Title: SITE EVALUATION OF THE OPG NEW NUCLEAR AT ...€¦ · 4.1 Derivation of Accident Scenario...

OPG Proprietary Document Number:

NK054-REP-01210-00016 Sheet Number: Revision:

Coversheet

N/A R001 Title:

SITE EVALUATION OF THE OPG NEW NUCLEAR AT DARLINGTON - PART 2: DISPERSION OF RADIOACTIVE MATERIALS IN AIR AND WATER

Associated with REP N-TMP-10179-R001 (Microsoft® XP)

© Ontario Power Generation Inc., 2009. This document has been produced and distributed for Ontario Power Generation Inc. purposes only. No part of this document may be reproduced, published, converted, or stored in any data retrieval system, or transmitted in any form or by any means (electronic, mechanical, photocopying, recording, or otherwise) without the prior written permission of Ontario Power Generation Inc.

Prepared by: [AMEC NSS – Per attached] Verified by: [AMEC NSS – Per attached] Elizabeth Alderson

Badi-Uz-Zaman Khan Arindam Mukherjee Mark Gerchikov Jimin Peng AMEC NSS

Date Yung Cheung AMEC NSS

Date

Reviewed by: [AMEC NSS – Per attached] Reviewed for QA by: [AMEC NSS – Per attached]

Michael Lee Robert Zimmermann Marcello Oliverio Richard J. Fluke AMEC NSS

Date Charles Gordon AMEC NSS

Date

Approved by: [AMEC NSS – Per attached] Ron Henry

AMEC NSS Date

Recommended for OPG Acceptance by:

Accepted for OPG Use by:

John Marczak Manager Safety Analysis Review Dept Darlington New Nuclear Project

Date Bob Goodman Director - Engineering Darlington New Nuclear Project

Date

Site Evaluation of the OPG New Nuclear at Darlington - Part 2: Dispersion of Radioactive

Materials in Air and Water

NK054-REP-01210-00016-R001

2009-08-31

Project ID: 10-27600 AMEC NSS: P1093/RP/001 R05

OPG Proprietary

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AMEC NSS Limited

P1093/RP/001 R05 Page 2 of 132 Form 114 R14

Confidentiality, Copyright and Intellectual Property Notice 2009

This document and its contents are strictly confidential. It has been produced by AMEC NSS Limited

for Ontario Power Generation Inc. (OPG) under the terms of the Nuclear Safety and Technology Support and Services Agreement dated December 13, 2005 between OPG and AMEC NSS.

Rights of copying and of ownership and use of the intellectual property in or embedded in this document are solely determined by the terms of such Agreement.

No part of this document shall be used, reproduced, published, converted or stored in any data retrieval system or transmitted in any form or by any means (electronic, mechanical, photocopying,

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If you are not the intended recipient please notify the Contracts Manager, AMEC NSS, (416) 592 4094

or return by post to AMEC NSS Limited, 700 University Avenue H4, Toronto, Ontario M5G 1X6.

Revision Summary

Rev Date Author Comments

R00 August 25, 2008 Mark Gerchikov et al. Framework Report Issued to Client

R01 November 03, 2008 Mark Gerchikov et al. Report Issued to Client

R02 January 21,2009 Mark Gerchikov et al. Final Report Incorporating Client’s Comments

R03 February 27,2009 Mark Gerchikov et al. Updated Report Issued to Client

R04 March 20,2009 Mark Gerchikov et al. Updated Report Incorporating Client’s Comments

R05 August 2009 Mark Gerchikov et al. Revised to address internal and external client comments

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Table of Contents

Page

EXECUTIVE SUMMARY................................................................................................... 10

1.0 INTRODUCTION .................................................................................................. 13

1.1 Project and Task Background....................................................................................... 13

1.2 Objectives .................................................................................................................. 14

1.2.1 Normal Operations ...................................................................................................... 14

1.2.2 Accidents ................................................................................................................... 15

1.3 Regulatory Framework Overview .................................................................................. 15

1.3.1 CNSC RD-346 ............................................................................................................. 15

1.3.2 IAEA NS-G-3.2 ............................................................................................................ 15

1.3.3 IAEA NS-R-3............................................................................................................... 16

1.3.4 CNSC RD-337 ............................................................................................................. 16

1.4 Document Structure .................................................................................................... 17

2.0 GENERAL SITE DESCRIPTION AND CHARACTERISTICS...................................... 19

2.1 Site location and description ........................................................................................ 19

2.2 Programme for meteorological investigation .................................................................. 24

2.2.1 Meteorological and atmospheric monitoring................................................................... 24

2.2.2 Regional Climatology ................................................................................................... 26

2.2.3 Temperature............................................................................................................... 26

2.2.4 Precipitation ............................................................................................................... 26

2.2.5 Wind.......................................................................................................................... 26

2.2.6 Lake Effect ................................................................................................................. 27

2.2.7 Atmospheric Stability................................................................................................... 27

2.2.8 Atmospheric Pressure.................................................................................................. 36

2.3 Monitoring Programme for Surface Water and Groundwater ........................................... 37

2.3.1 Lake Water................................................................................................................. 37

2.3.2 Groundwater .............................................................................................................. 38

2.4 Population distribution................................................................................................. 41

2.5 Uses of Land and Water in the Region .......................................................................... 45

2.5.1 Agriculture ................................................................................................................. 45

2.5.2 Industry ..................................................................................................................... 45

2.5.3 Transportation............................................................................................................ 46

2.5.4 Fishing....................................................................................................................... 47

2.5.5 Biological Data............................................................................................................ 47

2.5.6 Baseline ambient radioactivity and Pre-existing Hazardous substances............................. 47

3.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR NORMAL OPERATIONS....................................................................................................... 50

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3.1 Radioactive Source Parameters for Normal Discharges to Air .......................................... 50

3.2 Radioactive Source Parameters for Normal Discharges to Surface Water and Groundwater 52

3.3 Identification of Potential Critical Groups and Their Characterization................................ 56

3.3.1 Selection of Potential Critical Groups............................................................................. 56

3.3.2 Characterization of Potential Critical Groups................................................................... 58

3.3.3 Determination of average dilution factor in fish home range ........................................... 59

3.3.4 Dairy cow’s diet .......................................................................................................... 60

3.4 Radioactive Effluent Dose Consequence for Normal Operations....................................... 60

3.4.1 Environmental Pathway Model...................................................................................... 60

3.4.2 Pathway Analysis Model and its Implementation ............................................................ 66

3.4.3 Assessment Results..................................................................................................... 66

3.4.4 Sensitivity Analysis ...................................................................................................... 77

3.5 Radioactive Effluent Dose Consequence for Refurbishment and Decommissioning ............ 79

3.5.1 Radiological Implications from Decommissioning and Refurbishment ............................... 80

3.5.2 Source Term and Dose Assessment for Refurbishment and Decommissioning Operations.. 82

3.6 Assessment of Consequences from Radioactive Waste and Used Fuel Management .......... 85

4.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR ACCIDENTAL DISCHARGES....................................................................................................... 86

4.1 Derivation of Accident Scenario and Release Characteristics............................................ 86

4.2 Representative Source Term for Radioactive Airborne Releases....................................... 87

4.2.1 Core Inventory ........................................................................................................... 88

4.2.2 Derivation of Release Fractions for RSGB Releases......................................................... 90

4.3 Source Term for Accidental Radioactive Discharges to Surface Water and Groundwater .... 92

4.4 Offsite Public Dose Consequence for Radioactive Airborne Release Accidents ................... 93

4.4.1 Dose Targets and Limits .............................................................................................. 93

4.4.2 Dose Consequence Calculations.................................................................................... 94

4.4.3 RSGB Dose Consequence Results ................................................................................. 96

4.5 Offsite Public Dose Consequence for Radioactive Waterborne Release Accidents .............106

4.6 Impact of Mitigation on Dose Consequence for Accidents ..............................................106

5.0 CONSIDERATION OF THE FEASIBILITY OF AN EMERGENCY PLAN.................... 107

5.1 Objectives for Emergency Planning..............................................................................107

5.2 Existing Emergency Plans ...........................................................................................108

5.3 Protective Action Levels..............................................................................................111

5.4 Comparison with IAEA Requirements and Recommendations from Other Agencies ..........113

5.5 Potential Impact on Emergency Planning of OPG’s New Nuclear at Darlington .................114

6.0 QUALITY ASSURANCE PROGRAMME ................................................................. 115

7.0 CONCLUSIONS .................................................................................................. 116

8.0 REFERENCES ..................................................................................................... 118

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APPENDIX

Appendix A: Abbreviations………………………………………….………………………………………………..123

Appendix B: Extracts from IAEA GS-R-2……………………………………….………………………………..125

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Figures

Page

Figure 2.1-1: Darlington Location. [OPG 2004] ...................................................................20

Figure 2.1-2: Darlington Site with the Surrounding Regions. [OPG 2004] ..............................21

Figure 2.1-3: Site Access Locations [OPG 2004]...................................................................22

Figure 2.3-1: Map of Darlington Monitoring Wells [CH2MHill 2009a] ......................................40

Figure 2.4-1: Population Distribution Grid for the Area Surrounding the Proposed Darlington Nuclear Site......................................................................................................................42

Figure 2.4-2: Land Use Prediction to 2031[Regional Municipality of Durham 2008] .................43

Figure 2.5-1: Darlington Environmental Monitoring Locations [OPG 2008c] ............................49

Figure 3.3.1-1: Locations of potential critical groups ............................................................57

Figure 3.4.1-1: Environmental transfer model [CSA 2008] ....................................................61

Figure 4.4-1: Variation of Committed Effective Dose with Distance for RSGB SRF Release – EARLY 7 Day...................................................................................................................101

Figure 4.4-2: Variation of Dose to Thyroid with Distance for the RSBG SRF Release – EARLY 7 Day................................................................................................................................103

Figure 4.4-3: Variation of Committed Effective Dose with Distance for RSGB LRF Release - LATE.....................................................................................................................................104

Figure 5.2-1: Darlington Specific Response Sectors in Primary Zone ....................................109

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Tables

Page

Table 2.2-1: Data collected at the Darlington Site from 1996-2000 [AMEC NSS 2008a] ...........28

Table 2.2-2: Data collected at the Bowmanville Station from 1971-2000 [OPG 2004] ..............29

Table 2.2-3: Precipitation data from nearby sites 1971-2000 [OPG 2004]...............................29

Table 2.2-4: Triple Joint Wind Frequencies – Stability Class A 1996-2000 [AMEC NSS 2008a] .30

Table 2.2-5: Triple Joint Wind Frequencies – Stability Class B 1996-2000 [AMEC NSS 2008a] .31

Table 2.2-6: Triple Joint Wind Frequencies – Stability Class C 1996-2000 [AMEC NSS 2008a] .32

Table 2.2-7: Triple Joint Wind Frequencies – Stability Class D 1996-2000 [AMEC NSS 2008a].33

Table 2.2-8: Triple Joint Wind Frequencies – Stability Class E 1996-2000 [AMEC NSS 2008a] .34

Table 2.2-9: Triple Joint Wind Frequencies – Stability Class F 1996-2000 [AMEC NSS 2008a] .35

Table 2.2-10: Wind Speed Frequencies – Grand Total 1996-2000 [AMEC NSS 2008a] ............35

Table 3.1-1: Estimated maximum Airborne Emission to the Environment [OPG 2009d]............51

Table 3.1-2: Airborne Release Parameters [OPG 2009d] .......................................................52

Table 3.2-1: Estimated maximum waterborne emission to the environment [OPG 2009d]........53

Table 3.2-2: Discharge Rate of Liquid Effluents [OPG 2009d] ................................................54

Table 3.2-3: Aquatic Plume Parameters [OPG 2008c]............................................................55

Table 3.3.1-1: Summary of Potential Critical Groups [OPG 2008c] .........................................56

Table 3.4.1-1: Transfer Compartments and their Units [CSA 2008]........................................63

Table 3.4.1-2: Transfer Parameters and their Units [CSA 2008] ............................................64

Table 3.4.3-1: Radionuclides of Importance Released from ACR Reactor [OPG 2008c]* ...........67

Table 3.4.3-2: Limiting radionuclides in particulate released from ACR-1000 reactor................68

Table 3.4.3-3: Limiting Radionuclide in Beta-Gamma Released from ACR-1000 Reactor via Water pathway...........................................................................................................................69

