E+04 1.20E-04 2.50E-07 7.98E-11 1.00E-09 4.34E-10 0.00E+00

Pu-240 6.54E+03 1.20E-04 2.50E-07 3.96E-11 2.60E-09 0.00E+00 0.00E+00

Pu-241 1.44E+01 2.30E-06 4.80E-09 1.60E-12 3.30E-12 0.00E+00 0.00E+00

Pu-242 3.76E+05 1.10E-04 2.40E-07 3.46E-11 3.07E-09 0.00E+00 0.00E+00

Am-241 4.32E+02 9.60E-05 2.00E-07 1.18E-08 1.70E-08 5.48E-08 0.00E+00

Cm-243 2.91E+01 3.11E-05 1.50E-07 1.58E-07 7.99E-09 1.94E-06 3.42E-08

Cm-244 1.81E+01 2.71E-05 1.21E-07 3.41E-11 2.17E-09 0.00E+00 0.00E+00



Radionuclide UF freshwater fish

TF cow meat TF cow milk UF green veg WR green veg UF root veg UF grain UF grass

name (m3 kg-1) (d kg-1 fresh weight)

(d kg-1) (Bq kg-1Bq-1 kg) (y-1) (Bq kg-1Bq-1 kg)

(Bq kg-1Bq-1 kg)

(Bq kg-1Bq-1 kg)

H-3 1.00E-03 2.90E-02 1.00E-02 5.00E+00 1.83E+01 5.00E+00 1.00E-02 5.00E+00

C-14 9.00E+00 1.20E-01 1.00E-02 1.00E-01 1.83E+01 1.00E-01 1.60E-01 1.00E-01

Cl-36 5.00E-02 4.30E-02 1.70E-02 5.00E+00 1.83E+01 5.00E+00 8.80E-02 5.00E+00

Fe-55 1.00E-01 2.00E-02 3.00E-05 2.00E-04 1.83E+01 3.00E-04 1.00E-01 4.00E-04

Co-60 3.00E-01 1.00E-02 3.00E-04 3.00E-02 1.83E+01 3.00E-02 8.00E-02 6.00E-03

Ni-63 1.00E-01 5.00E-03 1.60E-02 3.00E-02 1.83E+01 3.00E-02 5.00E-02 2.00E-02

Sr-90 6.00E-02 8.00E-03 3.00E-03 3.00E+00 1.83E+01 9.00E-02 1.20E-01 3.00E+00

Nb-94 3.00E-01 3.00E-07 4.10E-07 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02

Tc-99 2.00E-02 1.00E-04 2.30E-05 1.00E+01 1.83E+01 1.00E+01 1.00E+01 1.00E+01

Ru-106 1.00E-02 5.00E-02 3.30E-06 4.00E-03 1.83E+01 1.00E-02 1.00E-01 4.00E-02

Ag-108m 5.00E-03 3.00E-05 5.00E-05 2.70E-04 1.83E+01 1.30E-03 8.80E-02 1.50E-01

Sb-125 1.00E-01 4.00E-05 2.50E-05 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02

Sn-126 1.00E+00 1.90E-03 1.00E-03 1.00E-01 1.83E+01 1.00E-01 2.00E-01 2.00E-01

I-129 3.00E-02 4.00E-02 1.00E-02 1.00E-01 1.83E+01 1.00E-01 2.80E-01 1.00E-01

Ba-133 4.00E-03 5.00E-04 5.00E-04 4.00E-03 1.83E+01 1.00E-02 1.00E-01 4.00E-02

Cs-134 2.00E+00 5.00E-02 7.90E-03 3.00E-02 1.83E+01 3.00E-02 2.00E-02 3.00E-02

Cs-137 2.00E+00 5.00E-02 7.90E-03 3.00E-02 1.83E+01 3.00E-02 2.00E-02 3.00E-02

Pm-147 3.00E-02 5.00E-03 2.00E-05 3.00E-03 1.83E+01 3.00E-03 3.00E-03 3.00E-03

Eu-152 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03

Eu-154 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03

Eu-155 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03

Pb-210 3.00E-01 4.00E-04 3.00E-04 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02

Ra-226 5.00E-02 9.00E-04 1.30E-03 4.00E-02 1.83E+01 4.00E-02 4.00E-02 4.00E-02

Ac-227 8.00E-01 1.60E-04 4.00E-07 1.00E-03 1.83E+01 1.00E-03 1.00E-03 1.00E-03

Th-229 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04

Th-230 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04

Th-232 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04

Pa-231 1.00E-02 5.00E-05 5.00E-06 4.00E-02 1.83E+01 4.00E-02 4.00E-02 4.00E-02

U-232 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

U-233 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

U-234 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

U-235 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

U-236 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

U-238 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03

Np-237 1.00E-02 1.00E-03 5.00E-06 1.00E-02 5.11E+01 1.00E-03 3.00E-04 5.00E-03

Pu-238 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03

Pu-239 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03

Pu-240 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03

Pu-241 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03

Pu-242 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03

Am-241 3.00E-02 4.00E-05 1.50E-06 1.00E-03 1.83E+01 1.00E-03 1.00E-05 5.00E-03

Cm-243 3.00E-02 1.00E-04 1.00E-06 1.00E-04 1.83E+01 1.00E-03 3.00E-05 1.00E-03

Cm-244 3.00E-02 1.00E-04 1.00E-06 1.00E-04 1.83E+01 1.00E-03 3.00E-05 1.00E-03



Radionuclide Kd soil Kd barrier Irradiation cloudshine

Irradiation groundshine Attenuation coefficient

name (m3 kg-1) (m3 kg-1) (Sv hr-1 Bq-1 m3) (Sv y-1 Bq-1 m2) (m-1)

H-3 1.00E-04 1.00E-04 1.19E-15 0.00E+00 0.00E+00

C-14 1.00E-01 1.00E-01 8.08E-16 5.08E-13 -5.59E+01

Cl-36 1.50E-02 1.50E-02 8.03E-14 2.12E-11 -2.04E+01

Fe-55 2.20E-01 8.00E-01 0.00E+00 0.00E+00 0.00E+00

Co-60 6.00E-02 1.00E+01 4.55E-10 7.43E-08 -1.20E+01

Ni-63 4.00E-01 6.00E-01 0.00E+00 0.00E+00 0.00E+00

Sr-90 1.30E-02 1.40E-01 7.12E-13 1.77E-10 -2.88E+01

Nb-94 1.60E-01 7.60E+00 2.77E-10 4.84E-08 -1.38E+01

Tc-99 1.40E-04 1.90E-01 5.83E-15 2.46E-12 -3.85E+01

Ru-106 5.50E-02 4.00E-01 3.73E-11 6.69E-09 -1.40E+01

Ag-108m 9.00E-02 1.80E-01 2.81E-10 5.03E-08 -1.49E+01

Sb-125 4.50E-02 2.40E-01 7.27E-11 1.34E-08 -1.50E+01

Sn-126 1.30E-01 6.70E-01 5.03E-10 8.94E-08 -3.59E+01

I-129 1.00E-03 1.00E-03 1.37E-12 8.15E-10 -1.31E+02

Ba-133 4.10E-03 4.00E-02 6.42E-11 1.25E-08 -1.40E+01

Cs-134 2.70E-01 2.00E+00 2.72E-10 4.81E-08 -1.40E+01

Cs-137 2.70E-01 2.00E+00 1.04E-10 1.85E-08 -2.61E+01

Pm-147 2.40E-01 1.30E+00 2.49E-15 1.08E-12 -2.78E+01

Eu-152 2.40E-01 7.80E+00 2.03E-10 3.48E-08 -1.30E+01

Eu-154 2.40E-01 7.80E+00 2.21E-10 3.75E-08 -1.29E+01

Eu-155 2.40E-01 7.80E+00 8.96E-12 1.86E-09 -3.37E+01

Pb-210 2.70E-01 4.90E+00 3.23E-13 1.12E-10 -8.36E+01

Ra-226 4.90E-01 9.00E+00 3.20E-10 5.26E-08 -2.29E+01

Ac-227 4.50E-01 5.00E+00 7.50E-11 1.40E-08 -2.75E+01

Th-229 3.00E+00 1.43E+01 4.15E-10 6.87E-08 -2.65E+01

Th-230 3.00E+00 1.43E+01 6.28E-14 2.37E-11 -2.82E+01

Th-232 3.00E+00 1.43E+01 8.70E-10 1.34E-07 -2.93E+01

Pa-231 5.40E-01 1.00E+01 8.78E-12 1.29E-09 -1.91E+01

U-232 3.30E-02 6.00E+00 5.10E-14 3.19E-11 -6.00E+01

U-233 3.30E-02 6.00E+00 5.87E-14 2.26E-11 -3.74E+01

U-234 3.30E-02 6.00E+00 2.74E-14 2.36E-11 -2.29E+01

U-235 3.30E-02 6.00E+00 2.78E-11 5.25E-09 -3.58E+01

U-236 3.30E-02 6.00E+00 1.80E-14 2.05E-11 -2.32E+01

U-238 3.30E-02 6.00E+00 3.40E-10 5.89E-08 -3.16E+01

Np-237 4.10E-03 4.60E-02 3.74E-11 7.07E-09 -8.74E+01

Pu-238 5.40E-01 7.60E+00 1.75E-14 2.64E-11 -2.18E+01

Pu-239 5.40E-01 7.60E+00 1.53E-14 1.16E-11 -5.37E+01

Pu-240 5.40E-01 7.60E+00 1.71E-14 2.54E-11 -2.96E+01

Pu-241 5.40E-01 7.60E+00 2.61E-16 6.09E-14 -5.58E+01

Pu-242 5.40E-01 7.60E+00 1.44E-14 2.10E-11 -4.44E+01

Am-241 2.00E+00 3.20E+00 2.95E-12 8.67E-10 -3.73E+01

Cm-243 4.00E-01 4.00E+00 2.12E-11 3.95E-09 -1.66E+01

Cm-244 4.00E-01 4.00E+00 1.77E-14 2.77E-11 -2.89E+01



Appendix C Sensitivity Studies

C.1 Groundwater Pathway

This section presents the results from a set of sensitivity studies to assess how calculated doses via the groundwater pathway are affected by alternative assumptions about the disposal system. Results are presented that show the effect of variations in leachate head within the landfill, of changes to the lifetime and efficiency of the cap, in different assessment periods, and in the assumptions about the exposed group.

The results of these sensitivity studies are discussed in Section 5.1.


Radionuclide

1m head 2m head 5m head 10m head Increase 1m - 10m

H-3 5.99E-23 6.01E-23 6.07E-23 6.17E-23 1.03

C-14 1.47E-07 1.47E-07 1.47E-07 1.47E-07 1.00

Cl-36 1.60E-06 1.60E-06 1.60E-06 1.60E-06 1.00

Fe-55 1.90E-36 6.63E-36 3.45E-35 1.20E-34 63.08

Co-60 2.92E-32 3.00E-32 3.63E-32 1.26E-31 4.33

Ni-63 2.92E-15 2.92E-15 2.92E-15 2.92E-15 1.00

Sr-90 1.70E-17 1.70E-17 1.71E-17 1.71E-17 1.01

Nb-94 6.84E-09 6.84E-09 6.84E-09 6.84E-09 1.00

Tc-99 1.15E-07 1.15E-07 1.15E-07 1.15E-07 1.00

Ru-106 7.67E-39 2.67E-38 1.39E-37 4.84E-37 63.09

Ag-108m 1.11E-12 1.11E-12 1.11E-12 1.11E-12 1.00

Sb-125 8.25E-35 2.87E-34 1.49E-33 5.20E-33 63.08

Sn-126 3.72E-07 3.72E-07 3.72E-07 3.72E-07 1.00

I-129 2.72E-04 2.72E-04 2.72E-04 2.72E-04 1.00

Ba-133 2.45E-24 2.46E-24 2.49E-24 2.53E-24 1.03

Cs-134 3.95E-36 1.37E-35 7.15E-35 2.49E-34 63.08

Cs-137 8.29E-19 8.29E-19 8.31E-19 8.34E-19 1.01

Pm-147 6.04E-37 2.10E-36 1.09E-35 3.81E-35 63.08

Eu-152 3.63E-25 3.64E-25 3.66E-25 3.71E-25 1.02

Eu-154 6.00E-28 6.03E-28 6.13E-28 6.29E-28 1.05

Eu-155 9.68E-34 1.00E-33 2.18E-33 7.58E-33 7.83

Pb-210 1.07E-18 1.07E-18 1.08E-18 1.08E-18 1.01

Ra-226 1.04E-06 1.04E-06 1.04E-06 1.04E-06 1.00




Radionuclide

1m head 2m head 5m head 10m head Increase 1m - 10m

Ac-227 5.00E-19 5.01E-19 5.02E-19 5.05E-19 1.01

Th-229 9.80E-07 9.80E-07 9.80E-07 9.80E-07 1.00

Th-230 5.15E-07 5.15E-07 5.15E-07 5.15E-07 1.00

Th-232 2.73E-06 2.73E-06 2.73E-06 2.73E-06 1.00

Pa-231 2.32E-06 2.32E-06 2.32E-06 2.32E-06 1.00

U-232 4.58E-14 4.58E-14 4.59E-14 4.59E-14 1.00

U-233 3.04E-07 3.04E-07 3.04E-07 3.04E-07 1.00

U-234 2.94E-07 2.94E-07 2.94E-07 2.94E-07 1.00

U-235 2.88E-07 2.88E-07 2.88E-07 2.88E-07 1.00

U-236 2.86E-07 2.86E-07 2.86E-07 2.86E-07 1.00

U-238 2.94E-07 2.94E-07 2.94E-07 2.95E-07 1.00

Np-237 7.22E-05 7.22E-05 7.22E-05 7.22E-05 1.00

Pu-238 1.47E-13 1.47E-13 1.47E-13 1.47E-13 1.00

Pu-239 1.04E-06 1.04E-06 1.04E-06 1.04E-06 1.00

Pu-240 7.04E-07 7.04E-07 7.04E-07 7.04E-07 1.00

Pu-241 4.19E-24 4.20E-24 4.23E-24 4.27E-24 1.02

Pu-242 1.14E-06 1.14E-06 1.14E-06 1.14E-06 1.00

Am-241 5.52E-09 5.52E-09 5.52E-09 5.52E-09 1.00

Cm-243 2.79E-18 2.79E-18 2.80E-18 2.81E-18 1.01

Cm-244 5.23E-21 5.24E-21 5.26E-21 5.30E-21 1.01

Table C.1 Sensitivity studies on the effect of increased leachate head in the landfill. Specific doses are to members of the public via use of water from a borehole at the site boundary for drinking. Results do not include the effects of ingrowth of daughter radionuclides.

Specific dose (µSv y-1 per MBq) Radionuclide

20 y cap 60 y cap 100 y cap

Difference 20 - 100 y

H-3 5.56E-22 5.99E-23 5.63E-24 98.76

C-14 1.49E-07 1.47E-07 1.45E-07 1.03

Cl-36 1.61E-06 1.60E-06 1.59E-06 1.01

Fe-55 1.06E-34 1.90E-36 4.80E-36 22.04

Co-60 5.26E-30 2.92E-32 2.99E-33 1758.56






Ni-63 3.88E-15 2.92E-15 2.16E-15 1.80

Sr-90 4.37E-17 1.70E-17 6.26E-18 6.97

Nb-94 6.93E-09 6.84E-09 6.74E-09 1.03

Tc-99 1.16E-07 1.15E-07 1.13E-07 1.03

Ru-106 1.70E-39 7.67E-39 3.97E-38 0.04

Ag-108m 1.38E-12 1.11E-12 8.85E-13 1.56

Sb-125 5.99E-33 8.25E-35 2.03E-34 29.57

Sn-126 3.77E-07 3.72E-07 3.67E-07 1.03

I-129 2.72E-04 2.72E-04 2.72E-04 1.00

Ba-133 3.18E-23 2.45E-24 1.62E-25 196.36

Cs-134 2.80E-35 3.95E-36 1.20E-35 2.33

Cs-137 2.07E-18 8.29E-19 3.14E-19 6.58

Pm-147 2.52E-35 6.04E-37 1.56E-36 16.13

Eu-152 2.85E-24 3.63E-25 4.07E-26 70.00

Eu-154 1.35E-26 6.00E-28 2.21E-29 613.41

Eu-155 2.40E-31 9.68E-34 1.89E-34 1275.31

Pb-210 3.67E-18 1.07E-18 2.91E-19 12.61

Ra-226 1.05E-06 1.04E-06 1.02E-06 1.04

Ac-227 1.76E-18 5.00E-19 1.32E-19 13.36

Th-229 9.93E-07 9.80E-07 9.66E-07 1.03

Th-230 5.22E-07 5.15E-07 5.07E-07 1.03

Th-232 2.76E-06 2.73E-06 2.69E-06 1.03

Pa-231 2.36E-06 2.32E-06 2.29E-06 1.03

U-232 6.82E-14 4.58E-14 3.00E-14 2.27

U-233 3.08E-07 3.04E-07 2.99E-07 1.03

U-234 2.98E-07 2.94E-07 2.90E-07 1.03

U-235 2.92E-07 2.88E-07 2.84E-07 1.03

U-236 2.90E-07 2.86E-07 2.82E-07 1.03

U-238 2.99E-07 2.94E-07 2.90E-07 1.03

Np-237 7.30E-05 7.22E-05 7.13E-05 1.02

Pu-238 2.01E-13 1.47E-13 1.06E-13 1.91

Pu-239 1.05E-06 1.04E-06 1.02E-06 1.03

Pu-240 7.13E-07 7.04E-07 6.94E-07 1.03

Pu-241 2.82E-23 4.19E-24 5.57E-25 50.61






Pu-242 1.15E-06 1.14E-06 1.12E-06 1.03

Am-241 5.88E-09 5.52E-09 5.16E-09 1.14

Cm-243 7.16E-18 2.79E-18 1.03E-18 6.97

Cm-244 2.38E-20 5.23E-21 1.05E-21 22.70

Table C.2 Sensitivity studies on the effect of changes in the cap lifetime. Specific doses are to members of the public via use of water from a borehole 100 m from the site boundary for drinking. Results do not include the effects of ingrowth of daughter radionuclides.