Table 3.4.3-4: Doses to Potential Critical Groups Resulting from the Operation of AP1000 Reactor (Once-Through Cooling) ........................................................................................71

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Table 3.4.3-5: Doses to Potential Critical Groups Resulting from the Operation of EPR Reactor (Once-Through Cooling) ....................................................................................................72

Table 3.4.3-6: Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 Reactor (Once-Through Cooling) ........................................................................................73

Table 3.4.3-7: Doses to Potential Critical Groups Resulting from the Operation of AP1000 reactor (Cooling Tower Option)......................................................................................................74

Table 3.4.3-8: Doses to Potential Critical Groups Resulting from the Operation of EPR reactor (Cooling Tower Option)......................................................................................................75

Table 3.4.3-9: Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 reactor (Cooling Tower Option) ..........................................................................................76

Table 3.4.4-1: Effect of temperature change on public dose..................................................78

Table 3.4.4-2: Effect of flow rate change on public dose .......................................................78

Table 3.4.4-3: Doses to hypothetical groups ........................................................................79

Table 3.5-1: Comparative Analysis of Radionuclide Inventory During Refurbishment and Decommissioning ..............................................................................................................82

Table 3.5-2: Pickering A Releases to Air During Refurbishment of Units 3 and 4 (Bq) ..............84

Table 3.5-3: Pickering A Releases to Water During Refurbishment of Units 3 and 4 (Bq)..........84

Table 4.2-1 : EPR Core Inventory From Vendor Data (AREVA 2007) ......................................88

Table 4.2-2 : Baseline Release ...........................................................................................91

Table 4.2-3 : RSGB Releases Normalised against SRF and LRF..............................................92

Table 4.4-1: Protective Action Levels...................................................................................93

Table 4.4-2: Health Canada Recommended Intervention Levels ............................................94

Table 4.4-3: Variation of Committed Effective Early Whole Body Doses with Distance for RSGB SRF..................................................................................................................................98

Table 4.4-4: Variation of Early Equivalent Dose to the Thyroid with distance for RSGB SRF Release ............................................................................................................................99

Table 4.4-5: Variation of Late Committed effective Whole Body Doses with Distance for RSGB LRF Release....................................................................................................................100

Table 4.4-6: Dose (mSv) by Pathway at 1 km from Release Point for RSGB LRF ...................102

Table 4.4-7: Total Event Committed Effective Doses for RSGB LRF......................................105

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Table 5.3-1 Protective Action Levels (PALs) .......................................................................112

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EXECUTIVE SUMMARY

OPG has commenced the federal approvals process for site preparation, construction, and operation of up to four additional nuclear reactors for the purpose of generating up to 4,800 MW of electricity at the Darlington Nuclear Site (“the site”). An application for a License to Prepare the Site was submitted to the Canadian Nuclear Safety Commission (CNSC) in September, 2006. The three reactors that are currently under consideration are:

Advanced CANDU Reactor (ACR-1000)

The Advanced CANDU Reactor (ACR-1000) is a Pressurized Hybrid Reactor (PHR) by Atomic Energy of Canada Limited. The ACR-1000 is based on the Canadian Deuterium Uranium (CANDU) technology. It uses light water (H2O) to cool the fuel and heavy water (D2O) for the moderator. The ACR-1000 plant (planned) has a net electrical output of approximately 1085 MW. It is a Generation III+ reactor.

Areva US EPR (EPR)

Areva's US EPR Reactor (EPR) is an evolutionary Pressurized Water Reactor (PWR) by AREVA NP. It has a rated electrical power of 1,580 MW and a design life of 60 years. The reactor operates with enriched uranium fuel (enrichment of up to 5%). The EPR is classified as a Generation III+ reactor.

Advanced Passive (AP1000) Reactor

The Advanced Passive (AP1000) reactor is another PWR by Westinghouse. Its safety systems operate passively, using natural forces such as gravity and natural circulation in order to function. The AP1000 output is 1037 MWe (net). It operates with enriched uranium fuel (enrichment of up to approximately 4.5%). It is a Generation III+ reactor.

AMEC NSS has been contracted by Ontario Power Generation (OPG) to evaluate the Darlington Site for suitability for the new nuclear power plant. This Task 2 Dispersion of Radioactive Material in Air and Water report is produced as part of the Site Evaluation of the OPG New Nuclear at Darlington to satisfy the requirements of Canadian Nuclear Safety Commission (CNSC) Regulatory Document 346 (RD-346) Site Evaluation for New Nuclear Power Plants. RD-346 is consistent with, and was developed based largely on, the following six IAEA Safety Guidelines for which aligned work sub-packages are being completed:

• NS-G-3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants [IAEA 2002a]

• NS-G-3.2, Dispersion of Radioactive Material in Air and Water and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants [IAEA 2002b]

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• NS-G-3.3, Evaluation of Seismic Hazards for Nuclear Power Plants [IAEA 2002c]

• NS-G-3.4, Meteorological Events in Site Evaluation for Nuclear Power Plants [IAEA 2002d]

• NS-G-3.5, Flood Hazard for Nuclear Power Plant on Coastal and River Sites [IAEA 2002e]

• NS-G-3.6: Geotechnical Aspects of Site Evaluation and Foundations for Nuclear Power Plants [IAEA 2002f]

This report addresses the requirements that are specified by the IAEA Safety Guide NS-G-3.2 [IAEA 2002b] including:

� Assessment of risks to the public for routine releases of radioactivity.

� Accident assessments for bounding release scenarios.

� Assessment of adequacy of Emergency Response plans to meet new reactor risk.

It also addresses the requirements specified in CNSC regulatory document RD-346 [CNSC 2008a] in relation to dispersion of radioactive materials, and IAEA NS-R-3, Site Evaluations for Nuclear Installations [IAEA 2003].

All work to evaluate the site’s compliance with the above site evaluation requirements was carried out in accordance with the AMEC NSS Quality Assurance system, which is fully compliant with the ISO 9001 standard. It is also compliant with the requirements of CSA Z299.1-85 and applicable portions of CSA N286.2-00 and N286.7-99 standards.

The study presents five key findings confirming that it is feasible to construct a new nuclear power plant at the site to satisfy the above requirements:

1. Monitoring data required to assess radiological impact from radioactive releases into the environment and the baseline information on environmental contamination, hydrogeology, meteorology and population distribution and habits have been collated. The site provides comprehensive datasets surpassing the requirements of IAEA and CNSC expectations with data collated over several decades to monitor performance of the operational nuclear generating station. These have been recently complemented with additional information as part of the Environmental Assessment which is being conducted by Ontario Power Generation.

2. Representative releases of radioactive materials have been estimated for both normal operations and accidents. These estimates were made on the basis of information provided by reactor vendors and, in the case of abnormal releases, on the basis of the Canadian regulatory requirements RD-337 [CNSC 2008b] specifying frequency and credible threshold releases. It is assumed that any

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reactor design licensed for construction in Canada will satisfy these requirements.

3. An assessment of atmospheric, surface water and groundwater dispersion of radioactive materials in the environment was conducted for both normal operations and accidental releases. Doses to the most exposed members of the public have been estimated.

4. Estimated maximum annual doses for normal operations are 5 µSv for both once-through cooling option and cooling tower option, which represent a small fraction of the dose limit of 1 mSv. This demonstrates that both options are feasible and can satisfy Canadian and IAEA requirements. Additional sensitivity studies were conducted to gauge the impact on dose if in the future population groups were to be located in locations with maximum exposure. Even such a conservative scenario resulted in estimated doses significantly below the regulatory limit.

5. Stylized accident release scenarios based on the RD-337 Small Release and Large Release Frequency safety goals were modeled to predict possible doses to the public. These results are consistent with predicted conditions for Emergency Planning. The overall impact from the New Nuclear at Darlington on emergency planning from an accident event has been considered in [OPG 2009c].

The results indicate that sheltering for an event such as that modeled could be mandatory to around 2 km from the point of release, while the option of sheltering may be considered for out to 10 km. Evacuation however would only require consideration for less than 2 km from the point of release. The mandatory evacuation dose is only exceeded within 500 m from the point of release.

The results indicate that if early phase evacuation takes place, approximately half of the total predicted dose for the release over 1 year is avertable. The relocation limit for dose over 1 year is not exceeded beyond 1 km from the point of release, and as such, no local residents would require permanent relocation. It is possible that temporary relocation measures will be required within this 1 km area for a time immediately following the release and for as long as 1 year. This relocation would only apply to permanent local residents and not to workers or businesses.

Based on the RD-337 Safety Goal Based (RSGB) release analyses, the existing Darlington Nuclear Emergency Plan is adequate for the OPG New Nuclear at Darlington as well.

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1.0 INTRODUCTION

1.1 Project and Task Background

OPG has commenced the federal approvals process for site preparation, construction, and operation of up to four additional nuclear reactors for the purpose of generating up to 4,800 MW of electricity at the Darlington Nuclear Site (“the site”). An application for a License to Prepare the Site was submitted to the Canadian Nuclear Safety Commission (CNSC) in September, 2006. The three reactors that are currently under consideration are:

Advanced CANDU Reactor (ACR-1000)

The Advanced CANDU Reactor (ACR-1000) is a Pressurized Hybrid Reactor (PHR) by Atomic Energy of Canada Limited. The ACR-1000 is based on the Canadian Deuterium Uranium (CANDU) technology. It uses light water (H2O) to cool the fuel and heavy water (D2O) for the moderator. The ACR-1000 plant (planned) has a net electrical output of approximately 1085 MW. It is a Generation III+ reactor.

Areva US EPR (EPR)

Areva's US EPR Reactor (EPR) is an evolutionary Pressurized Water Reactor (PWR) by AREVA NP. It has a rated electrical power of 1,580 MW and a design life of 60 years. The reactor operates with enriched uranium fuel (enrichment of up to 5%). The EPR is classified as a Generation III+ reactor.

Advanced Passive (AP1000) Reactor

The Advanced Passive (AP1000) reactor is another PWR by Westinghouse. Its safety systems operate passively, using natural forces such as gravity and natural circulation in order to function. The AP1000 output is 1037 MWe (net). It operates with enriched uranium fuel (enrichment of up to approximately 4.5%). It is a Generation III+ reactor.

AMEC NSS has been contracted by Ontario Power Generation to evaluate the Darlington Site for suitability for the new nuclear power plant. This Task 2 Dispersion of Radioactive Material in Air and Water report is produced as part of the Site Evaluation of the OPG New Nuclear at Darlington to satisfy the requirements of Canadian Nuclear Safety Commission (CNSC) Regulatory Document 346 (RD-346) Site Evaluation for New Nuclear Power Plants. RD-346 is consistent with, and was developed based largely on, the following six IAEA Safety Guidelines for which aligned work sub-packages are being completed:

• NS-G-3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants [IAEA 2002a]

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• NS-G-3.2, Dispersion of Radioactive Material in Air and Water and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants [IAEA 2002b]

• NS-G-3.3, Evaluation of Seismic Hazards for Nuclear Power Plants [IAEA 2002c]

• NS-G-3.4, Meteorological Events in Site Evaluation for Nuclear Power Plants [IAEA 2002d]

• NS-G-3.5, Flood Hazard for Nuclear Power Plant on Coastal and River Sites [IAEA 2002e]

• NS-G-3.6: Geotechnical Aspects of Site Evaluation and Foundations for Nuclear Power Plants [IAEA 2002f]

This report addresses the requirements that are specified by CNSC guidance document RD-346 [CNSC 2008a] in relation to dispersion of radioactive materials, and by the IAEA Safety Guide NS-G-3.2 [IAEA 2002b] including:

� Assessment of risks to the public for routine releases of radioactivity.

� Accident assessments for bounding release scenarios.

� Assessment of adequacy of Emergency Response plans to meet new reactor risk.

It also addresses the requirements specified in NS-R-3; Site Evaluations for Nuclear Installations [IAEA 2003] in relation to dispersion of radioactive materials.

1.2 Objectives

The objectives are as follows:

1.2.1 Normal Operations

Objective 1. The evaluation of the transport and diffusion of effluents discharged into the atmosphere for radioactive materials to assess the suitability of the Darlington Nuclear site.

Objective 2. The evaluation of the transport and diffusion of effluents discharged into the hydrosphere for radioactive materials to assess the suitability of the Darlington Nuclear site.

Objective 3. The uses of land and water in the region of the site should be investigated as part of the environmental assessment and to demonstrate feasibility of the emergency planning.

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1.2.2 Accidents

Objective 4. The study of the population distribution to evaluate the potential radiological impacts from accidents and to assist in the demonstration of the feasibility of emergency measures.

Objective 5. Demonstration of the feasibility of an emergency plan.

1.3 Regulatory Framework Overview

1.3.1 CNSC RD-346

All requirements specified in the CNSC Guidelines [CNSC 2008a] in relation to dispersion of radioactive materials are consistent with those identified by IAEA NS-G-3.2 [IAEA 2002b], with the exception of an additional statement relating to baseline data sample sizes:

“Baseline data is expected to be of sufficient sample size and duration to conduct hypothesis testing against post-commissioning (follow-up) monitoring data, with sufficient power to detect relevant effect sizes.” [CNSC 2008a]

The radiological characterization of the site and environs is very extensive due to the data collected for the on-going operation of the site [OPG 2008d] and the additional characterization done in support of the EA [OPG 2009b]. It is expected that this will provide sufficient baseline information for radiological hypothesis testing analysis going forward.

1.3.2 IAEA NS-G-3.2

IAEA Guide [IAEA 2002b] contains recommendations related to the following activities:

� Radioactive source parameters for normal and accidental discharge to air

� Programme for meteorological investigation, collection, analysis and presentation of meteorological data

� Atmospheric Dispersion Modeling

� Programme for monitoring of surface water and groundwater.

� Radioactive source parameters for normal and accidental discharge to surface water and groundwater

� Modeling of radionuclide dispersion in surface water

� Modeling of dispersion and retention of radionuclides in groundwater

� Land use, water use and population distribution in the region.

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� Consideration of the Feasibility of an Emergency Plan and Quality Assurance Programme.

1.3.3 IAEA NS-R-3

With respect to dispersion of radioactive materials this IAEA NS-R-3 Guide [IAEA 2003] does not provide any recommendations in addition to those specified in the IAEA Guide NS-G-3.2 [IAEA 2002b].

1.3.4 CNSC RD-337

Although compliance with requirements of the regulatory document Design of New Nuclear Power Plants [CNSC 2008b] is outside the scope of this assessment, it provides the design criteria for Design Basis Accident (DBA) frequencies. For the purposes of this Site Evaluation Study, it is implicitly assumed that any new nuclear plant will comply with these regulatory requirements. Section 4.2.2 of the regulatory document is reproduced below:

Safety Goals

Qualitative Safety Goals

A limit is placed on the societal risks posed by nuclear power plant operation. For this purpose, the following two qualitative safety goals have been established:

Individual members of the public are provided a level of protection from the consequences of nuclear power plant operation such that there is no significant additional risk to the life and health of individuals; and

Societal risks to life and health from nuclear power plant operation are comparable to or less than the risks of generating electricity by viable competing technologies, and should not significantly add to other societal risks.

Quantitative Application of the Safety Goals

For practical application, quantitative safety goals are established to achieve the intent of the qualitative safety goals. The three quantitative safety goals are:

1. Core damage frequency;

2. Small release frequency; and

3. Large release frequency.

A core damage accident results from a postulated initiating event (PIE) followed by failure of one or more safety system(s) or safety support system(s). Core damage frequency is a measure of the plant’s accident preventive capabilities.

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Small release frequency and large release frequency are measures of the plant’s accident mitigative capabilities. They also represent measures of risk to society and to the environment due to the operation of a nuclear power plant.

Core Damage Frequency: The sum of frequencies of all event sequences that can lead to significant core degradation is less than 10-5 per reactor year.

Small Release Frequency: The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1015 Becquerel of iodine-131 is less than 10-5 per reactor year. A greater release may require temporary evacuation of the local population.

Large Release Frequency: The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1014 Becquerel of cesium-137 is less than 10-6 per reactor year. A greater release may require long term relocation of the local population.

1.4 Document Structure

This report is organized as follows:

Section 2 - General Site Description and Characteristics:

This is the introduction to the report and contains a general description of the plant. It will mention different monitoring programs that are on the site and the population and land use within a 100 km radius.

Section 3 - Assessment of Dispersion of Radioactive Materials for Normal operation:

This section describes selection of bounding source terms for discharges to air and water under the normal operations and assessment of the resulting doses to potential critical groups. It contains the analysis of assessment results and evaluates consequences resulting from refurbishment and de-commissioning operations.

Section 4 - Assessment of Dispersion of Radioactive Materials for Accidental Discharges:

This section describes air and water source terms for beyond design basis accidents (BDBA) and design basis accidents (DBA). It will describe the dose calculation model and provide a detailed analysis of the dose consequences.

Section 5 - Consideration of the Feasibility of an Emergency Plan:

This section contains an assessment of how the Emergency Plan for Darlington is impacted by the calculated dose consequences.

Section 6 - Quality Assurance Programme:

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This section contains a description of the quality system under which the project was executed. It also contains the qualification status of every data stream that was used in this project.

Section 7 - Conclusion

This section provides a summary of the findings.

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2.0 GENERAL SITE DESCRIPTION AND CHARACTERISTICS

2.1 Site location and description

The Darlington Nuclear Site, where it is proposed the OPG New Nuclear at Darlington (NND) would be built, is located on the north shore of Lake Ontario, about 65 km east of the City of Toronto, in the Municipality of Clarington, Region of Durham in Ontario, Canada.

The site is at latitude 43º 53' north and longitude 78º 43' west. The entire site (including Darlington Nuclear Generating Station - DNGS) comprises a land parcel of about 485 hectares. Maps showing the site with the neighbouring area are provided as Figures 2.1-1 and 2.1-2.

The site is bounded by the South service road to the north and Lake Ontario to the south. To the west, the site is bounded by Solina Road. The St. Marys Cement plant occupies the land east of the site.

The site is situated in an undulating to moderately rolling limestone till plain. The previously irregular terrain has been graded in the existing powerhouse area to an elevation of about 100 m. The surface elevation rises towards the north with a mean elevation of 122 m just south of the railway tracks. To the north of the tracks, the ground is irregular ranging from 120 to 128 m elevation.

A higher ridge, starting from the shore just east of Raby Head, extends diagonally across the site in a north-westerly direction with levels ranging up to 15 m above the surrounding ground. On the west half of the site, the land slopes from 100 m elevation at the Darlington powerhouse to 130 m at the east end of the site.

Offshore from the site, the lake bottom slopes away gradually reaching a depth of 6 m at about 425 m from shore and 14 m at 1.2 km from shore. An undulating terrain usually increases atmospheric turbulence near ground level during times of moderate or strong winds, resulting in lower pollution concentrations at locations near the station.

The site is well-supplied by access roads for supply of off-site emergency aid and for ease of evacuation of non-essential personnel in case of an emergency. The site may be easily reached by car or rail. The multi-lane Macdonald-Cartier Freeway (The Highway of Heroes section of the 401) runs east/west, immediately north of the site. There are three controlled exits to the Macdonald-Cartier Freeway (Highway of Heroes), directly through Holt Road and two others less than 3 km on either side of the NGS.

Rail access can also be provided by the Canadian National Railway's main line, which bisects the site in an approximate east to west direction. The tracks are fenced and gated across the entire site. A rail siding area has been provided at the east boundary limit of the OPG Property.

Figure 2.1-3 shows the major access and physical features of the site.

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Figure 2.1-1: Darlington Location. [OPG 2004]

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Figure 2.1-2: Darlington Site with the Surrounding Regions. [OPG 2004]

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Figure 2.1-3: Site Access Locations [OPG 2004]

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2.2 Programme for meteorological investigation

2.2.1 Meteorological and atmospheric monitoring

OPG has been gathering on-site weather data at the site since 1991 [OPG 2004].

The meteorological tower at the site is located just north of the facility, just southeast of the intersection of Highway 401 and Holt Road (main access to the site). The tower has no significant obstructions from nearby buildings. Meteorological data available from the site consist of wind speeds at two heights and temperature at one height. Data are stored in a data logger and transmitted to a personal computer every fifteen minutes where the data are stored. A parallel fifteen minute back up of the same data is recorded by the Plant Information (PI) interface server, located at the Pickering Training Centre.

There are two anemometers on the monitoring tower, one at the standard ten-metre height, the other at fifty metres. The combined wind speed and direction sensors are RM Young Wind Monitors, model 05305, at both the ten and fifty metre heights. Each set of wind speeds is sampled at a three second interval. The data logger records sixty- and fifteen-minute averaged wind speeds, in km/h, and directions, in degrees, from both anemometers.

A Campbell Scientific Model 107 temperature probe measures the air temperature at a height of ten metres (in compliance with IAEA guidance). Although the guidance requires temperature measurement to be taken at the same height as the wind speed and recommend heights of 10 m and 50 m above ground, it should be noted that Meteorological Services Canada (MSC) measures temperature at a height of 2 m. For this reason, the MSC data sets are used here for discussion purposes only. Where data sets are required for calculation purposes, the on-site meteorological data measurement have been used.

The temperature is sampled at a three-second interval and the data logger records sixty-minute and fifteen-minute averaged temperatures. The fifteen-minute averaged data sets are used in dispersion models for short duration release consequence prediction and the sixty-minute averaged data sets are used for prolonged and long term release modelling.

Between 2001 and 2004, the observed wind direction data from the site meteorological tower appears to show a slight shift in direction1 from north to west in comparison with past historical and MSC data for the same period. A performance assessment of the on-site meteorological tower has been carried out and the root cause rectified; the subsequent data collection, post 2005 has not shown such deviations. For this reason, the weather data sets for the five-year period from 1996 to 2000 have been used for this assessment as the most recent contiguous five-year

1 Sensitivity analysis was carried out on the 2001-2005 data and the difference in accident impact results due to use of 1996-2000 weather data were found to be approximately 3%.

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period of data. Recent annual data observations show that this basis is supportable and that future five-year data sets will be consistent with this.

Other than the monitoring station in Darlington, meteorological data were collected from Environment Canada maintained stations [MSC 2008] that were present within 100 km from Darlington as follows:

Daily Data Belleville Bloomfield Cobourg STP Oshawa WPCP Thornhill Grandview Toronto Island Toronto Lester B. Pearson International Airport

Hourly Data Peterborough Airport Central Toronto (Precipitation post 2000) Toronto Island Toronto Lester B. Pearson International Airport Trenton Airport (Precipitation Post 2000) Trenton Municipal Airport Bowmanville Mozert (Precipitation 1971-2000 only)

Meteorological data has been collected from these nearby MSC stations over the period for which Darlington Meteorological Station data sets are in question. Given that the meteorological data averages for Trenton, Toronto Island, Toronto (Pearson) and Peterborough stations, in particular, are similar to those of the data averages for the Darlington meteorological tower, it is considered that these four locations form a representative set for current meteorological data at the site.

The proposed location of the land fill from the new building locations is to the east of the existing tower location and is built up to a maximum height of 140 m. The guardhouse to the south will be relocated during construction and it may be necessary to consider the relocation of the meteorological tower at the same time, purely for logistical purposes. At that time, consideration should be given to ensuring that the tower is not relocated to a position where building wake effects will influence the meteorological data obtained.

Objective S2.12 of the [IAEA 2002b] states that for the programme for meteorological investigation:

“The meteorological data collected should be compatible in terms of their nature, scope and precision with the methods and models in which they will be used in

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evaluating the radiation exposure of the public and the radiological impact on the environment for assessment against each regulatory objective.”

Humidity, air pressure and precipitation are currently not logged on-site by the meteorological tower; however the information is readily available from Environment Canada stations as listed above.

2.2.2 Regional Climatology

The meteorology in the vicinity of site is affected by meso-scale factors consisting of the general circulation of air masses and the effects of the Great Lakes, and micro-scale factors that include off-shore winds (for coastal areas due to diurnal temperature changes), terrain and topography.

Meso-scale factors affect meteorology beyond 10 km from the point of interest. Micro-scale factors affect weather within 10 km from the point of interest.

2.2.3 Temperature

The area has warm summers and cold winters. The mean annual temperature at the site is 7.9°C. Mean daily temperatures fall below zero in the winter months (December through March). The coldest recorded hourly temperature measurement was -30.5°C, whereas the warmest recorded hourly temperature was 31.5°C. The coldest month was January, with mean daily temperatures in the vicinity of -5.5°C. Summer temperatures averaged 17.7°C, or higher. The highest daily mean temperature recorded was 20.0°C, which occurred in July. Site temperature data from the period of 1996-2000 are summarized in Table 2.2-1. The data sets are consistent with temperatures recorded at the MSC stations in the surrounding regions for the same and subsequent periods.