Radionuclide

99% 95% 90% Difference 99 - 90 %

H-3 5.99E-23 6.94E-23 8.31E-23 1.39

C-14 1.47E-07 1.47E-07 1.47E-07 1.00

Cl-36 1.60E-06 1.6E-06 1.6E-06 1.00

Fe-55 1.90E-36 1.9E-36 1.9E-36 1.00

Co-60 2.92E-32 4.5E-32 7.16E-32 2.45

Ni-63 2.92E-15 2.96E-15 3.02E-15 1.03

Sr-90 1.70E-17 1.79E-17 1.91E-17 1.12

Nb-94 6.84E-09 6.84E-09 6.85E-09 1.00

Tc-99 1.15E-07 1.15E-07 1.15E-07 1.00

Ru-106 7.67E-39 7.67E-39 7.67E-39 1.00

Ag-108m 1.11E-12 1.12E-12 1.14E-12 1.03

Sb-125 8.25E-35 8.25E-35 8.25E-35 1.00

Sn-126 3.72E-07 3.73E-07 3.73E-07 1.00

I-129 2.72E-04 0.000272 0.000272 1.00

Ba-133 2.45E-24 2.89E-24 3.53E-24 1.44

Cs-134 3.95E-36 3.95E-36 3.95E-36 1.00

Cs-137 8.29E-19 8.72E-19 9.29E-19 1.12

Pm-147 6.04E-37 6.04E-37 6.04E-37 1.00

Eu-152 3.63E-25 4.12E-25 4.82E-25 1.33




Radionuclide

99% 95% 90% Difference 99 - 90 %

Eu-154 6.00E-28 7.39E-28 9.43E-28 1.57

Eu-155 9.68E-34 1.53E-33 2.5E-33 2.59

Pb-210 1.07E-18 1.15E-18 1.26E-18 1.17

Ra-226 1.04E-06 1.04E-06 1.04E-06 1.00

Ac-227 5.00E-19 5.37E-19 5.88E-19 1.18

Th-229 9.80E-07 9.8E-07 9.81E-07 1.00

Th-230 5.15E-07 5.15E-07 5.15E-07 1.00

Th-232 2.73E-06 2.73E-06 2.73E-06 1.00

Pa-231 2.32E-06 2.32E-06 2.33E-06 1.00

U-232 4.58E-14 4.68E-14 4.8E-14 1.05

U-233 3.04E-07 3.04E-07 3.04E-07 1.00

U-234 2.94E-07 2.94E-07 2.94E-07 1.00

U-235 2.88E-07 2.88E-07 2.88E-07 1.00

U-236 2.86E-07 2.86E-07 2.86E-07 1.00

U-238 2.94E-07 2.95E-07 2.95E-07 1.00

Np-237 7.22E-05 7.22E-05 7.22E-05 1.00

Pu-238 1.47E-13 1.49E-13 1.53E-13 1.04

Pu-239 1.04E-06 1.04E-06 1.04E-06 1.00

Pu-240 7.04E-07 7.04E-07 7.05E-07 1.00

Pu-241 4.19E-24 4.71E-24 5.44E-24 1.30

Pu-242 1.14E-06 1.14E-06 1.14E-06 1.00

Am-241 5.52E-09 5.53E-09 5.56E-09 1.01

Cm-243 2.79E-18 2.94E-18 3.14E-18 1.12

Cm-244 5.23E-21 5.72E-21 6.39E-21 1.22

Table C.3 Sensitivity studies on the effect of differences in cap efficiency. Cap lifetime is assumed to be 60 years. Specific doses are to members of the public via use of water from a borehole 100 m from the site boundary for drinking. Results do not include the effects of ingrowth of daughter radionuclides.




Radionuclide

5,000 y 1,000 y 500 y Difference

5,000 - 500 y

H-3 5.99E-23 5.99E-23 5.99E-23 1.0

C-14 1.47E-07 9.76E-11 9.90E-14 1.5E+06

Cl-36 1.60E-06 1.01E-09 9.92E-13 1.6E+06

Fe-55 1.90E-36 1.90E-36 1.90E-36 1.0

Co-60 2.92E-32 2.92E-32 2.92E-32 1.0

Ni-63 2.92E-15 2.72E-15 9.58E-17 3.0E+01

Sr-90 1.70E-17 1.70E-17 1.70E-17 1.00

Nb-94 6.84E-09 2.96E-12 2.86E-15 2.4E+06

Tc-99 1.15E-07 4.58E-11 4.37E-14 2.6E+06

Ru-106 7.67E-39 7.67E-39 7.67E-39 1.0

Ag-108m 1.11E-12 7.16E-13 1.04E-14 1.1E+02

Sb-125 8.25E-35 8.25E-35 8.25E-35 1.0

Sn-126 3.72E-07 1.46E-10 1.39E-13 2.7E+06

I-129 2.72E-04 1.03E-06 1.33E-09 2.0E+05

Ba-133 2.45E-24 2.45E-24 2.45E-24 1.0

Cs-134 3.95E-36 3.95E-36 3.95E-36 1.0

Cs-137 8.29E-19 8.29E-19 8.29E-19 1.0

Pm-147 6.04E-37 6.04E-37 6.04E-37 1.0

Eu-152 3.63E-25 3.63E-25 3.63E-25 1.0

Eu-154 6.00E-28 6.00E-28 6.00E-28 1.0

Eu-155 9.68E-34 9.68E-34 9.68E-34 1.0

Pb-210 1.07E-18 1.07E-18 1.07E-18 1.0

Ra-226 1.04E-06 2.16E-09 2.55E-12 4.1E+05

Ac-227 5.00E-19 5.00E-19 5.00E-19 1.0

Th-229 9.80E-07 5.38E-10 5.36E-13 1.8E+06

Th-230 5.15E-07 2.01E-10 1.92E-13 2.7E+06

Th-232 2.73E-06 1.03E-09 9.75E-13 2.8E+06

Pa-231 2.32E-06 9.52E-10 9.15E-13 2.5E+06

U-232 4.58E-14 4.58E-14 6.11E-15 7.50

U-233 3.04E-07 1.16E-10 1.11E-13 2.7E+06

U-234 2.94E-07 1.12E-10 1.07E-13 2.8E+06

U-235 2.88E-07 1.08E-10 1.03E-13 2.8E+06




Radionuclide

5,000 y 1,000 y 500 y Difference

5,000 - 500 y

U-236 2.86E-07 1.08E-10 1.02E-13 2.8E+06

U-238 2.94E-07 1.11E-10 1.05E-13 2.8E+06

Np-237 7.22E-05 3.26E-08 3.13E-11 2.3E+06

Pu-238 1.47E-13 1.45E-13 7.17E-15 2.1E+01

Pu-239 1.04E-06 4.38E-10 4.22E-13 2.5E+06

Pu-240 7.04E-07 4.05E-10 4.06E-13 1.7E+06

Pu-241 4.19E-24 4.19E-24 4.19E-24 1.0

Pu-242 1.14E-06 4.32E-10 4.11E-13 2.8E+06

Am-241 5.52E-09 1.70E-10 3.61E-13 1.5E+04

Cm-243 2.79E-18 2.79E-18 2.79E-18 1.0

Cm-244 5.23E-21 5.23E-21 5.23E-21 1.0

Table C.4 Sensitivity studies on the effect of changes in the assessment period. Specific doses are to members of the public via use of water from a borehole 100 m from the site boundary for drinking. Results do not include the effects of ingrowth of daughter radionuclides.


Adult Infant Child

Infant - Adult

H-3 3.66E-30 2.90E-30 2.62E-30 0.79

C-14 2.06E-09 7.29E-10 7.23E-10 0.35

Cl-36 6.52E-08 1.64E-07 6.88E-08 2.51

Fe-55 1.04E-43 5.62E-44 5.78E-44 0.54

Co-60 1.21E-39 7.44E-40 6.86E-40 0.61

Ni-63 7.94E-21 3.66E-21 2.58E-21 0.46

Sr-90 3.04E-24 8.49E-25 1.32E-24 0.28

Nb-94 3.76E-10 3.18E-10 3.08E-10 0.85

Tc-99 2.12E-09 1.60E-09 9.71E-10 0.76

Ru-106 3.44E-46 1.55E-46 1.18E-46 0.45

Ag-108m 1.68E-17 1.08E-17 1.04E-17 0.65

Sb-125 4.81E-42 2.29E-42 1.68E-42 0.48

Sn-126 1.2E-08 8.99E-09 8.11E-09 0.75

I-129 1.38E-05 2.14E-06 4.07E-06 0.16




Adult Infant Child

Infant - Adult

Ba-133 1.44E-31 1.23E-32 2.07E-32 0.09

Cs-134 1.82E-43 1.29E-44 2.37E-44 0.07

Cs-137 1.28E-25 4.17E-26 6.74E-26 0.33

Pm-147 3.41E-44 1.85E-44 1.24E-44 0.54

Eu-152 2.77E-32 1.28E-32 1.08E-32 0.46

Eu-154 3.28E-35 1.58E-35 1.26E-35 0.48

Eu-155 3.63E-41 1.86E-41 1.30E-41 0.51

Pb-210 1.25E-25 4.85E-26 5.74E-26 0.39

Ra-226 1.56E-08 4.58E-09 8.22E-09 0.29

Ac-227 5.66E-26 1.19E-26 1.29E-26 0.21

Th-229 1.44E-08 2.26E-09 3.08E-09 0.16

Th-230 8.24E-09 1.32E-09 1.80E-09 0.16

Th-232 4.04E-08 6.30E-09 8.79E-09 0.16

Pa-231 3.6E-08 5.45E-09 8.50E-09 0.15

U-232 5.07E-20 9.40E-21 1.46E-20 0.19

U-233 4.62E-09 9.50E-10 1.17E-09 0.21

U-234 4.35E-09 8.70E-10 1.10E-09 0.20

U-235 4.35E-09 9.21E-10 1.11E-09 0.21

U-236 4.22E-09 8.81E-10 1.05E-09 0.21

U-238 4.35E-09 8.80E-10 1.10E-09 0.20

Np-237 1.22E-06 1.74E-07 2.04E-07 0.14

Pu-238 1.86E-12 3.73E-13 4.70E-13 0.20

Pu-239 1.54E-08 1.97E-09 2.78E-09 0.13

Pu-240 1.04E-08 1.34E-09 1.89E-09 0.13

Pu-241 1.59E-10 2.27E-11 2.73E-11 0.14

Pu-242 1.69E-08 2.14E-09 3.06E-09 0.13

Am-241 5.37E-11 7.65E-12 9.17E-12 0.14

Cm-243 1.82E-10 2.33E-11 3.29E-11 0.13

Cm-244 1.29E-09 1.65E-10 2.33E-10 0.13

Table C.5 Sensitivity studies on the effect of changes in the exposed individual. Specific doses are to members of the public via use of water from a borehole 1500 m from the site for irrigation. Results include the effects of ingrowth of daughter radionuclides over 100 years.



C.2 Leachate Spillage



by leachate



Drinking water


associated products

Soil

H-3 2.96E-12 3.44E-15 1.92E-11 1.37E-12 9.52E-17

C-14 4.15E-11 4.34E-10 2.58E-10 2.09E-11 1.34E-15

Cl-36 9.85E-11 5.73E-12 6.40E-10 7.37E-11 3.17E-15

Fe-55 5.70E-11 6.63E-12 3.55E-10 3.37E-12 1.84E-15

Co-60 5.70E-10 1.99E-10 3.55E-09 1.92E-11 1.84E-14

Ni-63 1.45E-11 1.69E-12 9.03E-11 3.48E-12 4.68E-16

Sr-90 3.41E-09 2.38E-10 2.13E-08 3.65E-10 1.10E-13

Nb-94 1.76E-10 6.15E-11 1.10E-09 1.17E-15 5.68E-15

Tc-99 6.74E-11 1.57E-12 4.57E-10 1.29E-13 2.17E-15

Ru-106 7.49E-10 8.71E-12 4.66E-09 1.11E-10 2.41E-14

Ag-108m 2.23E-10 1.30E-12 1.39E-09 1.81E-13 7.18E-15

Sb-125 3.36E-10 3.91E-11 2.09E-09 1.57E-13 1.08E-14

Sn-126 7.03E-10 8.17E-10 4.38E-09 1.43E-11 2.26E-14

I-129 9.85E-09 3.44E-10 6.13E-08 2.59E-09 3.17E-13

Ba-133 5.73E-11 2.67E-13 3.56E-10 4.89E-13 1.85E-15

Cs-134 7.58E-10 1.76E-09 4.72E-09 1.96E-10 2.44E-14

Cs-137 2.07E-09 4.81E-09 1.29E-08 5.35E-10 6.66E-14

Pm-147 2.96E-11 1.03E-12 1.84E-10 4.43E-13 9.52E-16

Eu-152 1.35E-10 4.70E-12 8.39E-10 2.81E-13 4.34E-15

Eu-154 2.13E-10 7.42E-12 1.32E-09 4.42E-13 6.85E-15

Eu-155 3.53E-11 1.23E-12 2.19E-10 7.34E-14 1.14E-15

Pb-210 2.70E-07 9.42E-08 1.68E-06 1.45E-09 8.70E-12

Ra-226 3.22E-07 1.87E-08 2.00E-06 6.76E-09 1.04E-11

Ac-227 8.55E-08 7.95E-08 5.32E-07 4.07E-11 2.75E-12

Th-229 4.02E-08 1.40E-09 2.50E-07 3.22E-10 1.30E-12

Th-230 1.24E-08 4.34E-10 7.74E-08 9.96E-11 4.01E-13

Th-232 6.95E-08 2.43E-09 4.33E-07 5.56E-10 2.24E-12

Pa-231 4.77E-08 5.55E-10 2.97E-07 1.04E-11 1.54E-12

U-232 4.12E-08 4.79E-10 2.56E-07 2.67E-10 1.33E-12





by leachate



Drinking water


associated products

Soil

U-233 4.04E-09 4.70E-11 2.52E-08 2.62E-11 1.30E-13

U-234 3.84E-09 4.46E-11 2.39E-08 2.49E-11 1.24E-13

U-235 3.71E-09 4.31E-11 2.31E-08 2.40E-11 1.19E-13

U-236 3.63E-09 4.22E-11 2.26E-08 2.35E-11 1.17E-13

U-238 3.79E-09 4.41E-11 2.36E-08 2.46E-11 1.22E-13

Np-237 5.75E-09 6.69E-11 3.58E-08 1.73E-11 1.85E-13

Pu-238 1.24E-08 5.79E-11 7.74E-08 5.57E-13 4.01E-13

Pu-239 1.40E-08 6.51E-11 8.71E-08 6.27E-13 4.51E-13

Pu-240 1.40E-08 6.51E-11 8.71E-08 6.27E-13 4.51E-13

Pu-241 2.64E-10 1.23E-12 1.64E-09 1.18E-14 8.52E-15

Pu-242 1.35E-08 6.27E-11 8.39E-08 6.04E-13 4.34E-13

Am-241 1.14E-08 3.98E-10 7.10E-08 1.58E-12 3.67E-13

Cm-243 8.30E-09 2.89E-10 5.16E-08 2.55E-12 2.67E-13

Cm-244 7.32E-09 2.55E-10 4.55E-08 2.25E-12 2.36E-13

Table C.6 Specific doses to a child (10 years) via exposure pathways associated with spillage of leachate into a surface water resource. Results do not include the effects of ingrowth of long-lived daughter radionuclides.



by leachate



Drinking water


associated products

Soil

H-3 4.63E-12 2.89E-15 1.80E-11 2.74E-12 1.00E-15

C-14 6.17E-11 3.47E-10 2.30E-10 2.57E-11 1.34E-14

Cl-36 2.43E-10 7.60E-12 9.45E-10 2.41E-10 5.26E-14

Fe-55 9.26E-11 5.79E-12 3.45E-10 2.51E-12 2.00E-14

Co-60 1.04E-09 1.95E-10 3.88E-09 2.15E-11 2.25E-13





by leachate



Drinking water


associated products

Soil

Ni-63 3.24E-11 2.03E-12 1.21E-10 1.32E-11 7.01E-15

Sr-90 3.09E-09 1.16E-10 1.15E-08 4.31E-10 6.68E-13

Nb-94 3.74E-10 7.02E-11 1.39E-09 3.99E-15 8.10E-14

Tc-99 1.85E-10 2.32E-12 7.51E-10 4.07E-13 4.01E-14

Ru-106 1.64E-09 1.02E-11 6.09E-09 1.09E-10 3.54E-13

Ag-108m 4.24E-10 1.33E-12 1.58E-09 5.65E-13 9.18E-14

Sb-125 7.66E-10 4.79E-11 2.85E-09 5.20E-13 1.66E-13

Sn-126 1.52E-09 9.50E-10 5.66E-09 4.35E-11 3.29E-13

I-129 8.49E-09 1.59E-10 3.16E-08 2.62E-09 1.84E-12

Ba-133 4.87E-11 1.22E-13 1.81E-10 6.47E-13 1.05E-14

Cs-134 6.77E-10 8.46E-10 2.52E-09 1.79E-10 1.46E-13

Cs-137 2.07E-09 2.59E-09 7.72E-09 5.49E-10 4.49E-13

Pm-147 7.33E-11 1.37E-12 2.73E-10 5.19E-13 1.59E-14

Eu-152 2.86E-10 5.35E-12 1.06E-09 5.34E-13 6.18E-14

Eu-154 4.63E-10 8.68E-12 1.72E-09 8.66E-13 1.00E-13

Eu-155 8.49E-11 1.59E-12 3.16E-10 1.59E-13 1.84E-14

Pb-210 3.81E-07 7.14E-08 1.42E-06 3.06E-09 8.24E-11

Ra-226 2.87E-07 8.98E-09 1.07E-06 9.76E-09 6.22E-11

Ac-227 1.31E-07 6.57E-08 4.89E-07 2.90E-11 2.84E-11

Th-229 4.83E-08 9.06E-10 1.80E-07 1.78E-10 1.04E-11

Th-230 1.58E-08 2.97E-10 5.89E-08 5.82E-11 3.42E-12

Th-232 8.03E-08 1.51E-09 2.99E-07 2.95E-10 1.74E-11

Pa-231 5.02E-08 3.14E-10 1.87E-07 9.65E-12 1.09E-11

U-232 4.41E-08 2.76E-10 1.64E-07 4.59E-10 9.54E-12

U-233 5.40E-09 3.38E-11 2.01E-08 5.62E-11 1.17E-12

U-234 5.02E-09 3.14E-11 1.87E-08 5.21E-11 1.09E-12

U-235 5.05E-09 3.16E-11 1.88E-08 5.25E-11 1.09E-12

U-236 5.02E-09 3.14E-11 1.87E-08 5.21E-11 1.09E-12

U-238 4.98E-09 3.11E-11 1.85E-08 5.18E-11 1.08E-12

Np-237 8.17E-09 5.11E-11 3.04E-08 1.18E-11 1.77E-12





by leachate



Drinking water


associated products

Soil

Pu-238 1.54E-08 3.86E-11 5.74E-08 6.28E-13 3.34E-12

Pu-239 1.62E-08 4.05E-11 6.03E-08 6.59E-13 3.51E-12

Pu-240 1.62E-08 4.05E-11 6.03E-08 6.59E-13 3.51E-12

Pu-241 2.20E-10 5.50E-13 8.19E-10 8.94E-15 4.76E-14

Pu-242 1.54E-08 3.86E-11 5.74E-08 6.28E-13 3.34E-12

Am-241 1.43E-08 2.68E-10 5.31E-08 1.29E-12 3.09E-12

Cm-243 1.27E-08 2.39E-10 4.74E-08 1.99E-12 2.76E-12

Cm-244 1.13E-08 2.12E-10 4.20E-08 1.77E-12 2.44E-12

Table C.7 Specific doses to a infant (1 year) via exposure pathways associated with spillage of leachate into a surface water resource. Results do not include the effects of ingrowth of long-lived daughter radionuclides.

Annex C

ENRMF, IRRs 1999, Radiation Risk Assessment for Low Level Waste Disposal,

HPA

VERSION 3, 14 July 2009

1

EAST NORTHANTS RESOURCE MANAGEMENT FACILITY

IONISING RADIATIONS REGULATIONS 1999

RADIATION RISK ASSESSMENT FOR LLW WITH A SPECIFIC ACTIVITY UP TO

200Bq/g

1 SCOPE AND DEFINITIONS

1.1 INTRODUCTION

The East Northants Resource Management Facility (ENRMF) operated by Augean plc is

intending to dispose of low level radioactive wastes (LLW) with a specific activity of up to

200Bq/g. An application under the Radioactive Substances Act 1993 has been prepared, and

this includes an assessment of the potential radiation exposure of workers and members of

the public. In addition to this, the Ionising Radiations Regulations 1999 (IRR99) require that a

radiological risk assessment is undertaken for any work involving ionising radiation.

Specifically, Regulation 7 requires the radiation employer (Augean plc) to carry out a prior risk

assessment before commencing work with radioactive materials at the ENRMF site. This

document is intended to meet the requirements of this Regulation in relation to the operational

phase of the controlled burial operation.

1.2 RADIOACTIVE MATERIALS AND RADIATION HAZARDS

The type and quantities of radioactive materials that may be accepted at ENRMF are

described in the RSA93 application and supporting documents. In brief, the application

includes a range of potential radionuclides from nuclear and non-nuclear practices (including

radionuclides of natural origin) with a maximum total activity concentration of 200 Bq/g. This

assessment pessimistically assumes that the waste received contains radionuclides at the

maximum activity concentrations, which is unlikely to be the case in practice.

The radionuclides considered emit a combination of alpha and beta particles and gamma

rays. The handling of these materials can potentially give rise to a radiation hazard from:

- external gamma exposure from proximity to the waste (either during handling waste

containers or occupancy of the disposal areas);

- internal radiation exposure from the inhalation of contaminated dust (air contamination)

arising during the work;

- internal radiation from the transfer and inadvertent ingestion of material (surface

contamination) during the work; and

- internal radiation from any contaminated wounds incurred during the work.

This risk assessment focuses on the exposure of workers and other persons visiting the

ENRMF site. The potential radiation exposure of persons off-site (i.e. members of the public)

from a range of exposure pathways has been considered in detail in the RSA93 application.

This demonstrated that the maximum dose to a member of the public is expected to be

below 0.02 mSv per year1. This is well below the relevant IRR99 dose limit of 1 mSv per

1 A higher dose of up to 1 mSv was associated with accidental (public) intrusion into the landfill. This is a post-closure scenario and is beyond the scope of this risk assessment.


2

year, and consequently doses to persons off-site are not considered in detail in this risk

assessment.

1.2 RISK ASSESSMENT REQUIREMENTS

The purpose of this risk assessment is to identify the measures needed to restrict the

exposure of employees and other persons to ionising radiation from the controlled burial

waste (LLW) operations at ENRMF.

IRR99 Regulation 7 also requires that potential radiation accidents are identified and

quantified, and that steps are taken to prevent accidents, limit the consequences of any

accidents that do occur, and to provide any necessary information, instruction, training and

equipment to deal with such accidents.

Paragraph 44 of the Approved Code of Practice to IRR99 recommends that the following

matters should be considered when carrying out this risk assessment. The parts of this

document that correspond to these matters are listed in the table below.

Nature of the radiation source 1.2

Estimated radiation dose rates to which anyone can be exposed 2.1

Likelihood of contamination arising and being spread 2.2

Results of previous personal dosimetry or area monitoring 2.1.1

Advice from manufacturers or suppliers 4.5

Engineering control measures and design features 4.1

Any planned systems of work 4.1, 4.4.1

Estimated levels of airborne and surface contamination 2.2.1, 3

Effectiveness and suitability of personal protective equipment 4.5

Extent of unrestricted access to working areas where dose rates or

contamination levels are likely to be significant

4.2

Possible accident situations, their likelihood and potential severity 3

The failure of control measures or systems of work 3

Steps to prevent identified accident situations or limit their consequences 3, 4.1.3


3

Paragraph 45 of the ACoP states that the risk assessment should enable the radiation

employer to determine the following outcomes. Again, the relevant parts of this risk

assessment are indicated in the table below.

What action is needed to ensure that radiation exposures are as low as

reasonably practicable (ALARP)

4.1

What engineering controls, design features, safety and warning devices, and

systems of work are needed

4.1

Whether it is appropriate to provide personal protective equipment 4.1, 4.5

Whether dose constraints for planning purposes are needed 4.1.4

The need to alter the working conditions of any female employee who

declares she is pregnant or breastfeeding

4.1.6

A dose investigation level to check that exposures are ALARP 4.1.5

What maintenance and testing schedules are required 4.5

What contingency plans are necessary 4.1

The training of classified and non-classified employees 4.6

The need to designate specific areas as controlled or supervised and the

need for local rules

4.2

The actions needed to ensure restriction of access for controlled or

supervised areas

4.2, 4.4.1

The need to designate certain employees as classified persons 4.3

The need for individual dose assessment 4.3

The responsibilities of managers 4.4.2

An appropriate programme of monitoring or auditing of arrangements. 4.7


4

2 RADIATION RISKS FROM NORMAL OPERATIONS

2.1 RADIATION DOSE RATES AND EXTERNAL RADIATION RISKS

2.1.1 Augean employees engaged in the LLW operation

The radiation dose rates from a range of radionuclides have been calculated as part of the

supporting documents to the RSA93 application. Extracts from these documents, relevant to

the estimation of external dose, are reproduced in Appendix 1.

For all external dose scenarios, cobalt-60 is the limiting radionuclide (i.e. it gives rise to the

highest dose rates). For the purpose of this risk assessment, the following representative

dose rates, working patterns and estimated doses are used.

Work activity Dose rate

(μSv/h)

Occupancy

(hours/year)

Estimated

annual dose

(mSv)

Receipt of waste

consignments, including QA

and monitoring, etc.

10

2

50

100

0.5

0.2

Transfer and placement of

waste in landfill.

2

100

0.2

Occupancy of covered waste

area 2 100 0.2

TOTAL ESTIMATED ANNUAL DOSE 1.1 mSv

Summary

The above estimates are likely to be conservative, and it is unlikely that the same

person(s) will be exposed during all the work activities listed above. Nevertheless, it

is reasonable to assume, for planning purposes, that annual external doses of the

order of 1 mSv per year might be associated with the LLW operation.

2.1.2 External radiation risks to other persons

Such persons might include other employees (i.e. not involved in the LLW operation), visitors

to site, etc. Such persons would be unlikely to be exposed to dose rates above 1 µSv/h, and

exposure times would be expected to be short. Consequently, it is expected that external

doses to such persons should be negligible.

External doses to members of the public (during and after the LLW operation) were

estimated in the RSA93 application, and are a small fraction of the 20 µSv/y dose constraint.

2.2 CONTAMINATION LEVELS AND INTERNAL RADIATION RISKS

The waste will be in closed containers (either steel drums or bulk bags) throughout the LLW

operation. Furthermore, waste consigners will be required to demonstrate that the external

surfaces of these containers are effectively free of loose contamination. Consequently,

contamination levels, and hence internal radiation doses, during normal operations are

expected to be negligible.

There is the potential for contamination and internal exposures arising from accidents, in

particular a damaged container. This is considered in the Section 3.


5

3 RADIATION RISKS FROM ACCIDENTS

The following reasonably foreseeable incidents/accidents have been identified:

3.1 The delivery of waste containing unexpectedly high levels of radioactivity

The likelihood of receiving waste that is more radioactive than expected is limited by the strict

pre-acceptance criteria and associated procedures that are to be put in place. In addition, it is

expected that incoming consignments will be monitored, and a dose rate acceptance test

applied. Thus any radiation exposures from this scenario, should be limited to a brief external

exposure to increased dose rates at the receiving stage. Even if the dose rate is 10x the

acceptance criteria, the resulting doses to workers from the monitoring and subsequent

quarantine of the consignment would be expected to be negligible.

3.2 Dropping or otherwise damaging a container of waste and spilling the contents

The “dropped bag” scenario is specifically considered in the RSA93 application using a

pessimistic dispersion model to estimate the radiation doses (from dust inhalation) to workers

and persons off-site. This assessment is principally concerned with the exposure of workers,

in particular those that may be involved in cleaning up any spills. Consequently, for this risk

assessment the following general “spillage” scenario is assumed:

either type of waste container (drum or bag) could be damaged;

contaminated dust is released producing a localised dust loading of 10 mg/m3, which

is considered a pessimistic assumption for an accident outdoors;

workers remain in the above dust loading for a total of 4 hours (to allow for any clean-

up).

the worker breathing rate is 1.2 m3/h and no respiratory protective equipment (RPE)

is worn; and

dust is inadvertently ingested (e.g. during the clean-up) at a rate of 3.45 x 10-5 kg/h

(the same rate as assumed in the RSA application for excavation scenarios)

The above assumptions produce an inhaled dust mass of 48 mg, and an ingested dust mass

of 138 mg. ICRP dose coefficients for inhalation and ingestion (the same as those used in

the RSA93 application) are given in the Appendix to this risk assessment. Combining these

with the mass of dust inhaled and ingested, and an activity concentration of 200 Bq/g (i.e. a

worst case assumption) gives the following (rounded) internal doses:

Estimated internal dose from a single spillage (mSv) Radionuclide

Inhalation Ingestion Total

Ac-227 5 5

Th-229 2 2

Th-230,232

Pa-231

Pu-238, 239, 240, 242

Am-241

1 1

Ra-228, Th-228

U-232, Np-237

Cm-243, 244

0.1 to 1 0.1 to 1

All other radionuclides <0.1

All < 0.05 mSv

<0.1


6

Thus, dust inhalation is the dominant exposure pathway. The highest estimated doses are for

actinium-227 and thorium-229. However, it is considered highly unlikely that waste would

contain these radionuclides at 200 Bq/g.

There would also be an external dose associated with the clean up. Assuming a 4 hour

exposure to an average dose rate of 10 µSv/h gives an external dose of 0.04 mSv. Taking all

these factors into account, it is concluded that the radiation exposure (internal plus

external) from a spillage of waste containing up to 200 Bq/g is unlikely to exceed 1

mSv. This includes any exposures from the subsequent clean-up of the spill.

3.3 Internal exposure from contaminated wounds

Under normal circumstances this is not a reasonably foreseeable exposure scenario.

However, if contamination does arise, for example because of the spill scenario in 3.2 above,

then this additional accident exposure pathway becomes a possibility. It is considered that

doses from this pathway would be likely to be the same order of magnitude as from

inadvertent ingestion, i.e., less than 0.1 mSv.

The UKAEA Safety Assessment Handbook (UKAEA/SAH/D9, Issue 1, March 2006) gives

dose factors for contaminated wounds. Assuming that 0.1 g of material (at 200 Bq/g)

becomes incorporated into a wound, the highest estimated dose is approximately 3 mSv, from

actinium-227. As mentioned above, this radionuclide is most unlikely to predominate, and it is

concluded that internal doses from a contaminated wound would be very unlikely to exceed 1

mSv in practice.


7

4. RECOMMENDED ACTIONS - REQUIREMENTS OF IRR99

4.1 RESTRICTION OF EXPOSURE (IRR99 REGULATION 8)

4.1.1 Summary of estimated doses

Regulation 8 requires that every radiation employer shall take all necessary steps to ensure

that the radiation exposure of employees and other persons is as low as reasonably

practicable (ALARP). The preceding dose assessment produced the following estimated

effective doses:

Augean LLW workers Normal operations: 1 mSv/y from external exposure Accidents: Negligible (<0.1 mSv) increase in external dose due to

receipt of consignments containing higher than expected activity concentrations. Up to 1 mSv (principally from inhalation) from dealing with spills of loose waste material (without special precautions). Up to 1 mSv from contaminated wounds incurred when

dealing with spills of loose waste material (without

special precautions).

Other persons Normal operations and accidents

Doses are expected to be negligible, but even in the worst case should be much less than those estimated above for workers.

The estimated doses are a small fraction of the annual dose limits of 20 mSv (for workers)

and 1 mSv (for other persons) specified in IRR99. Notwithstanding this, there is still a

requirement to keep exposures ALARP, and the recommended steps needed to achieve this

are given below.

4.1.2 Protection during normal operations

Augean LLW workers

All LLW waste should be received, handled and disposed of in closed containers.

Consignments of waste should be checked and verified as they arrive on site.

External radiation exposures can be restricted by setting a limit on the dose rate from

each waste container. Based on this risk assessment, a limit of 10 µSv/h at 1 metre

from the container is recommended. A quarantine area should be provided for holding

containers that exceed this level.

External radiation exposures can be restricted through time and distance. Procedures

should be in place to ensure that LLW consignments are disposed of as quickly as

possible. Workers should be instructed to avoid loitering near waste containers.

Deposited waste should be covered with a compacted 300mm inactive layer, as soon as

practicable. The dose rate above the covered waste should be monitored and should

not exceed 2 µSv/h. If necessary, additional covering material should be applied until

this is achieved.

Although contamination is not expected during normal operations, it is good practice for

workers to wear suitable overalls and gloves during the LLW work, which will provide


8

protection in the event of a spillage of waste. This is existing practice at the ENRMF site

for al operatives.

Radiation monitoring (individual and environmental) is required – see 4.3 and 4.7 below.

Local rules and training should be provided - see 4.4 and 4.6 below.

Other persons

Other persons should be excluded from the immediate area during the LLW operation.

The dust suppression measures for spills, as described below, should also ensure that

the spread of airborne dust is minimised. No other specific protection measures are

required.

4.1.3 Accidents – prevention and mitigation

Dose rate checks on incoming consignments of waste should be undertaken, as

recommended above.

Contingency plans should be prepared for dealing with spillages of waste. These should

include the following precautions:

o Simple dust suppression measures (e.g. damping down, and avoiding dust

resuspension during clean-up) should be applied, where practicable.

o As an additional precaution, workers should wear respiratory protective

equipment when cleaning up spills – see 4.5 below.

o Spilled material must be placed into suitable containers for disposal, and the

affected area should be monitored to ensure that all contaminated material has

been removed.

The above precautions should ensure that the radiation doses from accidents are

negligible (<0.1 mSv).

4.1.4 Dose constraints

Regulation 8(3) requires that dose constraints are considered at the planning stage of

radiation protection. This has also been considered as part of the RSA93 application. Based

on the original application, and on the implementation of the findings of this risk assessment,

the following dose constraints are recommended:

LLW workers: 1 mSv/y

Other persons: 0.02 mSv/y

4.1.5 Dose investigation level

Regulation 8(7) requires that employers should set an investigation level for the purposes of

determining whether exposures are being kept ALARP. Based upon the findings of this risk

assessment it is recommended that Augean set a dose investigation level of 1 mSv for its

employees.

The monitoring required to compare exposures against the investigation level is discussed in

4.3 below.

If the dose investigation level is exceeded, Augean in consultation with their RPA should

undertake an investigation to determine whether the steps being taken to restrict exposures

are sufficient.


9

4.1.6 Pregnant and breast-feeding employees

Regulation 8(5) contains additional dose restriction provisions for such employees. For

pregnant women, it is recommended the dose to the foetus should be kept below 1 mSv.

Whilst exposures of over 1 mSv are unlikely to occur, as a precaution it is recommended that

pregnant employees are not allowed in the LLW work areas.

For breastfeeding women, it is recommended that they avoid situations where significant

bodily contamination might occur. As a general precaution, it is recommended that such

women are not allowed in the LLW work areas.

The risks associated with radiological hazards should be incorporated in the company risk

assessment for pregnant and breastfeeding employees.

4.2 DESIGNATED AREAS

4.2.1 Controlled areas

Regulation 16 requires the designation of a controlled area where either:

a) radiation doses are likely to exceed three-tenths of the annual dose limits for workers

(e.g. 6mSv/y effective dose); or

b) special working procedures are required to restrict radiation exposures.

Worker doses are not expected to exceed 6 mSv/y. However, it is considered that special

working procedures (as defined in Regulation 16(1)) are appropriate in respect of certain

operations. Consequently the following recommendations are made:

Incoming waste consignments should be rapidly processed, and should not remain in any

one area for an extended period of time. On this basis, a controlled area (for example,

around arriving vehicles) is not recommended.

A quarantine area should be provided for waste consignments that do not meet the dose

rate limits described previously, and this should be designated as a controlled area

whenever such consignments are quarantined. It should be ensured that the dose rate

outside this area is below 2 µSv/h.

During the deposition of waste containers, elevated dose rates are present, and there is

the potential for accidents (dropped containers, etc.). It is recommended that this area is

designated as a controlled area during the disposal operation, and should remain

designated until a satisfactory covering layer has been applied (see 4.1.2).

Controlled areas should, where practicable be physically demarcated and warning signs

posted at the points of entry. For the above areas, the following is recommended:

Quarantine area: the perimeter should be fully demarcated, ideally with fencing, but if

not, with rope barriers or similar. A controlled area warning sign should be posted at all

points of potential access.

Disposal area: during the operational period the area will be occupied or under

surveillance and it is considered sufficient to temporarily post controlled area warning

signs at the access points to the area.


10

Access to the controlled areas should be restricted to authorised personnel. Local rules and

Radiation Protection Supervisors (Regulation 17) should be provided for controlled areas –

see 4.4 below.

4.2.1 Supervised areas

The Regulations also require that a supervised area should be designated where it is

considered necessary to keep the radiological conditions under periodic review. Although

some confirmatory monitoring is recommended outside the controlled areas (see 4.7 below),

the designation of a supervised area is not considered necessary provided that the

aforementioned dose rate limits are met.

4.3 CLASSIFIED PERSONS AND INDIVIDUAL MONITORING

4.3.1 Designation of classified persons

Regulation 20 requires workers to be designated as classified persons if they are likely to

receive an effective dose in excess of 6 mSv per year. This risk assessment indicates that

doses are expected to be well below this value and, therefore, it is not recommended that

Augean employees are designated as classified persons.

4.3.2 Monitoring of individual dose

As a means of confirming the restriction of exposures, and for checking against the Dose

Investigation Level, it is recommended that a programme of individual dose monitoring is

implemented for all Augean employees engaged in the LLW operation. For monitoring

external exposure, it is recommended that passive whole body dosemeters (e.g. TLDs) are

worn, and Augean should make the necessary arrangements with an appropriate dosimetry

service.

Internal exposures during normal operations are expected to be negligible, and the

precautions listed in Section 4.1.3 should ensure that this is also the case for internal

exposures from accidents. The systematic assessment of individual internal dose is not,

therefore, warranted (see ACoP paragraph 386).

4.4 WORKING PROCEDURES AND SUPERVISION

4.4.1 Local rules

Regulation 17 requires that Local Rules are written for work in controlled areas. Augean

should draft Local Rules, consulting the RPA as required, to ensure that the format and

content of the rules (as specified in IRR99) are appropriate. The Local Rules must include:

- the dose investigation level;

- a description of each controlled area, and the means by which access is restricted;

- names of the Radiation Protection Supervisors (see below);

- for each controlled area, appropriate working instructions (PPE, good working

practice, monitoring arrangements, etc) including written arrangements for the entry of

non-classified persons into the controlled areas;

- details of any contingency arrangements, for example for dealing with spillages.


11

4.4.2 Radiation Protection Supervisors (RPSs)

Regulation 17 requires that Augean appoint one or more employees as RPS. The main role

of the RPS is to ensure that the Local Rules are being observed, and whoever is appointed

should be suitable for the role. In practice, this means that they are appropriately trained and

are able to properly supervise the work being undertaken. There should be an RPS present

on the ENRMF site whenever LLW is being processed.

It should be noted that the RPSs are not a substitute for line-management responsibilities.

Augean must ensure that line managers involved in the LLW operation project are familiar

with the contents of this risk assessment and the local rules, and their responsibilities for

health and safety.

4.5 PERSONAL PROTECTIVE EQUIPMENT

The internal dose to Augean employees from inhalation of dust during normal operations is

expected to be negligible. As indicated in Section 3.2, the inhalation dose from dealing with a

waste spill is likely to be below 1 mSv. Although this is well below the 20 mSv/y dose limit, it

is recommended that respiratory protective equipment be worn in the interests of keeping

exposures ALARP, and to ensure compliance with the dose constraint and dose investigation

level.

The RPE should be readily available in the event of a spill occurring, and must be put on

before attempting to clean up any spilt LLW material.

RPE with a minimum protection factor of 5 is recommended: this, combined with the dust

suppression measures described in Section 4.1.3, should ensure that inhalation doses are

below 0.1 mSv. In addition:

- RPE must be CE marked;

- the comfort of the wearer should be taken into account when choosing a particular type

of respirator;

- RPE should be fit-tested to ensure a good seal to individual faces;

- If the RPE is reusable, it should be thoroughly examined at suitable intervals and

properly maintained in accordance with the manufacturer’s instructions, and as required

by Regulation 10(2). Suitable records of examinations and maintenance should be

made and kept for at least 2 years; and (very importantly)

- training in the proper use and maintenance of RPE must be provided.