2.2.4 Precipitation

Precipitation is quite uniformly distributed throughout the year (see Table 2.2-2) with a slight maximum in the fall season. Total annual precipitation averages about 850 mm, of which about one-tenth occurs as snowfall [OPG 2004]. The greatest 24 hour rainfall recorded in August during the period of 1971-2000 is 81.2 mm and the greatest 24 hour snowfall recorded in January is 29 mm (as water). These results are based on data recordings taken at Bowmanville Mozert station, the nearest measuring point to the site between 1971 and 2000. Precipitation data sets post 2000 are available from Trenton and Toronto (Pearson). Results from 2000 to 2006 are presented in [AMEC NSS 2009a].

Also a comparison of the precipitation values recorded in the nearby stations has been provided in Table 2.2-3 for a quick comparison.

2.2.5 Wind

Table 2.2-4 to Table 2.2-9 provide the joint frequencies of wind speed and wind direction from hourly data collected from the on-site 50 m meteorological tower for

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each stability class (See Section 2.2.7). The summary of the wind speed frequencies is provided in Table 2.2-10. The data were measured at the 10 m height. The wind directions indicated are the directions from which the wind is blowing.

The prevailing winds at the site blow from the west, northwest or southwest over 45% of the time. In addition, there is a high frequency (55%) of light winds (<3 m/s or 10.5 km/h). Winds are lightest in the summer and strongest in the winter and spring.

2.2.6 Lake Effect

The proximity of the station to the lake affects the meteorology near the plant due to the “lake effect”. Lake effect results from temperature differences between land and water. A Thermal Internal Boundary Layer (TIBL) phenomenon develops as a result of this.

In the spring and summer, when the skies are clear and the geostrophic winds are light, a strong temperature gradient between the air over the land and the air over Lake Ontario begins to form in the morning. This is because the air over the ground is heated more rapidly to a higher temperature than the air over the water. The cooler air over the water is denser and flows inland at low levels, producing a lake breeze. At night, the situation is reversed. The land and the air above it cool faster than the water and the air above. The denser air is then over land and the low-level air flows from the land to the lake, creating a land breeze. The lake breeze usually has a much greater intensity than the land breeze. In the fall and winter, the lake is generally warmer than land resulting in more frequent land breezes.

In warm seasons, due to solar heating, the air over land is often warmer than that over water. When cold stable lake air flows over warmer land, the resulting upward heat flux gives rise to a TIBL. This TIBL grows in depth with distance inland as the stable air is advected over land and adjusts to changes in surface roughness, heat, and moisture input. The depth of the TIBL is typically hundreds of metres and extends about 10 km inland before a new equilibrium is reached.

For emission sources near the ground, pollutants emitted into the unstable boundary layer would also result in higher than expected ground level concentrations during on-shore flows because the stable layer aloft would limit vertical diffusion (compared to the situation when there is no TIBL).

2.2.7 Atmospheric Stability

The most commonly used method of categorizing the amount of atmospheric turbulence present is categorization of turbulence into six stability classes. The turbulent nature of the atmosphere strongly affects the concentration of contaminants downwind of the release point. A highly turbulent atmosphere is referred to as Stability Class A and occurs under warm sunny conditions. A highly stable atmosphere is referred to as Stability Class F and occurs typically under night-time conditions and low wind speeds. A neutral atmosphere, referred to as Stability Class D, is representative of average turbulence conditions and occurs typically under cloudy,

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windy conditions. All other things being equal, downwind contaminant concentrations for cold ground-level releases are highest when the atmosphere is highly stable (F-stability) and lowest when the atmosphere is highly unstable (A-stability).

Various schemes have been developed for predicting stability class. A widely accepted method and one used by OPG is the Sigma Theta (σθ) method. It is based primarily on the standard deviation of continuous measurements of wind direction, but also on the time of day and wind speed [Ontario Hydro, 1984]. The meteorological data used to predict stability class are measured at the 10 m and 50 m elevations of the on-site tower.

The meteorological assessment that qualifies the site as acceptable to build a new nuclear power plant is contained in Site Evaluation Report on Meteorological Events [AMEC NSS 2009a].

Table 2.2-1: Data collected at the Darlington Site from 1996-2000 [AMEC NSS 2008a]

Daily Mean

Mean Daily Max

Mean Daily Min

Months (1996-2000)

(°C) (°C) (°C)

January -5.5 0.6 -10.8

February -2.8 2.1 -9.2

March 0 6.6 -6

April 5.9 11.7 0.6

May 12.3 19.8 5.7

June 17.7 22.3 13.1

July 20 25.7 15.3

August 19.7 24.2 15

September 16.4 20.9 11.1

October 9.7 13.7 5.5

November 3.4 7.7 -1.9

December -1.5 4.3 -10.8

Year 7.9 25.7 -10.8

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Table 2.2-2: Data collected at the Bowmanville Station from 1971-2000 [OPG 2004]

Monthly Average Daily Extremes

Month Precipitation Rain Snow Precipitation Rain Snow

(mm) (mm) (mm) (mm) (mm) (mm)

Jan 63.1 33.1 30 46.2 46.2 29

Feb 47.2 30.8 16.4 42.2 42.2 19.4

Mar 60.7 47.2 13.5 47.6 47.6 20.8

Apr 72.9 70 2.9 43.2 43.2 10.2

May 73.7 73.7 0 36.2 36.2 0

Jun 81.5 81.5 0 50.6 50.6 0

Jul 63.7 63.7 0 51.1 51.1 0

Aug 81 81 0 81.2 81.2 0

Sep 90.5 90.5 0 84 84 0

Oct 67.9 67.8 0.1 48.6 48.6 12.2

Nov 84 77.9 6.1 71.4 71.4 15.5

Dec 71.6 47.4 24.2 41.1 41.1 24

Annual 857.9 764.6 93.2 - - -

Table 2.2-3: Precipitation data from nearby sites 1971-2000 [OPG 2004]

Precipitation (mm)

Month Toronto Oshawa Port Hope

January 52.2 71 59

February 42.6 52.7 55.8

March 57.1 62.3 76.6

April 68.4 73.1 69.5

May 72.5 74.7 65

June 74.2 80.6 72.1

July 74.4 67.3 53.3

August 79.6 83.3 75.9

September 77.5 87.9 80.7

October 64.1 66.3 68.7

November 69.3 79.9 75

December 60.9 78.7 80.5

Year 792.7 877.9 832

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Table 2.2-4: Triple Joint Wind Frequencies – Stability Class A 1996-2000 [AMEC NSS 2008a]

Wind

Direction Wind Speed, u (m/s) A

(wind

blowing

from)

u ≤ 2 2 6 Total

Frequency (%) at 10 m Height

N 0.08 0.01 0 0 0 0 0.09

NNE 0.09 0.01 0 0 0 0 0.1

NE 0.06 0.01 0 0 0 0 0.07

ENE 0.07 0.01 0 0 0 0 0.08

E 0.2 0.05 0.01 0 0 0 0.26

ESE 0.22 0.14 0.02 0 0 0 0.38

SE 0.25 0.09 0 0 0 0 0.34

SSE 0.19 0.04 0.01 0 0 0 0.24

S 0.2 0.09 0.01 0 0 0 0.3

SSW 0.3 0.31 0.08 0.01 0 0 0.7

SW 0.33 0.32 0.14 0.03 0 0 0.82

WSW 0.21 0.15 0.08 0.04 0 0 0.48

W 0.08 0.03 0.02 0.01 0 0 0.14

WNW 0.04 0.02 0.02 0.01 0.01 0 0.1

NW 0.07 0.03 0.01 0.01 0 0 0.12

NNW 0.08 0 0 0 0 0 0.08

Total 2.47 1.31 0.4 0.11 0.01 0 4.3

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Table 2.2-5: Triple Joint Wind Frequencies – Stability Class B 1996-2000 [AMEC NSS 2008a]

Wind

Direction Wind Speed, u (m/s) B

(wind blowing

from)

u ≤ 2 2 6 Total


N 0.05 0.01 0 0 0 0 0.06

NNE 0.08 0 0 0 0 0 0.08

NE 0.06 0.01 0 0 0 0 0.07

ENE 0.07 0.01 0 0 0 0 0.08

E 0.16 0.15 0.07 0.01 0 0 0.39

ESE 0.27 0.36 0.1 0.01 0 0 0.74

SE 0.2 0.31 0.05 0 0 0 0.56

SSE 0.14 0.11 0.01 0 0 0 0.26

S 0.07 0.14 0.02 0 0 0 0.23

SSW 0.02 0.15 0.21 0.03 0.01 0 0.42

SW 0.09 0.49 0.77 0.52 0.09 0.02 1.98

WSW 0.12 0.21 0.14 0.05 0 0 0.52

W 0.05 0.04 0.03 0.01 0 0 0.13

WNW 0.05 0.02 0.03 0.01 0 0 0.11

NW 0.04 0.01 0.01 0.02 0 0 0.08

NNW 0.06 0 0 0 0 0 0.06

Total 1.53 2.02 1.44 0.66 0.1 0.02 5.77

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Table 2.2-6: Triple Joint Wind Frequencies – Stability Class C 1996-2000 [AMEC NSS 2008a]

Wind

Direction Wind Speed, u (m/s) C

(wind

blowing from)

u ≤ 2 2 6 Total


N 0.11 0.06 0.02 0 0 0 0.19

NNE 0.15 0.05 0.01 0 0 0 0.21

NE 0.14 0.03 0.02 0 0 0 0.19

ENE 0.16 0.03 0 0 0 0 0.19

E 0.35 0.61 0.58 0.36 0.19 0.08 2.17

ESE 0.13 0.48 0.44 0.23 0.07 0.01 1.36

SE 0.18 0.42 0.19 0.04 0.02 0 0.85

SSE 0.17 0.37 0.1 0 0.01 0 0.65

S 0.08 0.22 0.07 0.01 0 0 0.38

SSW 0 0.03 0.13 0.06 0.03 0.03 0.28

SW 0.03 0.24 0.67 0.69 0.27 0.13 2.03

WSW 0.13 0.52 0.85 0.56 0.16 0.11 2.33

W 0.12 0.1 0.07 0.03 0.02 0 0.34

WNW 0.08 0.05 0.01 0.01 0.01 0 0.16

NW 0.09 0.08 0.05 0.02 0.01 0.01 0.26

NNW 0.1 0.05 0.04 0.02 0 0 0.21

Total 2.02 3.34 3.25 2.03 0.79 0.37 11.8

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Table 2.2-7: Triple Joint Wind Frequencies – Stability Class D 1996-2000 [AMEC NSS 2008a]

Wind

Direction Wind Speed, u (m/s) D

(wind

blowing from)

u ≤ 2 2 6 Total


N 1.4 0.46 0.41 0.34 0.23 0.22 3.06

NNE 0.89 0.37 0.26 0.13 0.05 0 1.7

NE 0.83 0.4 0.21 0.11 0.02 0 1.57

ENE 1.16 0.63 0.27 0.11 0.04 0.01 2.22

E 0.43 1.09 1.04 0.82 0.63 0.64 4.65

ESE 0.08 0.29 0.35 0.17 0.11 0.02 1.02

SE 0.08 0.13 0.13 0.07 0.03 0 0.44

SSE 0.11 0.21 0.19 0.08 0.01 0.01 0.61

S 0.08 0.23 0.21 0.14 0.06 0.02 0.74

SSW 0 0.03 0.13 0.1 0.05 0.01 0.32

SW 0.01 0.13 0.51 0.28 0.16 0.11 1.2

WSW 0.16 1 1.41 0.93 0.48 0.6 4.58

W 1.02 1.62 1.1 0.87 0.59 0.33 5.53

WNW 0.63 0.66 0.35 0.28 0.22 0.19 2.33

NW 0.83 0.53 0.43 0.42 0.39 0.34 2.94

NNW 1.24 0.42 0.47 0.51 0.41 0.28 3.33

Total 8.95 8.2 7.47 5.36 3.48 2.78 36.24

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Table 2.2-8: Triple Joint Wind Frequencies – Stability Class E 1996-2000 [AMEC NSS 2008a]