In addition to RPE, protective clothing should also be worn by Augean employees when

working in the area, as follows:

- coveralls must be worn, the type being selected according to the nature of the work.

- protective gloves must be worn, the type being selected according to the nature of the

work. Gloves should be impermeable and be sufficiently strong to withstand wear and

tear and provide protection against cuts/wounds;

- footwear – normal safety footwear is considered sufficient; and

- suitable washing and changing facilities should be provided for use by workers before

lunch breaks, ends of shift, etc.. This should include facilities for separate storage of


12

clean and dirty clothing, and hand/face washing facilities with elbow-operated taps. It is

suggested that a contamination monitor should also be considered for reassurance

purposes, i.e. so that workers can check themselves if they wish.

After dealing with a spill, coveralls and gloves may need to be disposed of. It is suggested

that disposable outer coveralls and gloves should be provided for use when cleaning up spills.

Gloves should be taped to coveralls where there is a risk of up-sleeve contamination during a

clean-up. Footwear should be washed down before leaving the area.

4.6 INFORMATION, INSTRUCTION AND TRAINING

To meet the requirements of Regulation 14, the following arrangements are recommended:

All Augean employees engaged in LLW work should receive training in radiation

protection prior to the work. This should cover:

o the nature of the radiation hazards associated with LLW;

o the risks to health associated with exposure to radiation;

o the precautions that need to be taken to restrict exposures, including the contents

of this risk assessment and the local rules;

o the correct use of RPE; and

o the regulatory requirements associated with the work, and the importance of

complying with these requirements.

In addition, specifically appointed Augean employees should receive additional

training to act as a Radiation Protection Supervisor(s) and (if applicable) to examine and

maintain RPE.

Other persons working on the ENRMF site should be provided with information to

indicate that the certain areas are designated as controlled areas, that access to these

areas is restricted, and that warning signs should be observed.

4.7 WORKPLACE MONITORING

The following programme of workplace monitoring is recommended.

Dose rates

All incoming LLW containers should be subject to dose rate monitoring, and the results

recorded. The dose rate a 1 metre from a container must not exceed 10 µSv/h.

Any containers that do not meet the above criteria should be placed in quarantine. The

dose rate around the perimeter of the quarantine area must be measured (and recorded)

whenever containers are placed inside. The dose rate at the perimeter must not exceed

2 µSv/h.

The dose rate on top of any newly deposited material must be measured after the

minimum 300mm cover is applied. The dose rate at a height of 1 metre must not exceed

2 µSv/h. If necessary, additional cover should be applied. The measured dose rate and

the thickness of cover applied should be recorded.


13

Annual environmental-level dose rate monitoring will be undertaken by the RPA at

representative locations around the site boundary.

Surface contamination monitoring

Surface contamination is not expected to arise during routine operations. However,

confirmatory monitoring should be undertaken once every month in the following areas:

o At the exit point from the disposal area

o After the vehicle wheel wash

o Change rooms including PPE

o At the main exit from the site.

In addition, contamination monitoring should be undertaken after cleaning up any waste

spillages. This should include:

o Monitoring the affected area, i.e. to confirm that all contamination has been

removed.

o Monitoring all persons and items leaving the area to ensure that the spread of

contamination is avoided.

P V Shaw

14 July 2009

Document History

Version 1: 27 March 2009. First complete draft produced by RPA

Version 2: 7 July 2009. Incorporating comments by Augean

Version 3: 14 July 2009. Revised by RPA to incorporate comments.


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APPENDIX TO ENRMF RISK ASSESSMENT

SUPPORTING RADIOLOGICAL DATA TAKEN FROM RSA93 APPLICATION

A.1 EXTERNAL DOSE DATA

WASTE IN CONTAINERS

Specific calculations were undertaken for cobalt-60 (the most restrictive radionuclide) and

caesium-137 (for comparison) at 200 Bq/g – for both high and low density waste in drums and

bulk bags. A summary of the results is given below.

Estimated dose rate (µSv/h)2 Exposure

scenario Cobalt-60 Caesium-137

Drums

- contact (1 cm)

- 1 metre

- 2 metres

100

6

2

25

1.5

0.5

Bulk waste bags

- contact (1 cm)

- 1 metre

- 2 metres

125

14

5

6

3

1

DEPOSITED WASTE

Specific calculations were also undertaken to estimate the dose rate above deposited waste

covered with 30 cm of compacted topsoil. The results are summarised in the following table.

Radionuclides Calculated dose rate (µSv/h)

Cobalt-60 at 200 Bq/g

Other radionuclides at 200 Bq/g

5 to 10

<1

In this assessment a maximum dose rate of 2 µSv/h above the covered waste has been

recommended (see 4.1.2). This value has, therefore, been used (in part 2.1.1) to estimate

doses to workers.

2 The values have been rounded and represent the average dose rate calculated for high density (2g/cm3) and low density (1 g/m3) waste. In the case of cylindrical drums, the average values calculated for the (curved) sides and (flat) ends are given.


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A.2 INTERNAL DOSE DATA - ICRP INTERNAL DOSE COEFFICIENTS

For consistency purposes, the data below are the same as those used in the RSA93

application, and are the relevant ICRP dose coefficients for members of the public. The ICRP

dose coefficients for workers are slightly different, but this does not materially affect the

outcome of this risk assessment.

Dose coefficient (Sv/Bq) Radionuclide

Inhalation Ingestion

H-3 2.6E-10 1.8E-11

C-14 5.8E-09 5.8E-10

Cl-36 7.3E-09 9.3E-10

Fe-55 7.7E-10 3.3E-10

Co-60 3.1E-08 3.4E-09

Ni-63 4.8E-10 1.5E-10

Sr-90 1.6E-07 2.8E-08

Nb-94 1.1E-08 1.1E-08

Tc-99 1.3E-08 6.4E-10

Ru-106 6.6E-08 7.0E-09

Ag-108m 3.7E-08 2.3E-09

Sb-125 5.5E-08 3.1E-09

Sn-126 3.1E-08 7.1E-09

I-129 3.6E-08 1.1E-07

Ba-133 3.1E-09 1.5E-09

Cs-134 6.8E-09 1.9E-08

Cs-137 3.9E-08 1.3E-08

Pm-147 5.0E-09 2.6E-10

Eu-152 4.2E-08 1.4E-09

Eu-154 5.3E-08 2.0E-09

Eu-155 6.9E-09 3.2E-10

Pb-210 5.6E-06 6.9E-07

Ra-226 9.5E-06 2.8E-07

Ac-227 5.5E-04 1.1E-06

Th-229 2.6E-04 6.1E-07

Th-230 1.0E-04 2.1E-07

Pa-231 1.4E-04 7.1E-07

Th-232 1.1E-04 2.3E-07

U-232 4.7E-05 4.6E-07

U-233 9.6E-06 5.1E-08

U-234 9.4E-06 4.9E-08

U-235 8.5E-06 4.7E-08

U-236 3.2E-06 4.7E-08

U-238 8.0E-06 4.5E-08

Np-237 5.0E-05 1.1E-07


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Dose coefficient (Sv/Bq) Radionuclide

Inhalation Ingestion

Pu-238 1.1E-04 2.3E-07

Pu-239 1.2E-04 2.5E-07

Pu-240 1.2E-04 2.5E-07

Pu-241 2.3E-06 4.8E-09

Pu-242 1.1E-04 2.4E-07

Am-241 9.6E-05 2.0E-07

Cm-243 3.1E-05 1.5E-07

Cu-244 2.7E-05 1.2E-07

Annex D

Dose Rate calculations in support of Low Level Waste disposal authorisation,

TSG(09)0487

NOT PROTECTIVELY MARKED

Reference: TSG(09)0487

Issue: Issue 2 Technical Services GroupDate: 15th July 2009

DOSE RATE CALCULATIONS IN SUPPORT OF A LOW LEVEL WASTE DISPOSAL AUTHORISATION UK-10497

SUMMARY Dose rate calculations were performed in MicroShield to support a low level waste disposal authorisation. The dose rate was calculated on contact, 1m and 2m from a 200-litre drum and a bulk waste bag of soil and rubble waste. Dose was found to be highest when dealing with a 60Co source.

Name and Organisation Signature Date

Prepared By:

Tony Lansdell

TSG

Checked By:

Barry Cook

TSG

Approved By:

Gráinne Carpenter

TSG

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ELECTRONIC COPY


Table of Contents

1 Introduction.......................................................................................................................3 2 Methodology.....................................................................................................................3

2.1 Background ..............................................................................................................3 2.2 Materials ...................................................................................................................3 2.3 200-litre drum case...................................................................................................3 2.4 Bulk waste bag case.................................................................................................4 2.5 MicroShield calculation details and uncertainties .....................................................5

3 Results..............................................................................................................................6 3.1 Low density case ......................................................................................................6 3.2 High density case .....................................................................................................6

4 References .......................................................................................................................7

2


1 INTRODUCTION

Dose rate calculations were required to support a low level waste disposal authorisation. Cases were run using MicroShield v7.02 [1] to determine the maximum dose rate at a series of distances from the wasteform, for two different wasteform geometries.

2 METHODOLOGY

2.1 Background

MicroShield was used to determine the maximum dose rates at various distances from packaged contaminated soil and rubble waste. Two cases were defined, one for waste packaged in a 200-litre drum, and one for waste in a flexible bulk waste bag. In each case, the dose rate was required on contact, at 1m and at 2m from the wasteform.

2.2 Materials

Two sub-cases were defined; one for soil/rubble waste containing 200 Bq/g of 60Co, and one for soil/rubble containing 200 Bq/g of 137Cs. As soil is not a material type available to MicroShield, concrete was chosen to represent the waste material composition.

The bulk density of the soil and rubble wastes will vary depending on the composition of the waste, the level of compaction used, and the packing efficiency. Cases were assessed for two wasteform density values to provide bounding results.

A search of literature revealed that the bulk density of soil was typically 1.0 g/cm3 for loose soil, 1.3 g/cm3 for undisturbed soil, and 1.6 g/cm3 for compacted soil [2]. Concrete rubble was assumed to be the same as normal density concrete, 2.35 g/cm3.

The minimum density case was taken to be packaged loose soil, with a density of 1.0 g/cm3.

The maximum density wasteform was taken to contain the maximum amount of concrete rubble, with the remaining space taken up by compacted soil. It was assumed solid pieces of rubble would have a packing efficiency no better than 50%, hence 50% of the volume was assumed to be rubble (2.35 g/cm3), with the remaining 50% consisting of compacted soil (1.6 g/cm3). The maximum density of the wasteform was therefore predicted to be 1.98 g/cm3, and 2.0 g/cm3 was used for simplicity. The maximum range of wasteform density used was therefore between 1 and 2 g/cm3.

2.3 200-litre drum case

200-litre drums are steel-walled cylindrical drums of diameter of 67 cm and height 87 cm. The shielding effect of the drum was ignored to be conservative, hence the drum wall was not modelled in MicroShield, and the wasteform was taken to be a cylindrical volume of the above drum dimensions. Dose points were positioned on contact, 1m, and 2m from the wasteform, both in a radial direction ( ) and an axial direction ( ). Radial dose points were located at half the height of the cylinder, where the dose rate is maximised. Axial dose points were on axis with the centre of the cylinder, where the dose rate is maximised.

Figure 1Figure 2

3


Figure 1: Radial dose points for 200-litre drum (images from MicroShield)

Figure 2: Axial dose points for 200-litre drum

2.4 Bulk waste bag case

The bulk waste bag is a cube of side length 1m, and the wasteform was modelled in MicroShield as a rectangular volume with all sides 1m in length ( ). Again, the wall of the bag was not explicitly modelled to be conservative. The dose points were positioned in line with the centre of a flat face, where the dose rate is maximised.

Figure 3

4


Figure 3: Dose points for bulk waste bag

2.5 MicroShield calculation details and uncertainties

Energy deposition to dose rate conversion was performed automatically in MicroShield using built-in tables of effective dose rate, taken from ICRP-51 [3]. This presents a series of possible dose rates depending on the assumed irradiation geometry. The highest biological dose rate is produced assuming anterior-posterior geometry (with the gamma rays entering a person from the front and exiting through the back), and to be conservative it was this maximum dose rate that was reported. Dose rates can vary by approximately 30%, depending on which geometry is assumed.

MicroShield approximates the contribution of scattered radiation to the resulting dose rate by the use of build-up tables. The dose rate is dependant on which material is chosen as the dominant scattering medium. In accordance with the MicroShield manual, the material containing the highest number of gamma ray mean free paths should be used as the build-up material – hence in these cases, the source was chosen as build-up material. If the air gap is chosen as the scattering medium, it was found that the resulting dose rates increased by 6% for 60Co cases, and increased by 12% for 137Cs cases, but these results would be over-pessimistic.

MicroShield uses a point-kernel integration technique to determine the dose rate. This involves splitting the geometry into pieces (kernels). The quadrature order of the calculation determines the number of kernels used and hence the accuracy of the approximation; the default quadrature order was used for the reported results. The order of the calculation was increased by a factor of two in each dimension, and the contact results only increased by 0.3%, which is well within the range of other sources of uncertainty in the calculation. Further increases in accuracy produced no change to the results.

In all cases assessed, the ‘contact’ dose rate point was actually positioned at 1 cm from the surface, as the method of calculation used by MicroShield is known to become unstable at distances closer than 1 cm, though this will strongly depend on the integration order used.. 137Cs is a beta emitter. Its daughter, 137mBa is the source of the gamma radiation. Where a source containing 137Cs was specified, its daughter product 137mBa was also included in equilibrium concentration with 137Cs. Since the half-life of 137mBa is short (2.5 minutes), it will almost always be found in equilibrium with its parent radionuclide.

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3 RESULTS

3.1 Low density case

Density = 1.0 g/cm3, specific activity = 200 Bq/g = 200 Bq/cm3.

Dose Rate (µSv/hr)

Curved cylinder face Flat cylinder face Case

Contact* 100 cm 200 cm Contact* 100 cm 200 cm 60Co 91.65 6.019 1.95 98.35 4.766 1.465 137Cs 21.6 1.421 0.458 23.57 1.109 0.335

Dose Rate (µSv/hr)

Flat cube face Case

Contact* 100 cm 200 cm 60Co 120.7 12.45 4.065 137Cs 27.6 2.875 0.925

3.2 High density case

Density = 2 g/cm3, specific activity = 200 Bq/g = 400 Bq/cm3.

Dose Rate (µSv/hr)

Curved cylinder face Flat cylinder face Case

Contact* 100 cm 200 cm Contact* 100 cm 200 cm 60Co 109.2 7.293 2.316 123.9 5.682 1.654 137Cs 24.35 1.639 0.515 28.01 1.283 0.368

Dose Rate (µSv/hr)

Flat cube face Case

Contact* 100 cm 200 cm 60Co 131.3 14.5 4.526 137Cs 28.72 3.261 1.01

* Contact doses were located at 1 cm from the wasteform.

6


7

4 REFERENCES

1 MicroShield v7.02, Grove Software Inc, 2007 2 “Soils and Soil Fertility”, page 54, Sixth Edition, F.R. Troeh and L.M. Thompson,

Blackwell Publishing, 1979. 3 ICRP-51 (1987) Data for use in protection against external radiation

Annex E

Development of a Framework for Assessing the Suitability of Controlled Landfills to Accept Disposals of Solid

Low-Level Radioactive Waste: Principles and Technical Manual

SNIFFER, 2005

These manuals describe in detail the models used for the assessment in the application. Hardcopies have not been included. The information can be read and downloaded at: http://www.sniffer.org.uk/Resources/UKRSR03/Layout_EnvironmentalRegulation/11.aspx?backurl=http%3a%2f%2fwww.sniffer.org.uk%3a80%2fthemes%2fenvironmental-regulation.aspx&selectedtab=completed

Annex F

Application Form

Environment AgencyRadioactive Substances Form RSA3 (interim)Application for authorisation to accumulate and dispose of radioactivewaste from non-nuclear premisesRadioactive Substances Act 1993 Sections 13 & 14

NoteThis application form should be read and completed in conjunction with the current EnvironmentAgency guidance, available on the Environment Agency web sitehttp://www.environment-agency.gov.uk/business/444304/945840/1064273/?version=1&lang=_eor on request from Environment Agency offices (including the Environment Agency’s InterimGuidance to users of sealed sources on the High-activity Sealed Radioactive Sources and OrphanSources Regulations 2005). Words used in this form have the same meaning as in the aboveguidance.

To get an authorisation to accumulate and dispose of radioactive waste you generally also need tohold a registration for the premises. If you do not already hold a relevant and suitable registration forradioactive substances, you should fill in an application form RSA1 to cover the open or sealedsources you use or intend to use, and send it in with this form. You should note that the EnvironmentAgency may inspect the premises and/or ask the Police to review security during consideration of thisapplication for authorisation.

The issue of the certificate of authorisation under Sections 13 and 14 of the Radioactive SubstancesAct 1993 does not allow you to contravene any other statutory legislation that might also apply to thepremises.

This form should only be used for the accumulation and disposal of radioactive waste from a singledefined premise by a single organisation. You do not need to hold an authorisation to coveraccumulation and disposal of radioactive waste which is within the scope of an exemption order,provided you can comply with all of the conditions in such an order.

If you need more space than this form allows, please continue on separate sheets. Please write thenumber of the question you are answering on the top of each continuation sheet.

http://www.environment-agency.gov.uk/business/444304/945840/1064273/?version=1&lang=_e

ContentsA For Office UseB Company or Organisation DetailsC Type of ApplicationD Premises DetailsE Contact DetailsF Producing Radioactive WasteG Incineration on the PremisesH Disposal of Gaseous WasteI Aqueous WasteJ Organic Liquid WasteK Very Low Level Solid WasteL Solid Waste (excluding HASS and sources of similar potential hazard)M NAIR ArrangementsN ChecklistO Data HandlingP PaymentQ DeclarationR SignatureAnnexS Sealed Sources

A For Environment Agency use onlyNew application number

Date received – Agency date stamp

Existing authorisation for premises? YesExisting numberNo

New operator account? YesInvoice codeDateNo

Commercially confidential? YesNoSign

National security? YesNoSign

Fee £Date receivedAmount received £

SignDeclaration signed? Yes

NoNuclear site tenant? YesYes No

B Company or organisation details

B1 Please give the name of the company or organisation carrying on the undertaking creatingradioactive waste on the premises.Name

Registered office or business addressIf no registered office please give principal place of business

PostcodeCompanies House registration numberif you have one

B2 On behalf of what type of organisation are you applying?Tick the option which is most appropriate

Sole traderPartnershipLimited liability companyPublic limited companyDistrict or county council or unitary authorityEducational establishmentNHS trustPrivate hospitalOther medical establishment please give details

Non-governmental public bodyMinistry of DefenceOther Government departmentOther please give details

C Premises Details

C1 Where are the premises you want to accumulate and dispose of radioactive waste?Address

PostcodeOrdnance Survey national grid referenceFor example SJ 123 456

C2 Are the premises located on a nuclear licensed site?ie as a tenant

YesNo

C3 Which council or unitary authority are the premises in?If premises are on a boundary please give names of all relevant authorities.Borough or district council or unitary authority

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Augean South Limited

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4 Rudgate Court, Walton, Wetherby

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LS23 7BF

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4636789

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R

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R

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East Northants Resource Management Facility, Stamford Road, Kings Cliffe, Northamptonshire

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PE8 6XX

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TF 010 000

County council unless there is a unitary authority

C4 Who is the sewerage undertaker for the premises?This is often the local water supply company

C5 Which Police Force area are the premises in?

Avon & Somerset Bedfordshire Cambridgeshire Central Scotland Cheshire City of London Civil Nuclear Constabulary Cleveland Cumbria Derbyshire Devon & Cornwall Dorset Dumfries & Galloway Durham Dyfed-Powys Essex Fife Gloucestershire

Grampian Greater Manchester Gwent Hampshire Hertfordshire Humberside Kent Lancashire Leicestershire Lincolnshire Lothian & Borders Merseyside Metropolitan Ministry of Defence Police Norfolk Northamptonshire Northern Northumbria

North Wales North Yorkshire Nottinghamshire Northern Ireland South Wales South Yorkshire Staffordshire Strathclyde Suffolk Surrey Sussex Tayside Thames Valley Warwickshire West Mercia West Midlands West Yorkshire Wiltshire

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R

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East Northamptonshire Council

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Northamptonshire County Council

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This is currently tankered to an offsite treatment works

D Contact DetailsWe need the names and details of members of your organisation to help us deal with your applicationand authorisation quickly and efficiently.

Application contactD1 Who can we contact with questions on your application?Name

Position

Address

PostcodePhoneFaxE-mail

Operational contactD2 Who will be responsible for day to day supervision of the accumulation and disposal ofradioactive waste? If different people are responsible for some wastes, please give details of eachsuch personName

Position

Address


Payments and invoicesD3 Who can we contact about payment of fees and charges?Name

Position

Address


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Dr Gene Wilson

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Group Technical Director

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PE8 6XX

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01780 444905

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01780 444901

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[email protected]

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Simon Moyle

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Site Manager

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PE8 6XX

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01780 444900

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01780 444901

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[email protected]

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4 Rudgate Court, Walton, Wetherby

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LS23 7BF

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Adam Emmott

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01937 844980

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01937 844241

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[email protected]

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Head of Group Finance

E Type of Application

E1 When would you like the authorisation to start?We will try to meet your needs but it can take up to 4 months from date of receiving a valid applicationwith all of the information we need (and fee), before you receive your authorisation. After you receiveyour authorisation there is usually another 28 days before you can start accumulating and disposingof radioactive waste.Date

E2 When would you like any current authorisation cancelled?This will be the same date on which your new authorisation starts unless you tell us otherwise.Date to cancel any existing authorisation

E3 Have you made any other application to the Environment Agency (or previously HMIP) forany permission under the Radioactive Substances Acts, 1960 or 1993?

YesNo go to **

E4 Where relevant, please give details for a current or previous authorisation for thesepremises.User

Authorisation numberDate of authorisation

E5 Are you applying fora new authorisation for premises you do not hold a current authorisation for?a variation to an authorisation for your existing premises?a new authorisation for a new legal entity?a variation to an authorisation because you have changed your name but not your legal

status?other please give details

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17 November 2009

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No current authorisation

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R

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N/A

F Producing Radioactive Waste

F1 PracticePlease indicate the practice or work activity which creates radioactive waste.Please tick each relevant box.Note this list is adapted from the definitive list of existing practices on the Defra web site (link toDefra)

1.1 Enrichment of uranium - Use of the centrifuge process

2.1 Production of nuclear fuel - Manufacture of uranium metal and oxide fuel for power reactors 2.2 Production of nuclear fuel - Manufacture of mixed oxide fuel for power reactors 2.3 Production of nuclear fuel - Manufacture of uranium fuel for research or materials testing

reactors 2.4 Production of nuclear fuel - Manufacture of experimental nuclear fuel

3.1 Generation of electricity by nuclear reactors - Operation of Magnox power stations 3.2 Generation of electricity by nuclear reactors - Operation of advanced gas-cooled power

stations 3.3 Generation of electricity by nuclear reactors - Operation of pressurised water power stations

4.1 Recovery of usable products from spent nuclear fuel - Reprocessing of uranium metal frompower reactors

4.2 Recovery of usable products from spent nuclear fuel - Reprocessing of uranium oxide fuelfrom power reactors

4.3 Recovery of usable products from spent nuclear fuel - Reprocessing of fuel fromresearch/materials testing/prototype reactors

5.1 Production of radioisotopes - Manufacture of radioisotopes using nuclear reactors andaccelerators

6.1 Production of radioactive products - Manufacture of radioactive sources, substances andradiopharmaceuticals

7.1 Non-destructive testing - Use of radioactive sources and substances for radiography

8.1 Radiation processing of food - Use of gamma radiation sources to reduce bacterial levels,sterilise, disinfect or modify foods

9.1 Radiation processing of products - Use of gamma radiation sources to reduce bacterial levels,sterilise, disinfect or modify materials

10.1 Substance measurement and process control - Use of sealed sources for thickness gauging,density gauging, mass gauging, level gauging, flow measurement, borehole and well logging, controlof pipeline crawlers

10.2 Substance measurement and process control - Use of neutron sources for moisture gauging

11.1 Detection and analysis - Use of sealed sources for analysis 11.2 Detection and analysis - Use of beta sources for gas chromatography detectors 11.3 Detection and analysis - Use of radioactive sources for leak detection, chemical and

explosives detection 11.4 Detection and analysis - Use of neutron sources for activation analysis

12.1 Elimination of static electricity - Use of radioactive sources to eliminate static electricity

13.1 Illumination - Use and repair of gaseous tritium light sources for illumination, in safety signsand equipment, sighting and location markers, watches and instruments

http://www.defra.gov.uk/environment/radioactivity/government/legislation/pdf/regulations-guidelines.pdf

http://www.defra.gov.uk/environment/radioactivity/government/legislation/pdf/regulations-guidelines.pdf

13.2 Illumination - Use of radioluminous paint (tritium and promethium-147) for the luminising oftimepieces and the repair of radioluminised timepieces

14.1 Electronic apparatus - Use of electronic apparatus containing radioactive substances e.g.tritium in spark gap devices

15.1 Safety devices - Use of ionising radiation in smoke and fire detectors and other safetyinstruments

16.1 Security screening - Use of gamma rays or neutron sources to examine packages, baggage,containers or vehicles

16.2 Security screening - Use of gamma rays to detect people seeking illegal entry to the UK invehicles or freight

16.3 Security screening - Use of back-scatter imaging for the detection of concealed items on theperson

16.4 Security screening - Use of gamma rays or neutron sources to detect concealed items inbuildings

[17 The Environment Agency does not register practices in category 17 since they do not involveradioactive materials.]

18.1 Radioactive tracers - Use of radioactive tracers in industrial process controls 18.2 Radioactive tracers - Use of radioactive tracers for medical or biological techniques 18.3 Radioactive tracers - Use of radioactive tracers for environmental tests

19.1 Diagnosis – medical - Use of ionising radiation in radiography, fluoroscopy, computedtomography, in-vivo nuclear medicine and in-vitro nuclear medicine

20.1 Treatment – medical - Use of ionising radiation in interventional radiology; in-vivo nuclearmedicine; teletheraphy; brachytherapy; radiography (for planning purposes); fluoroscopy (for planningpurposes); computed tomography (for planning purposes)

21.1 Occupational health screening - Use of ionising radiation in radiography and in-vitro nuclearmedicine.

22.1 Health screening - Use of ionising radiation in radiography and in-vitro nuclear medicine

23.1 Medical and biomedical research - Use of ionising radiation in radiography; fluoroscopy;interventional radiography; computed tomography; in-vivo nuclear medicine; in-vitro nuclear medicine;teletherapy; brachytherapy and neutron activation analysis.

24.1 Medico-legal procedures - Use of ionising radiation in radiography; fluoroscopy; interventionalradiography; computed tomography and in-vivo nuclear medicine

25.1 Diagnosis and therapy – veterinary - Use of ionising radiation in radiography, fluoroscopy,computed tomography, in-vivo nuclear medicine, in-vitro nuclear medicine, teletherapy andbrachytherapy

26.1 Teaching, including further and higher education and training - Use of radioactive sources andsubstances

27.1 Research and development - Operation of nuclear fission or fusion reactors for R & Dpurposes

28.1 Ionising radiation metrology - Use of all types of radiation sources to support NationalMeasurement System and use of calibration sources in the testing of equipment

29.1 Storage in transit of radioactive materials

The numbers between 30 and 100 have been left for future use

101.1 Use of NORM - as a chemical reagent 101.2 Use of NORM - as a balance weight 101.3 Use of NORM - as radiation shielding 101.4 Use of NORM - Adventitious arising from gas and oil production 101.5 Other uses of NORM

NORM means naturally occurring radioactive material

102.1 Use by MOD or the armed services 102.2 Use for military purposes by a contractor to the MOD

F2 Please indicate which Associated Activity(ies) are carried out and produce radioactivewastePlease tick each relevant box.This list is intended to give the Environment Agency more detailed information about the production ofradioactive waste.

A Research and development B Manufacture C Repair D Maintenance E Supply F Assembly G Handling H Holding I Testing (operation and quality assurance) J Storage K Use L Decommissioning and waste disposal M Other Please specify

F3 Please describe how the radioactive waste is produced.

F4 Please attach a brief statement covering the following issues (which will be included asconditions if we decide to issue an authorisation):

• A statement of the existence and scope of your management system for compliance withauthorisation requirements.

• A diagram or description of your organisational structure with respect to compliance withauthorisation requirements.

• An indication of the resources available to achieve compliance with authorisation requirements.• Assurance that consultation with an RPA or other qualified expert can take place when

necessary.• Assurance that written operating procedures are in place to cover the accumulation and disposal

of radioactive waste.• The arrangements for adequate supervision of disposal of radioactive waste.

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Radioactive waste is not produced at the premises. The premises are intended to be a final disposal facility for radioactive waste of low specific activity produced from various sources and primarily from the UK civil nuclear decommissioning programme. The premises are an existing permitted PPC hazardous waste disposal landfill. See Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information attached to this application for further details. Secondary waste/emissions could arise from the disposal facility under normal conditions through the management of leachate and the emission of landfill gas. Secondary waste/emissions could arise from the disposal under non-normal/post closure conditions which are assessed in detail in the application supporting information.

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The information is provided in the application document Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information.

Please note that Environment Agency Officers may seek more detailed information on compliancewith relevant authorisation conditions during determination of the application or subsequently.

F5 Please enclose your assessment of how you plan to use best practicable means tominimise the disposal of radioactive waste.See Environment Agency guidance on BPM (in Chapter 4 of RASAG (http://www.environment-agency.gov.uk/business/444304/945840/1064273/?version=1&lang=_e)

F6 Will the radioactive waste be produced for a limited time?Yes, how long?

No

F7 Do you intend to receive and dispose of radioactive waste from another person orpremises?

YesNo

F8 Please give details of each such personCompany 1 Company 2

Address

Postcode

Address

PostcodePhone PhoneFax FaxE-mail E-mailPlease use a continuation sheet if necessary

F9 What is the chemical and physical nature of the waste you intend to receive?

G Incineration on the Premises

G1 Do you intend to use an incinerator on your own premises?YesNo please go to next section

G2 Is there any environmental licence covering the use of your incinerator?Yes please give details

No

G3 What type of incinerator do you have?Please give the manufacturer and model or type number

G4 When was the incinerator installed?



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Indirect radioactive waste arising from leachate management and land fill gas management will arise over the operational period of the landfill and over the aftercare period. The current closure date for the landfill (subject to revision) is 2013 and the current aftercare period extends to completion as defined in the Landfill Regulations.

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See attached information, Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. There are no specific named consignors of radioactive waste at the time of application. Consignors will be established after the disposal route has been authorised and will comprise the nuclear decommissioning industry operating under the NDA programme and other consignors from the UK. The waste route is intended to be open to all potential users in the same manner as the LLWR facility or typical hazardouswaste facilities, with quality assurance for waste receipt established through “conditions for acceptance” derived from the authorisation requirements and established through commercial contracts. This would include all NDA nuclear decommissioning sites, UK nuclear power producing sites, commercial users of radioactivity, hospitals, MOD, the oil/gas industry, legacy wastes, other “small users” and other producers.

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See attached information, Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. The waste will be solid, low specific activity low level radioactive waste that is (in so far as is reasonably practicable) non putrescible and that complies with the non-radiological acceptance criteria for the landfill based upon existing non-radiological risk assessments. The waste may have non-radiological properties that would be classified as inert, non-hazardous or hazardous were the waste not a radioactive waste. The existing landfill is a permitted hazardous waste facility.

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G5 Briefly describe any gas clean-up system or filtration on your incinerator.

H Disposal of gaseous waste

H1 Do you intend to dispose of radioactive waste in the form of gas, mist or dust?Yes Please continue with the rest of this sectionNo Please go to next section

H2 How many discharge points do you intend to use to dispose of gaseous waste?Number of discharge points

Please supply the information in questions H3 to H5 for each discharge point if you have more thanone.

H3 Identify or describe the discharge point.

H4 List the radionuclides you intend to discharge.You should only include intentional and unavoidable discharges of radioactive waste that you expectto need to make after the application of Best Practicable Means to your processes. The EnvironmentAgency does not authorise the accidental release of radioactive material. The quantities specifiedshould be the maximum realistically likely within the normal range of operations.Radionuclide Maximum discharge

in a single day inbecquerels

Maximum discharge ina year in becquerels

Is the proposed annualdischarge greater thanone tenth of therelevant PollutionInventory Threshold?(available from theEnvironment Agency’sweb site)

H5 Maximum number of days in a year on which you intend to discharge

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H6 How do you intend to measure or estimate the activity of the discharge? Please explain themethods to be used and state whether the methods are capable of demonstrating compliance withany proposed discharges greater than one tenth of the Pollution Inventory Threshold.

Assessment

H7 Please attach your radiological assessment of the proposed discharges to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in your workwith radioactive material or waste. For each discharge point you should give details of• the height of the discharge point• the height of the discharge point above the highest part of the nearest building• the discharge rate• details of any filtration on the discharge systemPlease give details of the calculations you use.

I Aqueous Waste

Accumulation of aqueous waste

I1 Do you intend to accumulate radioactive aqueous waste?This includes accumulation of waste to enable short-lived radionuclides to decay.

Yes Please continue with the rest of this sectionNo Please go to next section

I2 Why do you intend to accumulate aqueous waste?It is not common to accumulate aqueous waste before you dispose of it. Please explain why you wantto do it. Any proposed accumulation should be part of the BPM assessment supplied under questionF4.

I3 How do you intend to accumulate aqueous waste?Please explain what facilities and controls you will use to store the accumulated aqueous wastesafely.

I4 How long do you intend to accumulate aqueous waste for?Please give the maximum time that radioactive aqueous waste will be stored from creation or receiptuntil final disposal.

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I5 How much radioactive waste do you intend to accumulate?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examples could be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This willgive you flexibility.Radionuclide Maximum activity in becquerels

I6 What is the maximum volume you intend to accumulate at any one time?

Cubic metres

Disposal of aqueous waste

I7 Do you intend to dispose of radioactive aqueous waste?Yes Please continue with the rest of this sectionNo Please go to next section

I8 What is the chemical and physical nature of the waste you intend to dispose of?

I9 How do you intend to measure or estimate the activity of the discharge? Please explain

I10 Where will you dispose of the radioactive aqueous waste?Please tick all that apply and answer the questions in the relevant section.

to a public sewerdirect to a watercourse or water bodyto your premises’ own sewage treatment worksother method(s)

Disposal to a public sewer

I11 What is the name of your sewerage undertaker?

I12 What is the OS national grid reference of the sewage treatment works discharge point?For example SJ 123 456

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The aqueous waste will be leachate collected from the operating landfill and will have the chemical properties typically associated with landfill leachate from a hazardous waste site. The leachate could conceivably contain leached radioactivity.

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The maximum activity of the discharge has been estimated for risk assessment purposes in Application for “ Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. The estimate is based upon conservative assumptions for overall risk assessment purposes. The actual amount of the discharge is uncertain and will depend on the actual amount of waste disposed in the void at any one time and the fraction of the inventory which transfers to the leachate. The activity in the leachate will be monitored.

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Leachate is currently tankered to an offsite treatment works

I13 What is the total monthly volume of water which you intend to discharge from the premisesinto the sewer?

Cubic metres

I14 What is the maximum monthly total of each radionuclide which you intend to discharge?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.Radionuclide Maximum total activity in any single month

In becquerels

I15 Please attach your radiological assessment of the proposed discharge to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in your workwith radioactive material or waste. Please give details of the calculations you use.

Disposal direct to a watercourse or water body

I16 What is the name of the watercourse or body of water that you intend to discharge into?

I17 Is the body of water a pond or lake?YesNo

I18 What is the OS national grid reference of the discharge point?For example SJ 123 456

I19 What is the maximum volume of water you intend to discharge from the premises in amonth?

Cubic metres

I20 What is the maximum monthly total of each radionuclide which you intend to discharge?Radionuclide Maximum total activity in any single month

In becquerels


Disposal to a sewage treatment works on the premises

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Discharge limits for leachate are indicated within Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information and are subject to agreement.

I22 What is the name of the watercourse or body of water that your sewage treatment worksdischarges into?

I23 What is the OS national grid reference of your sewage treatment works discharge point?For example SJ 123 456

I24 What is the total monthly volume of water which you intend to discharge from your sewagetreatment works?

Cubic metres

I25 What is the maximum monthly total of each radionuclide which you intend to discharge?Radionuclide Maximum total activity in any single month

In becquerels

I26 What do you intend to do with any sludge or solids which are left after treatment?

I27 How do you plan to assess the activity of any sludge or solids which are left after treatmentbefore final disposal?


Disposal of aqueous waste by other methods

I29 Please give details of the method on a separate sheet and attach it to this form, including• a description of the type and quantity of radioactive waste• a description of the disposal route, including water and residual solids• a description of the measurement methods for the radioactivity• a brief summary of any agreement with a contractor and attach it to this form• your radiological assessment of the proposed discharge to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in your workwith radioactive material or waste. Please give details of the calculations you use.

J Organic Liquid Waste

Accumulation of organic liquid waste

J1 Do you intend to accumulate radioactive organic liquid waste?Yes Please continue with the rest of this sectionNo Please go to next section

J2 How do you intend to accumulate organic liquid waste?Please include details of measures and controls used to help keep the waste safe, for examplesecurity, fire precautions and alarms etc.

J3 How long do you intend to accumulate organic liquid waste for?Please give the maximum time that radioactive organic liquid waste will be stored from creation orreceipt until final disposal.

J4 How will you measure the activity of the organic liquid waste?

J5 How much radioactive waste do you intend to accumulate?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.Radionuclide Maximum activity in becquerels

J6 What is the maximum volume you intend to accumulate at any one time?

Cubic metres

Disposal of organic liquid waste

J7 Do you intend to dispose of organic liquid waste?Yes Please continue with the rest of this sectionNo Please go to next section

J8 What is the chemical and physical nature of the waste that you intend to dispose of?

J9 How do you intend to dispose of organic liquid waste?

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Please tick all that apply and answer the relevant questions.incineration on the premisestransfer to a contractorby other means

Incineration of organic liquid on the premises

J10 What is the maximum daily and monthly activity of each radionuclide which you intend toincinerate?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.Radionuclide Maximum activity in becquerels

J11 What is the maximum volume you intend to dispose of in the following periods?Day Month

Cubicmetres

Cubicmetres

J12 How do you intend to assess the activity content of the ash from the incinerator or solidsfrom any filtration system?

J13 How do you intend to dispose of ash from the incinerator or solids from any filtrationsystem?

J14 What will you do if your incinerator fails or breaks down?

J15 Please attach your radiological assessment of the proposed disposal to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in the workwith the radioactive material or waste.You should give details of• the height of the incinerator discharge point• the height of the discharge point above the highest point of the nearest building• the discharge rate• details of any filtration on the incineratorPlease give details of the calculations you use.

Transfer to a contractorPlease provide relevant details for each contractor if you want more than one on the authorisation.