Wind Direction

Wind Speed, u (m/s) E

(wind

blowing

from)

u ≤ 2 2 6 Total


N 0.93 0.83 0.76 0.53 0.37 0.28 3.7

NNE 0.6 0.46 0.2 0.1 0.03 0.03 1.42

NE 0.56 0.41 0.36 0.24 0.11 0.08 1.76

ENE 0.97 0.94 0.78 0.64 0.41 0.32 4.06

E 0.7 0.26 0.01 0 0 0 0.97

ESE 0.45 0.2 0 0 0 0 0.65

SE 0.23 0.06 0 0 0 0 0.29

SSE 0.09 0.05 0 0 0 0 0.14

S 0.09 0.03 0.01 0 0 0 0.13

SSW 0.03 0.07 0.01 0 0 0 0.11

SW 0.06 0.18 0.02 0 0 0 0.26

WSW 0.39 0.51 0.04 0 0 0 0.94

W 0.6 0.53 0.2 0.09 0.04 0.02 1.48

WNW 0.7 1.36 1.12 0.72 0.55 0.83 5.28

NW 1.58 1.29 0.72 0.54 0.43 0.8 5.36

NNW 1.71 1.32 1.25 1.22 0.9 1.48 7.88

Total 9.69 8.5 5.48 4.08 2.84 3.84 34.43

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Table 2.2-9: Triple Joint Wind Frequencies – Stability Class F 1996-2000 [AMEC NSS 2008a]

Wind

Direction Wind Speed, u (m/s) F

(wind

blowing

from)

u ≤ 2 2 6 Total


N 0.26 0.01 0 0 0 0 0.27

NNE 0.29 0 0 0 0 0 0.29

NE 0.33 0 0 0 0 0 0.33

ENE 0.28 0.01 0 0 0 0 0.29

E 0.65 0.05 0 0 0 0 0.7

ESE 0.61 0.1 0 0 0 0 0.71

SE 0.41 0.04 0 0 0 0 0.45

SSE 0.22 0.03 0 0 0 0 0.25

S 0.22 0.04 0 0 0 0 0.26

SSW 0.26 0.09 0 0 0 0 0.35

SW 0.5 0.19 0 0 0 0 0.69

WSW 0.74 0.24 0 0 0 0 0.98

W 0.25 0.04 0 0 0 0 0.29

WNW 0.32 0.13 0.01 0.01 0 0 0.47

NW 0.45 0.12 0 0 0 0 0.57

NNW 0.27 0.17 0.08 0.02 0 0 0.54

Total 6.06 1.26 0.09 0.03 0 0 7.44

Table 2.2-10: Wind Speed Frequencies – Grand Total 1996-2000 [AMEC NSS 2008a]

Wind Speed, u (m/s)

u ≤ 2 2 6 Total


30.72 24.63 18.13 12.27 7.22 7.01 99.98

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2.2.8 Atmospheric Pressure

Hourly observations of atmospheric pressure, which were adjusted to sea level equivalent, were obtained from four MSC weather stations located near the Darlington Nuclear Generating Station. This information is presented in detail in [AMEC NSS 2009a] and indicates a maximum atmospheric pressure of 1050.3 hPa and a minimum of 961.7 hPa over the period 1977-2006.

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2.3 Monitoring Programme for Surface Water and Groundwater

2.3.1 Lake Water

Lake Currents and Waves

Water movement near the site is predominantly along the shore, occurring for 73% of the time (35% to west and 38% to east) [OPG 2002]. Onshore and offshore movement occurs about 15% of the time [OPG 2004].

The mean current speed during the spring-summer-fall periods was 1.8 cm/s. Mean speeds recorded during the winter period were higher, averaging almost 2.8 cm/s. The maximum speed was 56 cm/s, as recorded during the winter of 1982 [Ontario Hydro, 1992].

Water levels

The mean water level is at elevation 74.20 m relative to the International Great Lakes Datum [AMEC NSS 2009b]. Minimum and maximum daily means are estimated at 73.83 m and 75.73 m respectively. The site is protected from high lake levels by the protected face of the shoreline which was built to elevation 79 m, 1 m above site grade level and about 3.0 m above the highest water level ever recorded. This shoreline will provide an adequate safety barrier against the severest anticipated combination of spring runoff and wave action [OPG 2004]. Flooding analysis has indicated that for the New Nuclear at Darlington, identified potential flood hazards can all be mitigated through conventional engineering means and methods [AMEC NSS 2009b].

Ice Conditions in Lake Ontario

Ground observations of the lake since 1971 near the site revealed no extensive ice cover, only shore and slush ice, which has developed into ridges extending up to 30 m away from the shore [OPG 2004].

Water Temperature

Water temperatures have been recorded in Lake Ontario, from 0.2 to 2.2 km offshore from the site, since 1971. Water temperatures are discussed in detail in the Surface Water Technical Supporting Documentation of the Environmental Assessment for the New Nuclear at Darlington [Golder 2009b] and the Darlington Safety Report [OPG 2004].

Major up-welling and down-welling events, resulting in temperature changes of over 10°C in a period of several days, were recorded each year during the 17 years the measurements were taken. In a typical year, events of this magnitude occurred about three times during July to September [Ontario Hydro, 1991].

Further information on Lake Hydrology is provided in the Site Evaluation Report on Flood Hazards [AMEC NSS 2009b].

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The specific requirements of the IAEA Guide NS-G-3.2 have been covered in a variety of reports and are referenced below:

� The general shore and bottom configuration in the region, and unique features of the shoreline in the vicinity of the discharge. Data on bathymetry out to a distance of several kilometres, and data on the amount and character of sediments in the shallow shelf waters: Site Evaluation Report on Flood Hazards [AMEC NSS 2009b]

� Speeds, temperatures and directions of any near shore currents that could affect the dispersion of discharged radioactive material. Measurements should be made at appropriate depths and distances, depending on the bottom profile and the location of the point of discharge: [AMEC NSS 2009b]

� The duration of stagnation and characteristics of current reversals. After stagnation, a reversal in current usually leads to a large scale mass exchange between inshore and offshore waters that effectively remove pollutants from the shore zone: [AMEC NSS 2009b]

� The thermal stratification of water layers and its variation with time, including the position of the thermocline and its seasonal changes: [AMEC NSS 2009b]

� The load of suspended matter, sedimentation rates and sediment distribution coefficients, including data on sediment movements characterized by defining at least the areas of high rates of sediment accumulation: Aquatic Environment, Existing Environmental Conditions TSD – NND Environmental Assessment [Golder & SENES 2009a]

� The background levels of activity in water, sediment and aquatic food due to natural and artificial sources: Radiological Environmental Monitoring Program [OPG 2008c]

� Seasonal cycles of phytoplankton and zooplankton, with at least the periods of their presence and cyclical evolutions of their biomass: [Golder & SENES 2009a]

� Spawning periods and feeding cycles of major fish species: [Golder & SENES 2009a]

2.3.2 Groundwater

Groundwater Monitoring Program

Since 2002, OPG has undertaken a voluntary program for a reconnaissance-level groundwater impact evaluation in the area. These wells are monitored for:

• Metals (aluminium, antimony, arsenic, barium, beryllium, bismuth, boron, cadmium, calcium, caesium, chromium (Total), cobalt, copper, iron, lead, lithium, magnesium, manganese, molybdenum, nickel, phosphorus, potassium,

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selenium, silver, sodium, strontium, thallium, thorium, tin, titanium, tungsten, uranium, vanadium, zinc, zirconium)

• Anions (chloride, sulphate, bromide, fluoride, nitrate, nitrite, phosphate)

• Alkalinity*

• Organics* (base neutral and acid extractible which includes polycyclic aromatic hydrocarbons and phenolics)

• Petroleum Hydrocarbons (Fractions F1 (C6 to C10), F2 (>C10 to C16), F3 (>C16 to C34) and F4 (>C34))

• BETX (benzene, ethylbenzene, toluene, xylene)

• PCBs*

• Radionuclides (Tritium, Sr-90*, C-14*, Cs-134, Cs-137, Co-60, K-40, Th series, U series)

* Monitoring of alkalinity, organics, PCBs, Sr-90 and C-14 was discontinued in 2008 since no abnormal readings have ever been detected, nor would they be expected to occur.

In the beginning of 2007 there were 20 monitoring wells and 8 Solinst multilevel wells. For the Environmental assessment on the site, 76 new monitoring wells have been setup. One multilevel well has 2 intake zones in the bedrock and the other multilevel well has 3 groundwater intake zones in the overburden material.

Further information on the most recent groundwater monitoring programme at Darlington Nuclear Site is provided in the Geological and Hydrogeological Technical Support Document for Darlington New Build Environmental Assessment [CH2MHill 2009a] and the Site Evaluation Study Report – Geotechnical Aspects [AMEC NSS 2009c].

Locations of the Monitoring wells existing in Darlington post-commencement of Environmental Assessment study are provided in Figure 2.3-1 on the next page. Multilevel wells are marked with a MW and the standard boreholes are marked with a DN.

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Figure 2.3-1: Map of Darlington Monitoring Wells [CH2MHill 2009a]

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2.4 Population distribution

The current population distribution for the region immediately surrounding the site was generated by Statistics Canada using their Geographical Information system (GIS). It is based on the most recent census information (2006 for Canada and 2000 for the USA). The 2006 population data are summarized in [OPG 2009a]. Figure 2.4-1 shows the grid used to represent demographic distribution for a 100 km circular zone around the proposed site for OPG New Nuclear at Darlington based on the 2006 area. There are relatively few people living within 4 km of the station, give that within the immediate 8 km radius of the station the area is primarily rural with the exception of the city of Bowmanville. The population increases substantially beyond 40 km with the inclusion of the City of Toronto.

For the purposes of Emergency Response planning activities over the lifetime of the proposed OPG New Nuclear at Darlington, it is necessary to consider the projected population distributions. This can be done with some accuracy, for projections of around 20-30 years, by using municipal plans for land use, projected development and economic growth. This information is based on the municipal land use planning projects proposed in the Growing Durham report [Regional Municipality of Durham 2008]. The resulting land use prediction is shown in Figure 2.4-2.

The recommended growth scenario [Regional Municipality of Durham 2008] concludes that the majority of land areas north of the Darlington Nuclear site will be assigned for future employment (industrial) between 2031 and 2056.

The majority of residential growth to 2031 is expected to be within the current urban areas of Courtice and Bowmanville through increase in population density. The Recommended Growth Scenario has identified land north of the Canadian Pacific Railway corridor in the vicinity of the site as future residential growth areas between 2031 and 2056.

OPG has, through collaboration and formal input into municipal planning, influenced changes to the land use structure to ensure that Emergency Response capability is not diminished. As a result, the recommended growth scenario supports unimpeded emergency plan implementation through 2031 [OPG 2009c]. It is expected that the emergency preparedness consultative processes will continue and evacuation time estimates will be jointly reviewed and revised as required beyond 2031.

AMEC NSS Limited


`

`

Niagara-on-

the-LakeSt. Catherines

NiagaraFalls

Thorold

Lincoln

West Lincoln

GrimsbyStoney

Creek

Hamilton

Glanbrook

Niagara County

Erie County

Genessee County

Orleans

CountyMonroe

County

0 5 10 20

Kilometers

Sidney

Murray

HaldimandHamilton

Coburg

Colborne

Hope Port

Hope

Cramahe

Clarington

Oshawa

Whitby

Ajax

Pickering

Markham

Scarborough

EastYork

Toronto

York

Eto

bic

oke

Mississauga

Oakville

Milton

Burlington

Halton

Hills

Brampton

CaledonVaughan

King

Ric

hm

ond

Hill

Aurora

Newmarket

Whitechurch-

Stouffville

East

Gwillimbury Uxbridge

Bra

dfo

rd -

West

Guill

embur

y

New

Tecumseh

Adjala-

New

Tosorontio

Essa

Innisfil

Barrie

Springwater

Oro - Medonte

Ramara

Carden

Eldon

Brock

Mariposa

Sommerville

Fenelon

Verulam

Lindsay

Ops

Manvers

Cavan

Emily

Smith

South Monaghan

Peterborough Otanabee

Asphodel

Percy

Seymour

Rawdon

4-km8-km

16-km

24-km32-km

40-km

60-km

80-km

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N

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EN

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ES

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S

SSW

SW

WS

W

W

WN

W

NW

NNW

Lake Ontario

North York

Georgina

Harvey

Duoro

Dummer

Figure 2.4-1: Population Distribution Grid for the Area Surrounding the Proposed Darlington Nuclear Site

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Figure 2.4-2: Land Use Prediction to 2031[Regional Municipality of Durham 2008]

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2.5 Uses of Land and Water in the Region

This section addresses Objective 3.