J16 Please attach a brief summary of your agreement with the contractor to this form.

J17 How much radioactive waste do you intend to transfer to your contractor?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.Radionuclide Maximum annual activity

In becquerels

J18 What is the maximum volume you intend to dispose of in any one year?

Cubic metres

J19 What is the company name of the contractor?

J20 What is the address of the contractor’s site which will receive the waste?Address


J21 In which County, borough, district or unitary authority areas is the contractor’s premises?

J22 Please describe contingency arrangements if your planned transfer routes becomeunavailable.For example failure of contractor’s incinerator

Disposal of organic liquid waste by other meansJ23 Please describe any other method you intend to use to dispose of liquid organic waste ona separate sheet. Attach your description to this form.

K Very Low Level Solid WastePlease contact us if you wish to dispose of alpha-emitting radionuclides via this route.

K1 Do you intend to accumulate or dispose of very low level solid waste?Yes Please continue with the rest of this sectionNo Please go to next section

K2 What is the chemical and physical nature of the waste?

K3 What categories of very low level waste do you intend to accumulate or dispose of?VLLW Category 1 waste in which

• there are no alpha-emitting radionuclides• the sum of all radionuclides in any 0.1 cubic metre of refuse is less than 400kBq andless than 40kBq in any one article

VLLW Category 2 (higher limits for Tritium and Carbon 14) waste in which• the sum of all Tritium and Carbon 14 in any 0.1 cubic metre of refuse is less than 4MBq and less than 400 kBq in any one article• there are no other radionuclides

K4 If you are seeking category 2 please tell us why you need these higher limits

K5 How will you measure or assess the activity of the waste?

Accumulation of very low level solid waste

K6 Do you intend to accumulate very low level solid waste?Yes Please continue with the rest of this sectionNo Please go to next section

K7 How much very low level waste do you intend to accumulate at any one time?

Cubicmetres

K8 How long do you intend to accumulate the waste before you dispose of it?The usual time is two weeks

Weeks Where the accumulation time is longer than two weeks please tell us why you need the extra time

K9 How will you store the accumulated very low level waste until it is disposed of?Please give details of measures and controls used to help keep the waste safe, for example security,fire precautions and alarms, etc.

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The application is for the premises to be a disposal facility for controlled burial wastes which would include HV-VLLW wastes. See Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information.

Disposal of very low level solid waste

K10 What is the maximum amount of very low level solid waste you intend to dispose of withnormal refuse in any one month?

Cubic metres

K11 How do you intend to dispose of very low level solid waste?Landfill at a site under your controlCollection by a Local Authority or its contractorTransfer to another contractor for landfill

L Solid Waste (excluding Sealed Sources)

L1 Do you intend to accumulate or dispose of solid waste?Please note that solid waste in the form of sealed sources is covered in Annex S to this form - pleasecomplete that Annex if you need an authorisation to accumulate or dispose of sealed sources. Do notinclude waste which can be disposed of without an authorisation under the terms of an exemptionorder. Waste accumulated to enable short-lived radionuclides to decay should be included.

Yes Please continue with the rest of this sectionNo Please go to the next section

Accumulation of solid waste

L2 Do you intend to accumulate solid waste?This includes accumulation of waste to enable short-lived radionuclides to decay.

Yes Please continue with the rest of this sectionNo Please go to next section

L3 What is the chemical and physical nature of the waste?

L4 How much radioactive waste do you intend to store?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examples could be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This willgive you flexibility.Radionuclide Maximum activity in becquerels Maximum Time of

Accumulation

L5 How much waste do you intend to accumulate at any one time?

Cubic metres

L6 Why are you suggesting this time period(s) for accumulating the waste?

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L7 How will you record and label this solid waste?

Disposal of solid waste

L8 Do you intend to dispose of solid waste?Yes Please continue with the rest of this sectionNo Please go to next section

L9 How do you intend to dispose of solid waste?Please tick all that apply and answer the relevant sections below.

incineration on the premisestransfer to a person authorised to receive themtransfer to a manufacturer or supplier of similar sourcestransfer to Drigg or Sellafield sitescontrolled disposalby other means Please describe any other method you intend to use to dispose of solid waste

on a separate sheet and attach it to this form. You must give relevant details.

Incineration on the premises

L10 What is the maximum daily and monthly activity of each radionuclide which you intend toincinerate?Where one or a few radionuclides dominate the waste you should detail each of them.You must detail all alpha-emitting radionuclides.If you use just a few megabecquerels of similar radionuclides, you can list them as a group. Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.You must indicate which are sealed sources and if any of those are high-activity sealed sources.Radionuclide Maximum discharge

in a single dayin becquerels

Maximum discharge in a monthIn becquerels

L11 How much radioactive solid waste do you intend to incinerate each month?

Cubic metres


L13 How do you intend to assess the activity in the ash from the incinerator and solids fromany filtration system?

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Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information.

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on premises

1

0 Solid waste continuedL14 How do you intend to dispose of ash from the incinerator and solids from any filtrationsystem?

L15 What will you do if your incinerator fails or breaks down?

L16 Please attach your radiological assessment of the proposed disposal to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in your workwith radioactive material or waste. You should give details of:the height of the incinerator discharge pointthe height of the discharge point above the highest point of the nearest buildingthe discharge ratedetails of any filtration on the incineratorPlease give details of the calculations you use.

Transfer to a person authorised under RSA 93 to receive them or a manufacturer or supplier ofsimilar sealed sourcesFor the purposes of this form the person who will be receiving the waste is referred to as “thecontractor”.Give full separate details for each contractor

L17 How much radioactive waste do you intend to transfer to the contractor?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examples could be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This willgive you flexibility.Radionuclide Maximum annual activity

In becquerels

L18 What is the name of the company or organisation which will receive your solid waste?

L19 What is the address of the company or organisation which will receive your solid waste?

PostcodeContact numbers and e-mail

PhoneFaxE-mail

L20 What is the address of the site where solid waste will be sent (if different)?

PostcodeContact numbers and e-mailPhoneFaxE-mail

L21 What is the National Grid Reference Number of the site where solid waste will be sent?

L22 Is the site where solid waste will be sent on a nuclear licensed site (except LLWR Drigg orSellafield)?

YesNo

L23 What is the recipient’s Environment Agency authorisation number for the site where solidwaste will be sent, if known? Not needed for nuclear sites

L24 In which borough, district or unitary authority area is the site where solid waste will besent?If premises are on a boundary please give names of all relevant authorities.Borough or district council or unitary authority

Please give the county council unless there is a unitary authority

L25 Please attach a brief summary of your agreement with any relevant contractor

L26 Please describe contingency arrangements if your planned contractor is unavailable.

Transfer to Low Level Waste Repository (LLWR) Drigg or Sellafield sitesPlease attach a brief summary of your agreement with the site operator to this form.

L27 Will any consignment of waste contain alpha emitting radionuclides in excess of 4gigabecquerels per tonne or all other radionuclides in excess of 12 gigabecquerels per tonne?

YesNo


L29 What is the maximum annual disposal activity (at the time of transfer) for each of thefollowing? in becquerelsUraniumRadium 226 plus Thorium 232Other alpha emittersCarbon 14Iodine 129TritiumCobalt 60Other beta-emitting radionuclides (half-life greater than 3months)Other beta-emitting radionuclides (half-life less than 3 months)

L30 What is the maximum amount of waste you plan to send to the site operator at LLWRDrigg or Sellafield in any one year?

Cubic metre

L31 How many consignments are intended for BNFL at Drigg or Sellafield in a year?

Controlled BurialPlease attach a brief summary of your agreement with the site operator to this form.

L32 How much radioactive waste do you intend to bury at the operator’s disposal site?• Where one or a few radionuclides dominate the waste you should detail each of them.• You must detail all alpha-emitting radionuclides.• If you use just a few megabecquerels of similar radionuclides, you can list them as a group.Examplescould be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give youflexibility.Radionuclide Maximum activity in

any one monthIn becquerels

ConcentrationBq per cubic metre


L34 What is the name of the company or organisation which will receive your solid waste?

L35 What is the address of the company or organisation which will receive your solid waste?

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Please refer to the supporting application document for details.

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The waste will be solid, low specific activity low level radioactive waste that is (in so far as is reasonably practicable) non putrescible and that complies with the non-radiological acceptance criteria for the landfill based upon existing non-radiological risk assessments. The waste may have non-radiological properties that would be classified as inert, non-hazardous or hazardous were the waste not a radioactive waste. The existing landfill is a permitted hazardous waste facility.

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Within our own organisation

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Within our own organisation


L36 What is the address of the site where solid waste will be sent (if different)?


L37 What is the National Grid Reference Number of the site where solid waste will be sent?

L38 What is the recipient’s Environment Agency authorisation number for the site where solidwaste will be sent, if known?

L39 In which borough, district or unitary authority area is the site where solid waste will besent?If premises are on a boundary please give names of all relevant authorities.Borough or district council or unitary authority


L40 Please attach a brief summary of your agreement with any relevant contractor

L41 Please describe contingency arrangements if your planned contractor is unavailable.

L42 Please attach your radiological assessment of the proposed disposal to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in your workwith radioactive material or waste. You should give details of• the disposal arrangements at the disposal site• the type and approximate depth of the overlying material• any measurable radiation dose rates from the closed containers holding the waste.

M NAIR Arrangements

M1 Do you have an emergency role as a participant under the National Arrangements forIncidents involving Radioactivity (NAIR)?No Please go to next sectionYes Please continue with the rest of this section

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TF 010 000

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This authorisation application applies

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East Northamptonshire Council

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Northamptonshire County Council

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M2 Do you wish to have the standard conditions for NAIR participants to be entered into thisauthorisation?This allows you to accumulate waste arising from your participation in the NAIR scheme.

YesNo

M3 Do you have a separate current Variation Notice for the accumulation and disposal of NAIRwaste?

Yes What is the reference number of the Notice?

No

N ChecklistThis section is to help you check that you have• completed the correct parts of the form ( )• attached the right documents to help us process your application quickly ( Ο ).

Company or organisation details Type of application

About the application About the premises Contact details Producing radioactive waste Gaseous waste

Disposal of gaseous waste Discharge point description(s) Ο Radiological assessment of discharge

Aqueous waste Accumulation of aqueous waste Disposal of aqueous waste Disposal to a public sewer Ο Radiological assessment of discharge Disposal direct to a watercourse or water body Ο Radiological assessment of discharge Disposal to a sewage treatment works on the premises Ο Radiological assessment of discharge Disposal of aqueous waste by other methods Ο Radiological assessment of discharge

Organic liquid waste Accumulation of organic liquid waste Disposal of organic liquid waste Incineration on the premises Ο Radiological assessment of discharge Transfer to a contractor Disposal of organic liquid waste by other means Ο Description of method Ο Radiological assessment of disposal

Very low level solid waste Accumulation of very low level solid waste Disposal of very low level solid waste

Solid waste Accumulation of solid waste Disposal of solid waste Incineration on the premises Ο Radiological assessment of disposal Transfer to a contractor

Transfer to LLWR Drigg or Sellafield sites Controlled burial Ο Radiological assessment of disposal Other methods of solid waste disposal Ο Description of method

NAIR arrangements Data handling

Ο Claim of confidentiality Ο National security direction

Payment for your application Ο Cheque

Declaration

O Data Handling

O1 Commercial in confidenceIs there any information in the application which you believe should be restricted on thegrounds that the information relates to a “relevant process” or trade secret?“Relevant process” means any process applied for the purposes of, or in connection with, theproduction or use of radioactive material.

Yes Please describe the information and explain why you believe it should be restricted

No

O2 National securityIs there any information in the application which you believe should be restricted on thegrounds of national security?

YesPlease enclose a copy of any request for a Direction which you have made to the Secretary of Stateor National Assembly for Wales. The Environment Agency already holds a Direction requiring it toensure that no information relating to sealed source applications/registrations is to be included inpublic registers. Pursuant to Section 25(3)(b) of RSA 93 no such information will be sent to LocalAuthorities. Nor will we send similar information relating to accumulation or disposal of radioactivewaste to Local Authorities.

No

O3 Data Protection NoticeThe Environment Agency is responsible for regulating environmental protection, flood defence, waterresources and fisheries. It has a duty to discharge its functions to protect and enhance theenvironment and to promote conservation and recreation.The information provided will be processed by the Environment Agency to deal with your application,to monitor compliance with the licence/permit/registration conditions and to process renewals.

We may also process and/or disclose it in connection with the following:• offering/providing you with our literature/services relating to environmental matters.• consulting with the public, public bodies and other organisations (eg Health and Safety Executive,local authorities, emergency services, DEFRA on environmental issues)• carrying out statistical analysis, research and development on environmental issues• providing public register information to enquirers• investigating possible breaches of environmental law and taking any resulting action• preventing breaches of environmental law• assessing customer service satisfaction and improving our service.• reporting to the European Commission on the experience gained in implementing Council Directive2003/122/Euratom.• exchanging information and co-operation with European Union Member States, third countries orrelevant international organisations.

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Annex S - Sealed Sources

Accumulation and Disposal of Waste Sealed Sources

S1 Do you intend to accumulate or dispose of waste sealed sources?This does not include waste sealed sources which can be accumulated or disposed of under anexemption order without an authorisation. You will be responsible for complying with the conditions ofany such exemption order. It does include accumulation of sealed sources to enable short-livedradionuclides to decay.

Yes Please continue with this AnnexNo Please do not complete this Annex

Accumulation of Waste Sealed SourcesIf your premises is located on a nuclear licensed site please do not complete Questions S2 to S**

S2 Do you intend to accumulate waste sealed sources on the authorised premises?This includes accumulation of sources to enable short-lived radionuclides to decay.

Yes Please continue with this AnnexNo Please describe how you dispose of sources without accumulating them. Then go to

question **

S3 What waste sealed sources do you intend to store at any one time?Including those covered by any existing authorisatons, but excluding sources exempt fromauthorisation.In order, starting with the highest activity material and finishing with the lowest activity material.If you intend to accumulate several sources of the same radionuclide with approximately the sameactivity you can describe them together in a single line in the table below. Refer to the maximumactivity of an individual source. (For example, Caesium-137, three sources, maximum activity for each100 Megabecquerels would cover sources of 75, 85, and 95 megabecquerels activity).You do not need to include radionuclides which are present as a result of radioactive decay of thelisted radionuclides.You may apply for the maximum number of sources that you reasonably expect to accumulate in theforeseeable future (ie the next 1-2 years).If you want to accumulate large numbers of relatively small sources, you can opt to authorise them asa group. (For example, beta/gamma emitting radionuclides, alpha emitting radionuclides.) However, itwill help us process your application if you provide as much information as possible about theproposed individual radionuclides you intend to accumulate. If you do this the maximum activity of anysingle source must not exceed the HASS threshold (see Environment Agency HASS guidance annex)for that radionuclide.Using becquerelsYou should list activity in SI units (becquerels). Write the prefix kilo, mega, giga, tera or peta clearly(in full) to minimise the risk of error.Rounding up substances of nominal activityIf you accumulate radioactive substances of nominal activity (particularly with radionuclides of shorthalf life), you may round up the figure to ensure you do not risk exceeding your authorised limit. If youdo round up a figure, please make sure you say how and where you have done this.Depleted uraniumYou should be aware that some sources may be supplied in depleted uranium containers. Wherenecessary you should give the masses for depleted uranium (for example, in source containers,counterbalance weights) in kilogrammes.Radionuclide Maximum

activity inBecquerels

Maximum time ofaccumulation

Is the source a“new HASS”,existing HASS orother type? *

Number ofWasteSources ofeach type

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*Note – These terms are explained in the Environment Agency’s interim guidance on HASS.(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf).Please put in new, existing or other as appropriate

S4 Why are you suggesting this time period(s) for accumulating waste sealed sources?

S5 How will you record and label the waste sealed sources?

S6 How will you store the accumulated waste sealed sources until they are disposed of?

Security of SourcesThe Environment Agency now has regulatory powers over protective security of certain waste sealedsources. Consideration of security is required for waste high-activity sources and for other sourceswhich, in the opinion of the Environment Agency, constitute a similar level of potential hazard. See theEnvironment Agency's Interim Guidance on HASS.(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf) . It is ouropinion that any source, or aggregation of sources in a single premises, which falls in any of sourcecategories 1 to 4 in the scheme set out in the NSAC Document constitutes a similar level of potentialhazard to a HASS. All users, applicants and other interested parties who need to see the NSACDocument should ask their Police Force Counter Terrorism Security Adviser for a copy.

Where sources are not considered to constitute a similar level of potential hazard to that from high-activity sources, the Environment Agency will be requiring users to take simple precautions to protectthem.

S7 Please provide the following details of the maximum holding of waste sealed sources atany time

Building or Facilityname or number

Radionuclide(s)and Practice(s) fromTable 1 of NSACDocument *

Maximum totalactivity of eachradionuclideGBq

SourceCategory(1- 5) *

SecurityGroup(A – D) *

http://publications.environment-agency.gov.uk/pdf/PMHO0206BKIA-e-e.pdf


* Note – Security Requirements for Radioactive Sources (October 2005), NSAC

The NSAC Document (see Note above) describes how to calculate the category and relevant securitygroup. You should do this on the basis of aggregating all the sealed sources that may be held in asingle building on the premises at any one time (include both registered and waste sealed sources).If you hold HASS or sources of similar level of potential hazard, then we will need to consider whetheror not all of your sources are vulnerable to the same threat, and our assessment of security groupmay differ from your initial one. If this means that you need additional security measures, we will giveyou the opportunity to amend your application. If your assessment of the category of your sourcesindicates that you need significant expenditure to meet the requirements of the NSAC Document, or ifyour sources are distributed around more than one building on the premises, you may considerdiscussing your situation with your local Police Counter Terrorism Security Adviser before completingthis form.

S8 Please confirm that you hold a copy of NSAC Document “Security Requirements for Sitesand Sectors working with Radioactive Substances”, October 2005 and that you understand itsrequirementsThis is available from your local Police Counter Terrorism Security Adviser

YesNo

S9 If you consider your premises to be in Security Groups A, B or C, have you met all of therequirements of the NSAC Document for the security group you consider your premises to be?

YesNo

S10 Please indicate if you have the following security measures in place for the sources bothwhen they are in use and when they are being stored while not being used and waste sources.

General measures to prevent loss of the sources, ie care of sourcesPhysical security:

FenceBuildingRoomStoreSecurity provided by source containerAccess control

Storage of information and databasesSecurity of essential utilities (eg. electricity)Site procedures and security plan to:

Prevent unauthorised access to or loss or theft of the sourcesDetect unauthorised access to or loss or theft of the sourcesInclude options to upgrade the site security plan in response to increased threat

Information security plan covering:Unauthorised access to information on the sourcesUnauthorised access to information on the security measures taken

Personnel checksDetection:

PatrolsAlarmsCCTV

Response to detection:LocalPolice

Documentary evidence of measures takenOther measures

The NSAC Document specifies how to determine which of these security features are required atpremises in Security Groups A, B or C.YOU SHOULD NOT INCLUDE DETAILS OF YOUR SECURITY MEASURES WITH YOURAPPLICATION.

High-Activity Sealed SourcesSee the Environment Agency's Interim Guidance on HASS(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf).You need to complete these questions if you intend to accumulate or dispose of a “new HASS”, orboth new and existing HASS, under the terms of an authorisation under RSA 93. You do not need tocomplete them if you can dispose of your waste HASS without authorisation under the terms of anexemption order.

S11 Please confirm you have read the requirements of the Defra guidance on financial andother provision (High-activity Sealed Radioactive Sources and Orphan Sources Directive (CouncilDirective 2003/122/Euratom) Guidance to the Environment Agency) for each waste high-activitysealed source you intend to accumulate.