2.5.1 Agriculture

Agricultural data presented in this report were derived primarily from the Darlington Nuclear Generating Station (NGS) Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006]. Inventories of site-specific agricultural data, which are pertinent to the food chain pathway radiological analysis, are vegetable, food crops, livestock and dairy products.

The first major category is further divided into four groups based on phenotypic and agricultural transport characteristics. These include:

� Leafy vegetables

� Exposed produce

� Protected produce

� Wheat and grains

Leafy vegetables (e.g.: cabbage and lettuce) have a broad flat leaf surface that may directly intercept deposition material from the atmosphere. In this group, the edible portion of the plant is primarily concerned with vegetative growth (leaves and stems). Exposed produce refers to vegetables and fruits that may also intercept airborne deposition material like leafy vegetables, but the edible surfaces are relatively smaller for deposition interception. The edible portions are typically concerned with reproductive functions (seeds and fruits). Protected produce items are not directly exposed to airborne material because they grow underground, or if above ground, their edible portions are protected by pods, shells, or non-edible skins or peels. The edible portions of the protected produce are typically the reproductive or storage organs. Wheat and grains are similar to protected produce but they are used both as human food and livestock feeds.

The second major category includes livestock food items of beef, pork, poultry, mutton, milk, and eggs. These are of concern because of animals grazing on contaminated vegetation or of feeding contaminated plant material to animals.

2.5.2 Industry

There are no industrial plants manufacturing hazardous materials within 8 km of the site, although there is a water treatment plant which stores quantities of chlorine. OPG will maintain the monitoring of the area development and will provide formal input to planning and construction activities to ensure sustainability in the area.

St. Marys Cement Company owns the calcining facility and quarry on the property to the east of the site. Routine blasting and the possible detonation of stored explosives

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are possible external hazards that could affect the site; however, the design will take this into account.

Industry in the Bowmanville community has primarily been light to medium manufacturing. The two largest plants produce rubber products and iron castings. There are three major pipelines near the site. Two pipelines carrying natural gas are located about 10 km from the site and one pipeline carrying refined oil products is located about 8 km from the site.

The City of Oshawa is a major centre of heavy manufacturing activity with about 90 manufacturing plants employing over 20,000 people. The largest segment of this employment is associated with the automobile industry.

The 1993 Official Plan for the Regional Municipality of Durham [Durham 1993] designates approximately 3100 hectares as industrial lands in the region.

For more information, refer to the Darlington Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006].

It is proposed that a Durham York waste park will be constructed within Clarington Energy Park to the west of the site on a site of 12 hectares [MOE 2006]. The proposal includes facilities for the management of municipal solid waste, including incinerator and ash-processing facilities. Once it is fully operational, it will have the capacity to process 400,000 tonnes/year of waste.

2.5.3 Transportation

There are 4 forms of transportation available near the site:

Rail – Canadian National Railway’s Toronto to Montreal main line passes through the exclusion area around the Darlington station. An average of 53 trains per day use this track and, for freight trains, each has an average of 72 cars. The rail line is dual track and the type and condition of track are excellent. The track is curved and has a gentle slope. There are no level crossings on the site. Details of rail related hazard analysis are provided in Site Evaluation Report “Summary Report: Site Evaluation of External Human Induced Events” [AMEC NSS 2009d]. There are no significant hazards from rail incidents that can not be mitigated.

Airport – There are two airports within a distance of 40 km (Greenbank and Oshawa). Details of hazard analysis from airport related incidents are provided in Site Evaluation Report “Summary Report: Site Evaluation of External Human Induced Events” [AMEC NSS 2009d] There are no significant hazards from aircraft incidents that can not be mitigated.

Water - A large number of ships, from small pleasure craft to large lake and ocean vessels ply Lake Ontario. Winter conditions limit this traffic to about 8 months of the year. The larger cargo vessels move along shipping lanes, which are designated by the Ministry of Transport, and the nearest approach is about 27 km from the site. The ports of Oshawa, Whitby, and Cobourg are nearest to the site and receive small lake

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vessels, gasoline being the most dangerous cargo carried. St. Marys Cement Company Limited owns a pier to the east of the site. Large lake vessels are able to use this dock to load cement or unload gypsum/coal. There are no significant hazards from water related incidents that can not be mitigated. Details are provided in Site Evaluation Report “Summary Report: Site Evaluation of External Human Induced Events” [AMEC NSS 2009d].

Road - The nearest major road is the Macdonald-Cartier Freeway (the Highway of Heroes section of the 401). Details of road-related hazard analysis is provided in Site Evaluation Report “Summary Report: Site Evaluation of External Human Induced Events” [AMEC NSS 2009d]. There are no significant hazards from road related incidents that can not be mitigated.

2.5.4 Fishing

Major cold water streams are along the northern shore of Lake Ontario. These streams are concentrated within 50 km of the site. The major species of fish found either inhabiting or migrating up these streams during the spawning season are the salmonoids: coho salmon, Chinook salmon, rainbow trout, brown trout, and lake trout. Highest fishing activities are found at the branches of these streams.

Sport fish likely to be caught and consumed in lakeshore area include rainbow trout, salmon, lake trout, brown trout, and walleye. They are highly mobile, migrate about the entirety of Lake Ontario, are unlikely to be strongly attracted to any feature of the lake bottom diffuser, and feed primarily on schooling fish species that are equally migratory and therefore would not be exposed to the discharge for extended periods [Barry Myler, 2008].

For more information, refer to the Darlington Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006].

2.5.5 Biological Data

The biological data that have been collected in and around the region are collated in the Site Specific Survey Report [OPG 2006] and the REMP Report [OPG 2008c]. These data have been used in the pathway analysis. For further details, refer to Section 3.3 and Section 3.4 of this report.

2.5.6 Baseline ambient radioactivity and Pre-existing Hazardous substances

There has been an ongoing program to monitor the radioactivity in the environment in the region surrounding Darlington. It is summarized in detail in the Radiological Environmental Monitoring Program [OPG 2008c]. It covers:

• External Gamma dose rates in the air • Tritium Oxide and C-14 concentrations in air • Tritium Oxide concentrations in precipitation and gross beta activity in deposition

rates via dry/wet fallout • Tritium Oxide, C-14 and I-131 concentrations in milk

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• Tritium oxide and gross beta activity concentration in drinking water and lake water

• Tritium Oxide, Organically bound Tritium (OBT), C-14 and gamma emitting radionuclide concentrations in aquatic sample (e.g.: fish (HTO, C-14, gamma), lake sediment (C-14, gamma), lake water and beach sand (gamma).)

• Tritium Oxide, gamma and C-14 concentrations in terrestrial samples (e.g.: fruit (HTO, C-14), vegetable (HTO, C-14) and honey (gamma).)

Figure 2.5-1 displays the sampling locations used in this monitoring program. In 2007 an additional study of Baseline Assessment of Radiation and Radioactivity [AMEC NSS 2009e] was initiated to complement the Radiological Environmental Monitoring Program [OPG 2008c] that was already in place by collecting additional samples to ensure that all feasible human consumption pathways are included and extending the number of analyzed radionuclides. REMP sampling program [OPG 2008c] has collected samples over a period of in excess of ten years. Additional sampling has been conducted for the OPG New Nuclear Environmental Assessment to collect additional radionuclide data for the specified reactor types. Triplicate samples and field and trip blanks have been collected over a period of 8 months and the sampling program is on-going. The quantity of baseline data will be sufficient to conduct a meaningful comparison with the future monitoring data.

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Figure 2.5-1: Darlington Environmental Monitoring Locations [OPG 2008c]

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3.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR NORMAL OPERATIONS

3.1 Radioactive Source Parameters for Normal Discharges to Air

Objective 1 is addressed in this section. Airborne radioactive materials will be generated during the normal operation of the OPG New Nuclear at Darlington. Airborne materials discharges could come from the following areas:

• Reactor building

• Used fuel storage bay area (and used fuel reception bays)

• Decontamination centre

• D2O handling area (ACR-1000)

• Active ventilation exhausts (e.g.: from solid and liquid waste management, chemical lab, etc.)

All active or potentially active gases vapour and airborne particulate contained in the ventilation exhaust flow will be treated through a series of high efficiency filters and monitored before the air is released to the environment via the exhaust stack. The estimated maximum emissions from the three reactor designs are tabulated in Table 3.1-1 [OPG 2009d].

The information associated with the projected airborne release parameters from the proposed OPG New Nuclear at Darlington is summarized in Table 3.1-2.

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Table 3.1-1: Estimated maximum Airborne Emission to the Environment [OPG 2009d]

Emissions from different reactor types (Bq/y)

AP1000 (4 units) EPR(3 units) ACR-1000 (4 units)

H-3 5.18E+13 H-3 2.00E+13 H-3 4.80E+14*

C-14 1.08E+12 C-14 8.10E+11 C-14 1.10E+12

Iodine Iodine Iodine

I-131 1.78E+10 I-131 9.77E+08 I-131 6.40E+07

I-133 5.92E+10 I-133 3.55E+09

Noble gas Noble gas Noble gas

Kr-85m 5.33E+12 Kr-85m 1.67E+13 Total 2.36E+14

Kr-85 6.07E+14 Kr-85 3.77E+15

Kr-87 2.22E+12 Kr-87 5.88E+12

Kr-88 6.81E+12 Kr-88 2.00E+13

Xe-131m 2.66E+14 Xe-131m 3.89E+14

Xe-133m 1.29E+13 Xe-133m 2.00E+13

Xe-133 6.81E+14 Xe-133 9.55E+14

Xe-135m 1.04E+12 Xe-135m 1.55E+12

Xe-135 4.88E+13 Xe-135 1.33E+14

Xe-138 8.88E+11 Xe-138 1.33E+12

Ar-41 5.03E+12 Ar-41 3.77E+12

Particulate Particulate Particulates

Cr-51 9.03E+07 Cr-51 1.08E+07 Total 1.90E+08

Mn-54 6.36E+07 Mn-54 6.33E+06

Co-57 1.21E+06 Co-57 9.10E+05

Co-58 3.40E+09 Co-58 5.33E+07

Co-60 1.29E+09 Co-60 1.22E+07

Fe-59 1.17E+07 Fe-59 3.11E+06

Sr-89 4.44E+08 Sr-89 1.78E+07

Sr-90 1.78E+08 Sr-90 6.99E+06

Zr-95 1.48E+08 Zr-95 1.11E+06

Nb-95 3.70E+08 Nb-95 4.66E+06

Ru-103 1.18E+07 Ru-103 1.89E+06

Ru-106 1.15E+07 Ru-106 8.66E+04

Sb-125 9.03E+06 Sb-125 6.77E+04

Cs-134 3.40E+08 Cs-134 5.33E+06

Cs-136 1.26E+07 Cs-136 3.66E+06

Cs-137 5.33E+08 Cs-137 9.99E+06

Ba-140 6.22E+07 Ba-140 4.66E+05

Ce-141 6.22E+06 Ce-141 1.44E+06

* From [OPG 2008a]. Corresponds to the maximum emission of Tritium to air without the use of Tritium Removal Facility.

The activity of individual radionuclides for the category of noble gas and particulate is not available for ACR-1000 reactor design [OPG 2009d].

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Table 3.1-2: Airborne Release Parameters [OPG 2009d]

Reactor types

Parameters AP1000 EPR ACR-1000

Elevation of release

point (m) 48.8 64.3 71.0

Volumetric emission

rate (m3/s) 80.9 114.4 69.4

Building height (m) 2 71.3 71.3 65.8

Gas temperature (ºC ) 29.5 48.9 40.0

Calculations of doses have been done with an ambient air temperature of 20°C consistent with previous assessments [OPG 2008c], which is the highest daily mean temperature as discussed in Section 2. This results in a conservative estimate of doses to the public because it will reduce the height of the air plume due to the decreased buoyancy of the plume and correspondingly increase the contaminant transfer factor (from the source to receptor).

3.2 Radioactive Source Parameters for Normal Discharges to Surface Water and Groundwater

This section addresses Objective 2. Liquid radioactive effluent will be generated during the normal operation. The liquid radioactive effluent will be, after appropriate treatment, discharged to Lake Ontario. The location of the effluent discharge, assumed to be the point source for both once-through cooling option and cooling tower option, is illustrated in Figure 3.3.1-1. The estimated maximum radioactive emissions to water from three reactor designs are summarized in Table 3.2-1 [OPG 2009d]. It should be noted that the activity of individual radionuclide for the category of gross beta and gamma is not available for ACR-1000 reactor design [OPG 2009d].

The discharge rate and aquatic plume parameters associated with the projected waterborne release from the proposed OPG New Nuclear at Darlington are given in Table 3.2-2 and Table 3.2-3.

Under normal operational conditions, all liquid radioactive materials will be collected and treated prior to discharge to the environment (Lake Ontario). There will be no direct discharge of liquid radioactive materials to groundwater under normal operations.

2 Revision 1 of [OPG 2009d] was used for the assessment. The preliminary analysis shows that using EPR specific

building height will result in lower public dose than the previous estimate, i.e., the previous estimate represents a more conservative assessment.

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Table 3.2-1: Estimated maximum waterborne emission to the environment [OPG 2009d]


AP1000 (4units) EPR(3 units) ACR-1000 (4 units)

H-3 1.49E+14 H-3 1.84E+14 H-3 1.40E+15**

C-14 1.0E+11* C-14 1.0E+11* C-14 8.40E+10

Gross beta-gamma Gross beta-gamma

Gross beta-gamma

Na-24 2.41E+08 Na-24 6.80E+08 Total 5.60E+10

Cr-51 2.74E+08 Cr-51 1.14E+08

Mn-54 1.92E+08 Mn-54 5.99E+07

Fe-55 1.48E+08 Fe-55 4.55E+07

Fe-59 2.96E+07 Fe-59 1.11E+07

Co-58 4.97E+08 Co-58 1.72E+08

Co-60 6.51E+07 Co-60 2.00E+07

Zn-65 6.07E+07 Zn-65 1.89E+07

W-187 1.92E+07 W-187 5.11E+07

Np-239 3.55E+07 Np-239 6.44E+07

Br-84 2.96E+06 Br-84 0

Rb-88 4.00E+07 Rb-88 0

Sr-89 1.48E+07 Sr-89 5.55E+06

Sr-90 1.48E+06 Sr-90 0

Sr-91 2.96E+06 Sr-91 8.88E+06

Y-91m 1.48E+06 Y-91m 0

Y-93 1.33E+07 Y-93 4.00E+07

Zr-95 3.40E+07 Zr-95 1.44E+07

Nb-95 3.11E+07 Nb-95 1.11E+07

Mo-99 8.44E+07 Mo-99 1.94E+08

Tc-99m 8.14E+07 Tc-99m 1.89E+08

Ru-103 7.30E+08 Ru-103 2.79E+08

Rh-103m 7.30E+08 Rh-103m 2.79E+08

Ru-106 1.09E+10 Ru-106 3.39E+09

Rh-106 1.09E+10 Rh-106 3.39E+09

Ag-110m 1.55E+08 Ag-110m 4.88E+07

Ag-110 2.07E+07 Ag-110 6.66E+06

Te-129m 1.78E+07 Te-129m 6.66E+06

Te-129 2.22E+07 Te-129 4.44E+06

Te-131m 1.33E+07 Te-131m 3.44E+07

Te-131 4.44E+06 Te-131 6.66E+06

I-131 2.09E+09 I-131 3.80E+09

Te-132 3.55E+07 Te-132 5.33E+07

I-132 2.43E+08 I-132 1.28E+08

I-133 9.92E+08 I-133 3.87E+09

I-134 1.20E+08 I-134 0

Cs-134 1.47E+09 Cs-134 2.94E+08

I-135 7.36E+08 I-135 1.67E+09

Cs-136 9.32E+07 Cs-136 3.44E+07

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AP1000 (4units) EPR(3 units) ACR-1000 (4 units)

Cs-137 1.97E+09 Cs-137 3.90E+08

Ba-137m 1.84E+09 Ba-137m 3.64E+08

Ba-140 8.17E+08 Ba-140 4.67E+08

La-140 1.10E+09 La-140 8.47E+08

Ce-141 1.33E+07 Ce-141 5.55E+06

Ce-143 2.81E+07 Ce-143 6.77E+07

Pr-143 1.92E+07 Pr-143 5.55E+06

Ce-144 4.68E+08 Ce-144 1.47E+08

Pr-144 4.68E+08 Pr-144 1.47E+08

All others 2.96E+06 All others 2.22E+06

* No data on waterborne emission of C-14 were provided in OPG report for PWR reactor designs

[OPG 2009d]. Based on published data [Kunz C. 1985], the discharge of C-14 in liquid was estimated at 1.0E11 Bq/y (or 3.19E+3 Bq/s) for PWR reactors assuming the total electrical output

of 5124 GW year.

** From [OPG 2008a]. Corresponds to the maximum emission of Tritium to water without the use of Tritium Removal Facility.

Table 3.2-2: Discharge Rate of Liquid Effluents [OPG 2009d]

Reactor type

Cooling options

AP1000 (4units)

EPR (3

units)3 ACR-1000(4 units)

Once through(L/s) 4 222700 222700 253000

Mechanical draft

cooling(L/s)) 1400 1400 1200

Natural draft cooling(L/s))

1200 1200 1200

3 Revision 1 of [OPG 2009d] was used for this assessment. It is expected that the impact of using EPR specific

discharge rate on the total dose is negligible. 4 It is expected that the impact of using EPR specific discharge rate on the public dose is very small.

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Table 3.2-3: Aquatic Plume Parameters [OPG 2008c]

Water plume parameters Value

Once-through cooling 1.15 Effluent

recirculation factor Cooling tower 1

Once-through cooling 7.10E-06 Proportionality

coefficient Cooling tower 3.39E-7

Once-through cooling 5 7 Initial dilution

Cooling tower 1

Velocity of flow to West(m/s) 0.10

Fraction of time current to West 0.35

Velocity of flow to East(m/s) 0.16

Fraction of time current to East 0.38

.

5 Initial dilution of a once through cooling system is 7 due to the diffuser. Without a diffuser the dilution is 1. The above results imply that a diffuser is included in the design.

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3.3 Identification of Potential Critical Groups and Their Characterization

3.3.1 Selection of Potential Critical Groups

Doses received by individual members of the public, as a result of a given radionuclide release, will vary depending on factors such as proximity to the release, dietary and behavioural habits, age and metabolism, and variations in the environment. To determine the radiological effects on members of the public, a concept of critical groups is adopted in this study. A critical group is a fairly homogeneous group of people whose location, characteristics or diet, cause them to receive doses higher than the average received by typical people in all other groups in the exposed population. Based on the Darlington Radiological Environmental Monitoring Program (REMP) review [OPG 2008c], which was conducted in 2008, eleven Potential Critical Groups (PCGs) have been identified in this study. These PCGs were updated for the NND location, summarized in Table 3.3.1-1 and are illustrated in Figure 3.3.1-1.

Table 3.3.1-1: Summary of Potential Critical Groups [OPG 2008c]

No. Potential critical group6 Wind sector (direction to)-

distance from OPG NND

Note

1 Farm WNW – 2.8 km Existing PCG used in REMP

2 Dairy Farm N - 2.3 km The closest Dairy farm to the NND

site

3 Industrial (St. Marys Cement)

NE - 0.8 km Existing PCG used in REMP

4 West East Beach Residents ENE – 2.2 km Existing PCG used in REMP

5 Camper W – 5.2 km Existing PCG used in REMP

6 Bowmanville Residents NE - 3.1 km Existing PCG used in REMP

7 Fisher E – 1.1 km Revised from an existing PCG used in

OPG’s REMP

8 Oshawa Residents WNW – 7.3 km Existing PCG used in REMP

9 Rural Residents NE – 1.8 km Existing PCG used in REMP

10 New resident NNW - 3.0 km New identified PCG

11 New industrial NW - 3.6 km New identified PCG

6 The representative location of potential critical group 1-9 was based on a previous work [OPG 2008c]. It is expected

that the dose to workers in Clarington Energy Park will be bounded by that for St. Marys Cement plant which is down wind and closest to the NND.

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Figure 3.3.1-1: Locations of potential critical groups

L

ocati

on

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Ou

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sed

in

th

e S

tud

y

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Compared to the existing Darlington NGS’s REMP program, the following changes have been implemented: 1. The location of the “Fisher” Potential Critical Group has been amended to place in

the area which is close to the NND, located 1100 m offshore from the new plant. 2. One additional “Industrial” PCG and one “Residential” PCG were included to reflect

future land use. These two groups represent the future locations of urban residents and industrial site which are the closest to the NND on the basis of the Durham Regional Population Growth estimate [Regional Municipality of Durham 2008]. The locations of these two groups are illustrated in Figure 3.3.1-1.

As different age classes may have quite different habits, intake rates, and dose coefficients, people in each PCG are grouped into three age classes as follows [CSA 2008].

• one year old infant, • 10 year old child, and • Adult

3.3.2 Characterization of Potential Critical Groups

Consistent with the previous work [OPG 2008c, 2008d], the following assumptions were made regarding the characterization of the potential critical groups:

• All potential critical groups, except fisher group, industrial /commercial group and camper group, live at the location 100 percent of the time.

• Fisher group only resides at the location for 1 percent of the time. • Industrial/Commercial groups are considered to be at the location for only

23 percent of the time since they do not reside at the location.

• Camper group live at the location for 50% of the time. • The fraction of local food intake is as determined by the site specific survey

[OPG 2006].

• Drinking water is consumed from local sources as determined by the site specific survey.

• Only dairy farm residents ingest local cow’s milk. • No local grain products are consumed by humans. • Beach recreational activities for all potential critical groups except fisher group

and industrial/commercial groups are assumed to take place at Darlington Provincial Park.

Detailed characteristics of potential critical groups, including local water and food usage and food and water intake rates, can be found in the site survey, Darlington REMP review report and CSA standard [OPG 2006 and 2008b, and CSA 2008]. It was assumed that the characteristics of two additional groups identified for this study, new resident group and new industrial group, were consistent with those of Bowmanville residents and the existing Industrial (St. Marys Cement) group, respectively.

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3.3.3 Determination of average dilution factor in fish home range

For the site evaluation purpose, the concept of home range was adopted to determine the concentration in fish due to waterborne emission. This assumed that fish would only be immersed in average water concentrations over a home range area. A recent study on the fish home range recommended that a 2 km x 2 km area would be a conservative estimate of the species of concern [SENES 2009a].

On this basis, the average dilution factors (DF) in this area were estimated for both once-through cooling and cooling tower options for all three designs with the assumption that the home range was centered at outfall. The equation used to calculate the average dilution factor is given below. The parameters used in the calculation of the average dilution factor are summarized in Table 3.3.3-1.

Average DF= ∫1000

0)( dxxDF [CSA 2008]

Where DF(x)=

17.1

17.1

117.1

1

17.0

1000

+

Oi

CV

DxUQ

d πκβ

Table 3.3.3-1: Parameters Used for the Estimate of Average DF [CSA 2008, OPG 2009d, SENES 2009a]

Parameters Once-through Cooling tower

Initial dilution , D0 7 1

Average water depth in this range,

d, (m) 15 15

Recirculation factor, β 1.15 1

Proportionality coefficient, κ 7.1E-6 3.39E-07

ACR-1000 253000 1200 Discharge rate, QV, (L/s) AP1000/EPR 222700 1400

Flow speed, Uc, (m/s) 0.16 0.16

Range, xi (m) 1000 1000

The average dilution factors in this range are summarized in Table 3.3.3-2.

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Table 3.3.3-2: Average Dilution Factors in Home Range Area

Once-through Cooling tower Designs

Dilution Factor Dilution Factor

ACR-1000 8.06 28.1

AP1000/EPR 8.19 24.3

3.3.4 Dairy cow’s diet

For the site evaluation purpose, it was assumed that the dairy cow’s diet consisted of forage and generic feed crops. Furthermore, it was conservatively assumed that the forage accounted for 30% of the dry weight of the total feed consumed by the dairy cow and the generic feed crops accounted for 70% [SENES 2009b]. It was also assumed that the hold up time for forage and generic feed crops was one day and three months, respectively [SENES 2009b].