YesNo

S12 Which mechanism are you proposing to use for this purpose?You will need to include with the application, sufficient documentation to enable the EnvironmentAgency to assess whether your proposed provision is adequate.

Disposal of waste sealed sources

S13 Do you intend to dispose of waste sealed sources?Yes Please continue with the rest of this sectionNo Please describe what happens to the sources after accumulation. Then leave the

remaining questions blank

Method of Disposal of Waste Sealed Sources

S14 How do you intend to dispose of waste sealed sources?Please tick all that apply and answer relevant questions below in addition to the following generalquestions

1 transfer to a person authorised under section 13 of RSA 93 to receive them2 transfer to a manufacturer or supplier of similar sources3 transfer to a nuclear licensed site except LLWR at Drigg4 controlled burial5 transfer to LLWR at Drigg6 disposal with local authority refuse (in the form of VLLW)7 by other means Please specify and attach full details


http://www.defra.gov.uk/corporate/consult/hass-regs-directions/hass-eaguidance.pdf

http://www.defra.gov.uk/corporate/consult/hass-regs-directions/hass-eaguidance.pdf

Please give the relevant details for each route to be used.

S15 What waste sealed sources do you intend to dispose of?Include those covered by any existing authorisatons, but excluding sources exempt fromauthorisation. Please list them in order, starting with the highest activity material and finishing with thelowest activity material.If you intend to dispose of several sources of the same radionuclide with approximately the sameactivity you can describe them together in a single line in the table below. Refer to the maximumactivity of an individual source. (For example, Caesium-137, three sources, maximum activity for each100 Megabecquerels would cover sources of 75, 85, and 95 megabecquerels activity.)You do not need to include radionuclides which are present as a result of radioactive decay of thelisted radionuclides.You may apply for the maximum number of sources that you reasonably expect to dispose of in theforeseeable future (ie the next 1-2 years).If you want to dispose of large numbers of relatively small sources, you can opt to authorise them as agroup. (For example, beta/gamma emitting radionuclides, alpha emitting radionuclides.) However, itwill help us process your application if you provide as much information as possible about theproposed individual radionuclides you intend to dispose of. If you do this the maximum activity of anysingle source must not exceed the HASS threshold (see Environment Agency HASS guidance annex)for that radionuclide. You must detail all alpha-emitting radionuclides.Using becquerelsYou should list activity in SI units (becquerels). Write the prefix kilo, mega, giga, tera or peta clearly(in full) to minimise the risk of error.Rounding up substances of nominal activityIf you dispose of radioactive substances of nominal activity (particularly with radionuclides of shorthalf life), you may round up the figure to ensure you do not risk exceeding your authorised limit. If youdo round up a figure, please make sure you say how and where you have done this.Depleted uraniumYou should be aware that some sources may be supplied in depleted uranium containers. Wherenecessary you should give the masses for depleted uranium (for example, in source containers,counterbalance weights) in kilogrammes.Radionuclide Maximum annual activity

in becquerels

Disposal by Transfer (Methods 1, 2 or 3 above)

S16 What is the name of the company or organisation whch will receive your waste sealedsources?

S17 What is the address of the company or organisation which will receive your waste sealedsources?

PostcodeContact numbers and e-mailPhoneFax

E-mail

S18 What is the address of the site where waste sealed sources will be sent (if different)?


S19 What is the National Grid Reference Number of the site where sources will be sent?

S20 What is the recipient’s Environment Agency authorisation number for the site where thesources will be sent? If known

S21 In which County, borough, district or unitary authority area is the site where waste sealedsources will be sent?If premises are on a boundary please give names of all relevant authorities.Borough or district council or unitary authority


S22 Please attach a brief summary of your agreement with any relevant contractor

S23 Please describe contingency arrangements if your planned contractor is unavailable.

Disposal of Waste Sealed Sources to LLWR (Low Level Waste Repository) at DriggYou should answer these questions if you intend to dispose of sealed sources to LLWR at Drigg.

S24 Will any consignment of waste sealed sources transferred to the Site Operator of theLLWR at Drigg contain alpha emitting radionuclides in excess of 4 gigabecquerels per tonneor all other radionuclides in excess of 12 gigabecquerels per tonne?

YesNo

S25 What is the maximum annual disposal activity (at the time of transfer) for each of thefollowing? in becquerelsUraniumRadium 226 plus Thorium 232Other alpha emittersCarbon 14Iodine 129TritiumCobalt 60

Other beta-emitting radionuclides (half-life greater than 3months)Other beta-emitting radionuclides (half-life less than 3 months)

S26 What is the maximum amount of waste you plan to send to LLWR at Drigg in any oneyear?

Cubicmetres

S27 How many consignments are intended for LLWR at Drigg in a year?

Disposal of Sealed Sources by Controlled Burial (Method 4)You should answer these questions if you intend to dispose of sealed sources by controlled burial.

S28 What is the name of the company or organisation whch will receive your waste sealedsources?

S29 What is the address of the company or organisation which will receive your waste sealedsources?


S30 What is the address of the site where waste sealed sources will be sent (if different)?


S31 What is the National Grid Reference Number of the site where sources will be sent?

S32 What is the recipient’s Environment Agency authorisation number for the site where thesources will be sent? If known

S33 In which County, borough, district or unitary authority area is the site where waste sealedsources will be sent?If premises are on a boundary please give names of all relevant authorities.Borough or district council or unitary authority


S34 Please attach a brief summary of your agreement with any relevant contractor

S35 Please describe contingency arrangements if your planned contractor is unavailable.

S36 What is the maximum volume in any one year to be sent for burial?

Cubicmetres

S37 Please attach your radiological assessment of the proposed disposal to this form.You should assess the dose to the most likely exposed individual(s) who are not involved in the workwith the radioactive material. You should give details of• the disposal arrangements at the disposal site• the type and approximate depth of the overlying material• any measurable radiation dose rates from the closed containers holding the waste.

Disposal of Sealed Sources in VLLW (Method 6)

S38 What categories of very low level waste do you intend to accumulate or dispose of?VLLW Category 1 waste in which

• there are no alpha-emitting radionuclides• the sum of all radionuclides in any 0.1 cubic metre of refuse is less than 400kBq andless than 40kBq in any one article

VLLW Category 2 (higher limits for Tritium and Carbon 14) waste in which• the sum of all Tritium and Carbon 14 in any 0.1 cubic metre of refuse is less than 4MBq and less than 400 kBq in any one article• there are no other radionuclides

S39 If you are seeking category 2 please tell us why you need these higher limits

S40 How do you intend to dispose of very low level solid waste?Landfill at a site under your controlCollection by a Local Authority or its contractorTransfer to another contractor for landfill

Annex G

Example Capacity Calculation Layout

Example Capacity Calculation Layout The following table is a possible layout for a spreadsheet to administer the radiological capacity of the landfill. A table is required for each of the scenarios that could be restrictive to landfill capacity and the conditions noted in the table must be satisfied for all the scenarios in order for the landfill to have remaining capacity. The table forecasts remaining capacity on the assumption that the “fingerprint” (radionuclide distribution) of the wastestream to be received is the same as that received to date. By changing the “current inventory” to include proposed shipments or hypothetical forecast future waste arisings the table will forecast the remaining capacity (if any) based on the resulting “overall” fingerprint of the waste received to date and to be received in the future. In a finalised version additional columns could be added to archive the current inventory at any particular date and consider proposed shipments separately. Additional features to codify particular shipments and their final location in the landfill could also be added. If the “fingerprint” of waste to be shipped over the life of the facility were known in advance the table would forecast the capacity for each nuclide and the overall capacity. However, since the radionuclide distribution is not known that far in advance (it is known prior to shipment) the table enables ongoing optimisation.

Example Radiological Capacity Table for Scenario “X” Type of Radionuclide Rni

For example Am-241

Specific Dose SDi (microSv yr-1 per MBq) Scenario specific value obtained from Annex B

Current Radiological Inventory of the Landfill Qi

(MBq) This is all waste received to date and could include a future amount proposed to be received to test the remaining capacity is adequate for that shipment

fi The fraction of the overall activity arising from Rni (such that fi=1)

SDi x fi

fi / (Σ SDi fi) x DC RCi (MBq) The radiological capacity for radionuclide Rni

Where DC is the dose constraint (microSv/yr) which is specific to the scenario: DC = 20 microSv/yr or 3000 microSv/yr for intrusion scenarios

RCi - Qi

(MBq)

The remaining radiological capacity for each nuclide Rni based on the waste stream received to date represented by fi. Numbers must remain >0.

Qi,l (MBq) The activity limit for radionuclide Rni if it were the only radionuclide to be disposed of. = DC / SDi

Qi / Qi,l

Qi (MBq) is the actual activity of radionuclide Rni disposed Qi,l (MBq) is the activity limit for radionuclide Rni if it were the only radionuclide to be disposed of.

Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Rni SDi Qi fi = Qi /Σ Qi SDi x fi RCi = fi / (Σ SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l Totals = Σ Qi = fi= 1 = Σ (SDi fi ) = Σ RCi

The total forecast overall radiological capacity of the landfill based on the wastestream received to date represented by fi. For scenario “X”.

= Σ (RCi - Qi) The total forecast remaining radiological capacity of the landfill based on the wastestream received to date represented by fi. For scenario “X”.

Number must remain >0.

= Σ (Qi / Qi,) Must be <=1 for there to be remaining capacity

Annex H

Calculation of dose rate at landfill, TSG(09)0488


Reference: TSG(09)0488

Issue: Issue 2 Technical Services GroupDate: 15th July 2009

CALCULATION OF DOSE RATE AT LANDFILL IN SUPPORT OF A LOW LEVEL WASTE DISPOSAL AUTHORISATION UK-10497

SUMMARY Dose rate calculations were performed in MicroShield to support a low level waste disposal authorisation. An estimate of the dose rate at the landfill site was calculated based on lightly-contaminated rubble being covered by a 30 cm layer of soil material. Dose was found to dependant on the soil material density and largely independent on the distance from the source.

Name and Organisation Signature Date

Prepared By:

Tony Lansdell

TSG

Checked By:

Barry Cook

TSG

Approved By:

Gráinne Carpenter

TSG

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Table of Contents

1 Introduction.......................................................................................................................3 2 Methodology.....................................................................................................................3

2.1 Background ..............................................................................................................3 2.2 Case details..............................................................................................................3 2.3 MicroShield calculation details, uncertainties and assumptions ...............................4

3 Results..............................................................................................................................5 3.1 Cobalt-60 case .........................................................................................................5 3.2 Caesium-137 case....................................................................................................5 3.3 Covering soil material thickness ...............................................................................6

4 References .......................................................................................................................7 Appendix A ...............................................................................................................................8

2


1 INTRODUCTION

Dose rate calculations were required to support a low level waste disposal authorisation. Cases were run using MicroShield v7.02 [1], to determine the dose rate above the layer of lightly contaminated soil at a landfill waste disposal site.

2 METHODOLOGY

2.1 Background

MicroShield was used to determine the maximum resulting dose rate from disposal of soil and rubble to a landfill site. This can be assumed to be uniformly contaminated to 200 Bq/g of either 60Co or 137Cs. An infinite slab of contaminated soil and rubble was assumed to be covered with 30 cm of uncontaminated soil material, and the dose rate assessed.

2.2 Case details

MicroShield was used to model a slab of waste, infinite in horizontal extent, and 100 cm thick. Preliminary study found that if the slab thickness was increased above 50 cm thick, the resulting dose rate was effectively unchanged, and hence a thickness of 100 cm was used to be conservative.

Preliminary studies also indicated that when dose rate was determined on contact, 1m and 2m above the shielding soil layer, dose rate was independent of dose point height and so only the contact dose was reported. This work can be found in Appendix A.

It has been outlined that MicroShield did not correctly include the effects of build-up (scattered flux) when using the infinite slab geometry, and that the calculation was instead performed using a (finite) rectangular slab that was chosen to have a very large extent such that it was effectively infinite. The extent was chosen such that the results were unchanged with further increases in size, and it was found that beyond 200 cm in width the dose rate on contact was effectively constant, and 1000 cm was chosen to be conservative.

After these initial tests, on the assumption that dose rates were in the worst case for each nuclide greater than 2.5 µSv/hr, it was to be found what thickness of soil material would result in a dose rate of 2.5 µSv/hr.

3


2.3 MicroShield calculation details, uncertainties and assumptions

Energy deposition to dose rate conversion was performed automatically in MicroShield using built-in tables of effective dose rate, taken from ICRP-51 [2]. This presents a series of possible dose rates depending on the assumed irradiation geometry. The highest biological dose rate is produced assuming anterior-posterior geometry (with the gamma rays entering a person from the front and exiting through the back), and to be conservative it was this maximum dose rate that was reported. Dose rates can vary by approximately 30%, depending on which geometry is assumed.

MicroShield approximates the contribution of scattered radiation to the resulting dose rate by the use of build-up tables. The dose rate is dependant on which material is chosen as the dominant scattering medium. In accordance with the MicroShield manual, the material containing the highest number of gamma ray mean free paths should be used as the build-up material – hence in these cases, the source was chosen as build-up material. It was found that choosing the shielding soil as the build-up material produced identical results; hence the results are insensitive to this assumption.

MicroShield uses a point-kernel integration technique to determine the dose rate. This involves splitting the geometry into pieces (kernels). The quadrature order of the calculation determines the number of kernels used and hence the accuracy of the approximation, at the expense of a longer calculation time. Due to the extent of the source relative to the dose rate distance, the quadrature order was increased in the y and z axes until the result was unchanged. Beyond a quadrature of 30, the results were unchanged, and 50 was used to be conservative.

In all cases assessed, the ‘contact’ dose rate point was actually positioned at 1 cm from the surface, as the method of calculation used by MicroShield is known to become unstable at distances closer than 1 cm. The dose rate was found to be nearly independent of distance, with only a 1-2% drop in dose rate from contact to 2m, hence the results are insensitive to this assumption as well.

4


3 RESULTS

3.1 Cobalt-60 case

The contact dose rate from high density (2 g/cm3) soil containing 200 Bq/g 60Co covered with 30 cm of uncontaminated soil was determined for a series of soil material densities from 1.0 to 1.6 g/cm3. The results are given in Table 1.

Soil material density (g/cm3)

Contact dose rate (µSv/hr)

1.0 18.35 1.2 13.02 1.4 9.28 1.6 6.63

Table 1: Contact dose rates for various soil material densities

The resulting dose rate above the shielding layer of soil will be between 18.35 µSv/hr for loose soil and 6.63 µSv/hr if the shielding surface soil has been compacted.

3.2 Caesium-137 case

The contact dose rate from high density (2 g/cm3) soil containing 200 Bq/g 137Cs covered with 30 cm of uncontaminated soil was determined for a series of soil material densities from 1.0 to 1.6 g/cm3. 137Cs is a beta emitter. Its daughter, 137mBa is the source of the gamma radiation. Where a source containing 137Cs was specified, its daughter product 137mBa was also included in equilibrium concentration with 137Cs. Since the half-life of 137mBa is short (2.5 minutes), it will almost always be found in equilibrium with its parent radionuclide. The results are given in Table 2.

Soil material density (g/cm3)

Contact dose rate (µSv/hr)

1.0 2.58 1.2 1.67 1.4 1.08 1.6 0.70

Table 2: Contact dose rates for various soil material densities

The resulting dose rate above the shielding layer of soil will be between 2.58 µSv/hr for loose soil and 0.70 µSv/hr if the shielding surface soil has been compacted.

5


3.3 Covering soil material thickness

Following on from the scenario outlined in Section 3.1, the contact dose rate from high density (2 g/cm3) contaminated soil and rubble containing 200 Bq/g 60Co at a density of 2 g/cm3 covered by uncontaminated soil of various thickness was determined to find a relationship between the shielding material thickness and dose rate. The density of the covering soil material was taken as 1 g/cm3, which presented the worst case in section 3.1. and show the relationship between dose and soil material thickness.

Figure 1

Figure 1: The relationship between soil thickness and contact dose rate for soil and rubble contaminated by 60Co isotopes resulting in a uniform activity of 200 Bq/g

Table 3

Table 3: Tabulated data for Figure 1

0

2

4

6

8

10

12

14

16

18

20

25 35 45 55 65 75 85

Soil thickness (cm)

Dose

Rat

e (u

Sv/h

r)

Soil Thickness (cm) Dose Rate (µSv/hr) 30 18.35 35 13.78 40 10.39 45 7.84 50 5.93 55 4.49 60 3.40 65 2.58 70 1.96 75 1.48

These data show that to get a dose rate of 10 µSv/hr, the soil material must be at least 40cm thick, and to get a dose rate of 2.5 µSv/hr, the soil material must be at least 65cm thick [3].

6


4 REFERENCES

1 MicroShield v7.02, Grove Software Inc, 2007 2 ICRP-51 (1987) Data for use in protection against external radiation 3 Personal communication, Paul Atyeo, 23rd March, 2009

7


APPENDIX A – Calculations to show that dose is geometry and air distance independent Depth: 200cm Length: 2000cm Breadth: 2000cm Based on the soil material having a thickness of 30cm and a density of 1 g/cm3

Dose point Dose (µSv/h) ‘Contact’ 18.35 1m 18.21 2m 17.82

8

Annex I

Baseline Groundwater and Leachate Sample Results

Report Determination of 238U, 235U, 234U, 232Th, 230Th, 228Th, 226Ra, 3H and gross alpha and gross beta in 8 water samples. (Samples: KO2A etc…)

UKAEA Harwell Customer Jon Blackmore

UKAEA B175 Harwell International Business Centre Didcot Oxfordshire OX11 0RA

Customer reference number Quote620 GAU job number GAU1278 (Final) Date samples received 18th August 2008 Report date 1st October 2008 Report produced by Dr P. Gaca

(Radiochemist, GAU-Radioanalytical) Signed

Report checked by Signed

Job reference number

GAU1278 (Final)

Geosciences Advisory Unit, National Oceanography Centre, Southampton, European Way, SO14 3ZH Page 2 of 8 1/10/08

C:\Documents and Settings\carolearp\Desktop\Permit Application - LLW\Annex I.doc

Methodology Samples were received at the National Oceanography Centre, Southampton on 18th August 2008 in good condition. Gamma spectrometry (Method GAU/RC/2032: Accredited to ISO/IEC 17025:2005) 100ml of the sample was evaporated down to less than 20ml and transferred to a scintillation vial. The sample was then counted on a well-type HPGe detector previously calibrated with a mixed nuclide standard of identical geometry. The resulting spectrum was analysed using Fitzpeaks spectral analysis software. All anthropogenic radionuclides were identified and quantified. In addition 60Co, and 137Cs were specifically searched for and limits of detection reported where no activity was detected. Gross alpha / beta in waters (Method GAU/RC/2034) 200 ml of the sample was acidified with H2SO4 and evaporated to dryness and the residue ignited at 350 C. The ignited residue was ground and mounted onto a 47 mm filter paper. The source was then counted on a gas flow proportional counter previously calibrated against 241Am (alpha) and 137Cs (beta). 3H in aqueous samples (Method GAU/RC/2004) 50ml of the sample was removed for 3H analysis. The sub-sample was purified by distillation. The 3H content of the distillate was then measured using a Quantulus ultra-low level liquid scintillation counter. 226Ra in aqueous samples (Method GAU/RC/2038) An aliquot of the aqueous sample is mixed with a water-immiscible scintillation cocktail in a glass vial. The vial is sealed and immediately counted on a Perkin Elmer Quantulus liquid scintillation counter with alpha-beta discrimination activated to determine the total 222Rn activity. The sample is then stored for two weeks and recounted to determine the activity of supported 222Rn/226Ra. Th isotopes by alpha spectrometry (Method GAU/RC/2027) An aliquot of the sample is spiked with 229Th and acidified. An iron hydroxide precipitation followed by anion exchange chromatography is used to isolate Th from the solution. The activities of 230Th and 232Th are then determined by alpha spectrometry. U by alpha spec & ICPMS (Method GAU/RC/2026) An aliquot of the sample is spiked with 232U and acidified. A combination of anion exchange and extraction chromatography is used to isolate U from the solution. 238U and 234U are determined by alpha spectrometry, and the 235U content is determined relative to 238U by ICP-MS.


GAU1278 (Final)



Limits of detection / quantification For gamma data, limits of quantification, LQ, is calculated as defined by Currie (1968) and Gilmore & Hemingway (2000)

ggammaQ MYEt

C

m

nL

11001001

214115.0

22

)(

.

where is set at 2.00, C is the background counts, n is the number of channels covering the peak, m is the number of background channels taken either side of the photopeak, t is the count time in seconds, E is the counting efficiency, Y is the gamma emission probability and Mg is the mass of sample analysed in grams Limits of detection for H-3 analyses are quoted as LD as defined by Currie, 1968.

g

D MREt

CgBqL

110010065.471.2)/(

where C is the background count, t is the count time in seconds, E is the measurement efficiency, R is the chemical recovery and m is the sample mass in grams. References Currie L.A. (1968). Limits of qualitative detection and quantitative determination. Anaytical Chemistry, 40 (3), 586-593. Gilmore G. and Hemingway J. (2000). Practical gamma-ray spectrometry. John Wiley, Chichester, UK


GAU1278 (Final)



Summary of samples and results All uncertainties quoted are propagated method uncertainties unless otherwise stated. * Indicates results obtained using an accredited method.

GAU ID Customer ID Sample type GAU1278/1 KO2a Water GAU1278/2 KO3 Water GAU1278/3 KO5 Water GAU1278/4 KO6 Water GAU1278/5 KO7 Water GAU1278/6 KO8 Water GAU1278/7 KCLW2A2 Water GAU1278/8 KCLW3A1 Water

Results Gross alpha/beta

GAU ID Gross alpha

[Bq/L] +/-

Gross beta [Bq/L]

+/-

GAU1278/1 <0.1 - 0.46 0.15 GAU1278/2 <0.1 - <0.2 - GAU1278/3 <0.2 - <0.3 - GAU1278/4 <0.2 - <0.3 - GAU1278/5 <0.2 - <0.3 - GAU1278/6 <0.1 - <0.2 - GAU1278/7 <2 - 90 3 GAU1278/8 <1 - 20 1

Coverage factor k=2 S.D. Uncertainties quoted are propagated method uncertainties


GAU1278 (Final)



3H

GAU ID 3H [Bq/L] +/- GAU1278/1 <5 - GAU1278/2 <5 - GAU1278/3 <5 - GAU1278/4 <5 - GAU1278/5 <5 - GAU1278/6 <5 - GAU1278/7 59 7 GAU1278/8 10 4

Coverage factor k=2 S.D. Uncertainties quoted are propagated method uncertainties 226Ra

GAU ID 226Ra [Bq/L] +/- GAU1278/1 0.30 0.07 GAU1278/2 0.29 0.07 GAU1278/3 0.29 0.07 GAU1278/4 0.30 0.07 GAU1278/5 0.35 0.07 GAU1278/6 0.33 0.07 GAU1278/7 0.34 0.07 GAU1278/8 0.58 0.08

Coverage factor k=2 S.D. Uncertainties quoted are propagated method uncertainties


GAU1278 (Final)



238U, 235U, 234U

GAU ID 238U

[Bq/L] +/-

235U [Bq/L]

+/- 234U

[Bq/L] +/-

GAU1278/1 0.039 0.012 <0.005 - 0.066 0.013

GAU1278/2 0.042 0.009 <0.005 - 0.039 0.009

GAU1278/3 0.018 0.005 <0.005 - 0.018 0.005

GAU1278/4 0.016 0.005 <0.005 - 0.028 0.006

GAU1278/5 0.014 0.005 <0.005 - 0.021 0.006

GAU1278/6 0.013 0.005 <0.005 - 0.010 0.006

GAU1278/7 <0.01 - <0.005 - <0.01 -

GAU1278/8 <0.06 - <0.005 - <0.05 - Coverage factor k=2 S.D. Uncertainties quoted are propagated method uncertainties 235U activity concentration calculated using 238U/235U ratio obtained with ICP-MS measurement. 232Th, 230Th, 228Th

GAU ID 232Th

[Bq/L] +/-

230Th [Bq/L]

+/- 228Th

[Bq/L] +/-

GAU1278/1 <0.003 - <0.003 - <0.005 -

GAU1278/2 <0.002 - 0.0024 0.0014 <0.004 -

GAU1278/3 <0.004 - 0.0030 0.0018 <0.004 -

GAU1278/4 <0.003 - <0.002 - <0.006 -

GAU1278/5 <0.003 - 0.0027 0.0018 <0.008 -

GAU1278/6 <0.003 - 0.0026 0.0016 0.0059 0.0022

GAU1278/7 <0.008 - <0.005 - <0.007 -

GAU1278/8 <0.002 - 0.0043 0.0020 0.013 0.003 Coverage factor k=2 S.D. Uncertainties quoted are propagated method uncertainties


GAU1278 (Final)



Gamma Spectrometry* Artificial Radionuclides

GAU ID 241Am +/- 60Co +/- 137Cs +/- 154Eu +/- 54Mn +/- 65Zn +/-

GAU1278/1 <1 - <3 - <2 - <30 - <1 - <3 - GAU1278/2 <1 - <2 - <1 - <20 - <1 - <3 - GAU1278/3 <1 - <3 - <2 - <30 - <2 - <4 - GAU1278/4 <1 - <3 - <1 - <20 - <1 - <4 - GAU1278/5 <0.9 - <2 - <1 - <20 - <1 - <3 - GAU1278/6 <0.9 - <2 - <1 - <20 - <1 - <3 - GAU1278/7 <1 - <2 - <1 - <20 - <1 - <3 - GAU1278/8 <0.7 - <2 - <0.9 - <20 - <0.9 - <3 -

*Indicates results obtained using an accredited method. Results are quoted in Bq/L. Coverage factor k=2 S.D. Reference date: 18/08/08


GAU1278 (Final)



Gamma spectrometry* Natural Radionuclides GAU ID 228Ac +/- 40K +/- 210Pb +/- 212Pb +/- 214Pb +/- 226Ra +/- 208Tl +/- 234Th +/- 235U +/-

GAU1278/1 <10 - <30 - <10 - 1.4 0.6 <3 - <20 - <10 - <10 - <4 -

GAU1278/2 <10 - <30 - <10 - <1 - <4 - <20 - <10 - <10 - <4 -

GAU1278/3 <10 - <40 - <10 - <1 - <4 - <20 - <10 - <10 - <4 -

GAU1278/4 <10 - <30 - <10 - <2 - <4 - <20 - <10 - <20 - <4 -

GAU1278/5 <7 - 18 9 <10 - <1 - <3 - <10 - <9 - <10 - <4 -

GAU1278/6 <7 - <30 - <10 - 1.3 0.6 <3 - <10 - <9 - <10 - <4 -

GAU1278/7 <20 - 93 15 <10 - 4.2 0.8 <4 - <20 - <10 - <20 - <4 -

GAU1278/8 <6 - 18 8 <10 - 2.3 0.7 <2 - <10 - <10 - <10 - <3 -

*Indicates results obtained using an accredited method. Results are quoted in Bq/L. Coverage factor k=2 S.D. Reference date: 18/08/08

Annex J

Capability Statements

Research Sites Restoration Limited (RSRL) (also referred to as UKAEA Harwell)

RSRL provided Augean with technical support in relation to the Low Level Wastes from the perspective of a waste producer and consignor. RSRL attended the public exhibition to provide information to the public on the wastes.

Research Sites Restoration Limited (RSRL) is the site licence company responsible for the closure programme at Harwell and Winfrith. Winfrith was a major centre for groundbreaking reactor development from the late 1950s to the 1990s whilst Harwell’s origins go back to the dawn of the UK’s nuclear industry in the 1940s.

RSRL is a wholly-owned subsidiary of UKAEA (the United Kingdom Atomic Energy Authority) and operates under contract to the Nuclear Decommissioning Authority (NDA).

UKAEA

The UKAEA group is a world leader both in decommissioning and regenerating nuclear sites and in developing fusion as a sustainable, secure and carbon-free energy source. With decades of experience as pioneers in these fields, UKAEA is making a key contribution to meeting the twin challenges of sustainable development and climate change.

UKAEA has over 50 years’ experience in nuclear site management, operations and decommissioning. Through projects spanning the nuclear lifecycle, UKAEA provides industry-leading technical, design, engineering, safety, and programme and project management consultancy services to organisations around the world.

The UKAEA Ltd. subsidiary of the UKAEA group provided technical assessments in support of the authorisation application for the proposal.

The involvement of RSRL was fronted by Paul Atyeo.

Paul Atyeo, RSRL

Paul is a Chartered Mechanical Engineer and Chartered Environmentalist with a first degree in Mechanical Engineering and a masters degree in Business Administration. Paul has worked in the nuclear industry for 21 years, specialising in nuclear reactor experimental systems, land remediation, nuclear waste management, nuclear decommissioning and site delicensing. Paul currently manages decommissioning of the Harwell nuclear site.

Peter Shaw

Peter Shaw Job Title – Group Leader, Consultancy Development Group

Qualifications

1979 HNC Chemistry

1986 Post Graduate Course in Radiological Protection (PGRP)

1993 Advanced Course in Radiological Protection (ARP)

1999 Diploma in Pollution Control

Professional Qualifications

2000 Member, Chartered Society for Radiological Protection (CRadP)

March 2001 and February 2006 - RPA2000 Certificate of Competence to be a Radiation Protection Adviser

Key skills

Expert in radiological protection. International expertise in ALARA. Special expertise in NORM.

Input to national and international standards

Communications, delivering presentations at national and international events, chairing meetings, etc.

Emergency response adviser

Project and team management

Customer liaison

Profile

Peter Shaw began his career with the National Radiological Protection Board in 1979, where he spent his formative years developing a sound grounding in the fields of health physics and radiological protection, including metrology, dosimetry, radiochemistry and radiological assessments. He has developed his expertise over the last 30 years, to become a well respected expert in radiological and environmental protection and hazard assessment. He develops and delivers professional level training modules for customers operating in the non-nuclear industrial sectors, and sits on a number of internal and external committees dedicated to developing expertise in this very specialised area. He sits on national and international committees and provides input to radiation protection standards. He is a highly qualified and experienced Radiological Protection Adviser and participates in national emergency exercises at off-site control centres. In addition to his well respected technical expertise, he is an accomplished manager, managing a multi-disciplinary team of technical employees as well as a portfolio of projects for internal and external customers.

Career history

Group Leader HPA, Leeds 1984–present

Consultancy Development Group

Manages the Consultancy Development Group within HPA, with specific responsibility for the

Peter Shaw

development of radiation protection services, including the Radiation Protection Advisor (Qualified Expert) service as well as various support activities.

A certificated RPA with extensive experience of advising on the use of industrial radioactive sources and x-ray equipment including gauging, non-destructive testing, security and analytical equipment. Also specialises in advising users of unsealed radioactive materials and NORM.

Extensive experience in the provision of radiation protection training at all levels. Provides professional level training to internal and external customers. Has managed the development and running of the HPA Radiological Protection Training Scheme Module on “Principle for Protection against Internal Radiation Sources”.

Participates in nuclear emergency exercises at off-site control centres and also participates in the multi-agency CBRN preparedness group for local government. Contributes to the development of national security standards for radioactive sources. Assists in the investigation of international radiological accidents.

Provides input to radiation protection standards and guidance at a national and international level. Currently Secretary of the European ALARA Network.

Scientific Officer National Radiological

Protection Board, Leeds 1979–1984

Operation and (later) management of radiation protection services, including metrology, external and internal dosimetry, radiochemistry and consumer product testing.

Experienced in developing environmental transfer.models and undertaking assessments of the radiological impact (public and worker) from the release of radionuclides into the environment.

Undertook environmental measurements and sampling following radiological accidents.

.

Galson Sciences Limited July 2009 Dr. Roger D. Wilmot, BA, PhD Principal Consultant Dr. Roger D. Wilmot has degrees in Earth Sciences from Cambridge University (BA) and Imperial College, London (PhD). He is a geologist with over 20 years experience in providing a broad range of research, consultancy and management services to a range of clients, starting with site characterisation work for the four proposed UK shallow sites in the 1980s. Dr. Wilmot is chair of SSM’s OVERSITE international panel responsible for regulatory review of SKB’s risk assessments for radioactive waste disposal in Sweden. He has also worked fro the Swedish regulators on Quality Assurance, development of a strategy for consideration of future human actions in assessments and the conduct of risk assessments. Dr Wilmot provided technical support to the Environment Agency and SEPA in the recent revision of the agencies’ Guidance on Requirements for Authorisation, and was part of the management team for the review on behalf of the Environment Agency of BNFL’s safety case for the LLWR. Dr. Wilmot developed a computer code for UKAEA to undertake radiological performance assessments of waste disposal and storage facilities, and assessments of the impact of radioactivity in the environment. He has led a variety of waste management options appraisals for UKAEA Dounreay, including an evaluation of the transport of radioactive waste, as part of BPEO studies. He authored a draft ESC for the Pits facility at Dounreay. Dr. Wilmot was responsible for development, implementation and trial application of an methodology for assessing the doses associated with landfill sites for Special Precautions Burial of LLW, and has extended and used this methodology for PA of on-site disposal of radioactive wastes and for assessing the dose implications of dustbin disposal of VLLW. Dr. Dev Reedha, BEng(Hons), PhD Senior Consultant Dr. Dev Reedha has degrees in Mechanical Engineering and Energy Systems (BEng First Class Honours) and Engineering (PhD) from the University of Manchester. He has six years experience in radioactive waste management and nuclear consultancy and research, and ten years experience in computational fluid dynamics, mathematical modelling and simulation of fluid flow in sedimentary units, and in computer code development. He also has experience in software consultancy. Dr. Reedha has carried out key technical work in radioactive waste management. On behalf of the UK environmental regulators (through SNIFFER), he developed a computational model to evaluate allowable disposals of radioactive waste to landfill, and provided technical support for the evaluation of doses associated with VLLW disposals. On behalf of Defra, he provided technical support on a project to advise Government on the feasibility of gathering data on the geographical generation of non-nuclear LLW, including VLLW, within the UK. For ONDRAF-NIRAS, he has reviewed the treatment of concrete degradation in safety assessments, and is currently working (lead author) on the Belgian Category A inventory report. Dr. Reedha is an experienced groundwater flow and contaminant transport modeller using state-of-the-art software packages such as FEFLOW and GoldSim-RT. He contributed to an assessment of post-closure safety of the LLW disposal facility near Drigg, on behalf of the Environment Agency. For DSRL, he worked on the development of a risk assessment model in GoldSim-RT for solid LLW disposal at Dounreay, and is currently undertaking hydrogeological modelling analysis of the proposed LLW facilities using FEFLOW. On behalf of the Nuclear Decommissioning Authority’s Radioactive Waste Management Directorate (NDA RWMD), he recently undertook transient three-dimensional groundwater flow calculations using FEFLOW to evaluate potential hydrological interactions in a geological disposal facility during the operational phase and after facility closure.

July 2009 Page 1 of 2

NAME: GENE BARRY WILSON BORN: 1957 NATIONALITY: British QUALIFICATIONS & PROFESSIONAL AFFILIATIONS: Doctor of Philosophy - Imperial College London Diploma of Imperial College - Imperial College London B.Sc. Honours in Botany/Genetics - University College Cardiff Chartered Town Planner; Member of the Royal Town Planning Institute Chartered Waste Manager; Member of the Chartered Institution of Wastes Management Chartered Biologist; Member of the Institute of Biology Chartered Environmentalist Member of the Institute of Quarrying Member of the Institute of Ecology and Environmental Management Registered Principal Environmental Auditor CAREER SUMMARY:

2005 – Present: Group Technical Director, Augean plc Dr Wilson is an experienced environmental manager with particular expertise in quarrying and waste management together with skills in industrial and applied ecology. Dr Wilson is responsible for the planning and permitting strategy and delivery for the two divisions of the Group, Landfill and Treatment. This involves regular interface with regulator bodies and the public. A key part of his role is monitoring and advising on compliance and regulatory matters for the Group. Dr Wilson manages a team of auditors and monitoring technicians who continuously assess the Group’s environmental and health and safety performance. The results of these assessments are reported annually in the Group Corporate Responsibility report.

Dr Wilson is responsible for the management and monitoring of the Group’s Integrated Management System which satisfies the requirements of ISO 14001 (Environmental Management System Standard), ISO 9001 (Quality Management System Standard) and OHSAS 18001 (Health, Safety and Welfare Management System Standard).

Dr Wilson manages a team of highly trained chemists within our laboratory services who provide our clients with accurately assessed data to identify and understand the nature of the waste they produce which then allows us to offer the best management solution. Dr Wilson actively engages with the industry, regulators and government departments at a national level promoting high standards and new technologies for the sector. He is a member of the Regulatory and Planning Committees of the Environmental Services Association and regularly comments on planning policy and technical guidance notes. He is also a member of the DEFRA Hazardous Waste Steering Group which is developing a strategy for the modernisation of the sector.

July 2009 Page 2 of 2

1987 - 2005: Director of Environmental Planning and Principal Ecological Consultant with MJCA

Dr Wilson directed the Company services in planning, environmental assessment, environmental audit, environmental management systems and waste management licensing bringing his strong organisational skills to manage successfully these complex environmental projects. Dr Wilson is an experienced expert witness and has given evidence on the subjects of minerals and waste management operations, planning policy, need and ecology. Dr Wilson was responsible for the Environmental Planning services of the Company which include town and country planning, minerals planning and in particular waste management planning. An essential part of his role at MJCA was the assessment of local policy and strategy in matters of land use and waste management and he has considerable experience of the structure, analysis and use of local plans. A significant proportion of Dr Wilson's work involved the preparation and negotiation of environmental assessments and planning applications where his knowledge of environmental science and understanding of the technical and practical demands of industrial development are critical. Dr Wilson is an experienced environmental manager who has assisted a range of companies from single site to multinational in the development and implementation of environmental management systems. Working closely with the client company he undertook environmental reviews, developed environmental policies, prepared manuals and codes of practice and provided advice on ISO 14001 and EMAS. Dr Wilson has managed several hundred environmental audits of companies and their facilities to demonstrate compliance with legislation and environmental policy, for the purpose of due diligence prior to acquisition and for insurance purposes. Dr Wilson is an experienced applied and industrial ecologist and provided advice on matters such as reclamation, conservation, bioengineering and landscape planting with a strong emphasis on the integration of land development with vegetation and wildlife. Dr Wilson has particular expertise in the restoration of quarries and landfill sites and in habitat creation. He lectured regularly to the waste management industry on reclamation and chaired the landfill reclamation course run by the Environmental Services Association.

1983 - 1987: Research Scientist at the School of Agriculture, Nottingham University

Dr Wilson was responsible for the design and management of experiments and reclamation procedures, including the propagation and establishment of native plant species in a fully active dolerite quarry in Wales, the organisation of surveys to describe native plant communities in the areas adjacent to the quarry and evaluation of their conservation status and the monitoring of meteorological and edaphic factors on field sites. Dr Wilson liaised closely with conservation bodies, local councils, landowners and in particular the sponsoring quarry company for whom periodic reports were produced and lectures given. This work culminated in the production of a manual for the continuing reclamation and conservation of the site.

AUGEAN SECTION 73 EAST NORTHANTS RESOURCE …€¦ · environmental statement section 73...

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