3.4 Radioactive Effluent Dose Consequence for Normal Operations

3.4.1 Environmental Pathway Model

Members of the public could be exposed to radiation and radioactivity via a variety of pathways, which may include:

(a) Air inhalation

(b) Air immersion (external exposure)

(c) Water ingestion

(d) Water immersion (external exposure – swimming and bathing)

(e) Soil external exposure or soil ground shine

(f) Terrestrial plant ingestion – fruits and vegetables

(g) Terrestrial animal product ingestion – milk, meat, eggs

(h) Aquatic animal ingestion – fish

(i) Sediment external exposure – beach ground shine

(j) Incidental soil and sediment ingestion

A generalized model of environmental radioactivity transport and human exposure pathways is illustrated in Figure 3.4.1-1.

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Figure 3.4.1-1: Environmental transfer model [CSA 2008]

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Using the concept of compartments, each environmental source/receptor is presented as a numbered compartment and the quantity in ‘compartment ’ is denoted by Xi. Transfer from ‘compartment ’ to ‘compartment j’ is characterized by a transfer parameter Pij and the amount present in ‘compartment j’ under steady-state conditions due to transfer from ‘compartment ’ to ‘compartment j’ is therefore PijXi. The magnitude of the quantity (concentration or dose) represented by any ‘compartment j’ is

i

i

ijj XPX ∑=

Where the summation is over all ‘compartment ’ transfer into ‘compartment j’.

The compartments and transfer parameters, along with their corresponding units are summarized in Tables 3.4.1-1 and 3.4.1-2.

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Table 3.4.1-1: Transfer Compartments and their Units [CSA 2008]

Compartment Number

Compartment Name Units*

0 Source Bq•s-1**

1 Atmosphere Bq•m-3**

1a Atmosphere (HTO from HT) Bq•m-3

2 Surface Water (lake, river or ocean) Bq•L-1

2p Surface Water (pond) Bq•L-1

2w Ground Water (well) Bq• L-1

3area Soil Surface Bq•m-2

3mass Bulk Soil Bq•kg-1 dw

3spw Soil Water Bq• L-1

4 Forage and Plant Produce Bq•kg-1 fw

5 Animal Produce Bq•kg-1 fw

5m Mother’s Milk*** Bq•L-1

6 Aquatic Animals (fish and shellfish) Bq•kg-1 fw

7 Aquatic Plants Bq•kg-1 fw

8 Sediment Bq•kg-1 dw

9 Dose Sv•a-1

Notes:

* Units involve mass (kg), fresh weight (fw) or dry weight (dw) as indicated.

** For noble gases, source units are BqMeV•s-1, atmosphere units are BqMeV•m-3.

*** CSA N288.1 does not explicitly address mother’s milk as a transfer compartment.

The parameters for this compartment have been taken from COG DRL Guidance

document [Ecometrix Inc. 2008].

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Table 3.4.1-2: Transfer Parameters and their Units [CSA 2008]

Compartments Transfer Parameter

From To

Parameter Units

P01 Source Atmosphere s•m-3

P11a Atmosphere (HT) Atmosphere (HTO)* Unitless

P3area1 Surface Soil Atmosphere m2•m-3

P3mass1† Bulk Soil Atmosphere kg dw • m-3

P12p Atmosphere Surface Water (pond) m3•L-1

P13area Atmosphere Surface Soil m3•m-2

P13mass Atmosphere Bulk Soil m3•kg-1 dw

P13spw Atmosphere Soil Water m3•L-1

P14 Atmosphere Forage and Plant Produce m3•kg-1 fw

P15 Atmosphere Animal Produce m3•kg-1 fw

P(i)19 Atmosphere Dose (inhalation) Sv•a-1•Bq-1•m3

P(e)19 Atmosphere Dose (immersion) Sv•a-1•Bq-1•m3

P02 Source Surface Water (lake, river, ocean) s•L-1

P3spw1‡ Soil Water Atmosphere L • m-3

P3area2p Surface Soil Surface Water (pond) m2•L-1

P3area2w Surface Soil Groundwater (well) m2•L-1

P3area3spw Surface Soil Soil Water m2•L-1

P3spw2w Soil Water Groundwater (well) Unitless

P3spw2p Soil Water Surface Water (pond) Unitless

P23area Surface Water Surface Soil L•m-2

P23mass Surface Water Bulk Soil L•kg-1 dw

P23spw‡ Surface Soil Soil Water Unitless

P24 Surface Water Forage and Plant Produce L•kg-1 fw

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Compartments Transfer

Parameter

From To

Parameter Units

P25 Surface Water (lake, river) Animal Produce L•kg-1 fw

P2p5 Surface Water (pond) Animal Produce L•kg-1 fw

P2w5 Well Water Animal Produce L•kg-1 fw

P26 Surface Water Aquatic Animal L•kg-1 fw

P27 Surface Water Aquatic Plants L•kg-1 fw

P28 Surface Water Sediment L•kg-1 dw

P(i)29 Surface Water Dose (ingestion) Sv•a-1•Bq-1•L

P(i)2w9 Well Water Dose (ingestion) Sv•a-1•Bq-1•L

P(e)29 Surface Water Dose (immersion) Sv•a-1•Bq-1•L

P(e)2w9 Well Water Dose (immersion) Sv•a-1•Bq-1•L

P3mass4 Bulk Soil Forage and Plant Produce kg dw•kg-1 fw

P3mass5 Bulk Soil Animal Produce kg dw•kg-1 fw

P(i)3mass9 Bulk Soil Dose (ingestion) Sv•a-1•Bq-1•kg dw

P(e)3area9 Surface Soil Dose (groundshine) Sv•a-1•Bq-1•m2

P45 Forage and Plant Produce Animal Produce kg fw•kg-1 fw

P49 Forage and Plant Produce Dose (ingestion) Sv•a-1•Bq-1•kg fw

P59 Animal Produce Dose (ingestion) Sv•a-1•Bq-1•kg fw

P69 Aquatic Animals Dose (ingestion) Sv•a-1•Bq-1•kg fw

P79 Aquatic Plants Dose (ingestion) Sv•a-1•Bq-1•kg fw

P(i)89 Sediment Dose (ingestion) Sv•a-1•Bq-1•kg dw

P(e)89 Sediment Dose (beachshine) Sv•a-1•Bq-1•kg dw

Notes:

* P11a is a composite transfer parameter encompassing transfer of HT to soil, oxidation of HT to HTO by

soil microbes, re-emission of HTO from soil to atmosphere, and dispersion of HTO in the atmosphere. † For C-14 and radioiodine only ‡ For HTO only

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3.4.2 Pathway Analysis Model and its Implementation

The environmental transfer model discussed above has been programmed in a computer code called “Integrated Model for Probabilistic Assessment of Contaminated Transport” (IMPACT) [Ecometrix Inc 2006a]. The IMPACT program is a customizable tool that allows the user to assess the transport and fate of contaminants through a user-specified environment. It also enables the quantification of the human exposure to those environmental contaminants and the calculation of DRLs for nuclear facilities (power generating stations, research reactors, waste management facilities). It covers all potential exposure and release scenarios, including atmospheric and aquatic pathways. Quality assurance during code development has followed CSA standard N286.7 [CSA 1999] and CNSC Guideline G-149 [CNSC 2000]. The code verification and validation was documented in Tool Qualification Report [Ecometrix Inc, 2006b].

In this assessment, the IMPACT code (version 5.2.2) was used for the calculation of doses to members of the public. The code is essentially consistent with CSA N288.1-08, reflecting the updates in scientific developments related to the understanding of environmental transport models and human dosimetry. Detailed discussion of various mathematical models and key model parameters comprising the IMPACT database can be found in CSA N288.1 [CSA 2008].

3.4.3 Assessment Results

This section summarizes dose predictions for potential critical groups using IMPACT code (version 5.2.2) assuming a life time of sixty years for all designs considered in the assessment.

Selection of limiting radionuclides

As the report [OPG 2009d] did not provide the breakdown of radionuclides in particulate (air emission) and gross beta-gamma (liquid effluent) for ACR-1000 design, the limiting radionuclide in particulate and gross beta-gamma for this design must be determined before the detailed calculation of dose to the public. Table 3.4.3-1 provides the radionuclides of importance released from CANDU reactor [OPG 2008c], which will be used for ACR reactor for the determination of the limiting radionuclide in air (particulate) and water (gross beta-gamma).

It should be noted that this method of using the limiting radionuclide, the radionuclide resulting in the highest dose to the public per unit release, will result in a conservative estimate.

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Table 3.4.3-1: Radionuclides of Importance Released from ACR Reactor [OPG 2008c]*

Radionuclides of importance in

particulate in air emission

Radionuclides of importance in water

Gross Beta-gamma emission

Co-60 Co-60

Cs-134 Cs-137

Cs-137 Eu-154

Hg-203 Fe-59

Ru-106 Gd-153

Sb-124 Mn-54

Sc-46b Nb-95

Zn-65 Sb-124

Sb-125

Zn-65

Zr-95

Note:

* The report [OPG 2009d] does not provide the list of radionuclides in particulate and gross beta-gamma released from ACR reactor design. The individual radionuclides presented in

this table are based on a list of radionuclides released from CANDU reactor with an assumption that the emission of radionuclide from ACR-1000 and CANDU reactor will be

similar.

To determine the limiting radionuclide to represent gross measurement of particulate in air and beta-gamma in water emission, the doses to each potential critical group selected were estimated using IMPACT assuming a unit release rate (1 Bq/second) of each radionuclide identified above. The results are presented in Table 3.4.3-2 and Table 3.4.3-3. Those limiting radionuclides identified through this exercise will be used in the subsequent dose calculations.

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Table 3.4.3-2: Limiting radionuclides in particulate released from ACR-1000 reactor

Potential Critical Group Age group

Limiting Radionuclide Annual Dose (Sv)

Bowmanville Residents 1 yr old Co-60 4.36E-10

10 yr old Cs-137+ 3.72E-10

Adult Cs-137+ 5.19E-10

Camper 1 yr old Co-60 1.04E-09

10 yr old Co-60 8.10E-10

Adult Cs-137+ 8.71E-10

Dairy Farm 1 yr old Cs-137+ 1.36E-09

10 yr old Cs-137+ 9.44E-10

Adult Cs-137+ 1.38E-09

Fisher 1 yr old Ru-106+ 8.59E-13

10 yr old Ru-106+ 9.18E-13

Adult Ru-106+ 6.71E-13

Farm 1 yr old Co-60 2.26E-09

10 yr old Cs-137+ 1.89E-09

Adult Cs-137+ 2.54E-09

New Industry Adult Co-60 1.38E-10

New Residents 1 yr old Co-60 9.59E-10

10 yr old Co-60 7.54E-10

Adult Cs-137+ 9.03E-10

Oshawa Residents 1 yr old Co-60 4.57E-10

10 yr old Co-60 3.59E-10

Adult Cs-137+ 4.19E-10

Rural Residents 1 yr old Co-60 8.67E-10

10 yr old Co-60 6.82E-10

Adult Cs-137+ 7.94E-10

St. Marys Adult Co-60 2.93E-10

West East Beach Residents 1 yr old Co-60 3.22E-09

10 yr old Co-60 2.50E-09

Adult Cs-137+ 2.53E-09

Note: “+” means the contribution from its progeny is included

From the preceding table, the one year old child in West East Beach Residents group had the limiting dose from particulate due to Co-60. Therefore, the limiting radionuclide in particulate (airborne emission) released from ACR-1000 reactor design is assumed to be Co-60.

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Table 3.4.3-3: Limiting Radionuclide in Beta-Gamma Released from ACR-1000 Reactor via Water pathway

Potential Critical Group Age group Limiting

Radionuclide

Annual

Dose (Sv)

Bowmanville Residents 1 yr old Co-60 8.02E-12

10 yr old Co-60 6.04E-12

Adult Cs-137+ 6.37E-12

Camper 1 yr old Zn-65 2.00E-12

10 yr old Cs-137+ 2.43E-12

Adult Cs-137+ 5.71E-12

Dairy Farm 1 yr old Fe-59 1.57E-12

10 yr old Fe-59 1.21E-12

Adult Fe-59 1.21E-12

Fisher 1 yr old Zn-65 1.09E-11

10 yr old Cs-137+ 1.50E-11

Adult Cs-137+ 4.49E-11

Farm 1 yr old Fe-59 1.58E-12

10 yr old Fe-59 1.22E-12


New Industry Adult Cs-137+ 2.96E-13

New Residents 1 yr old Fe-59 2.01E-12

10 yr old Fe-59 1.45E-12

Adult Cs-137+ 1.90E-12

Oshawa Residents 1 yr old Co-60 4.17E-12

10 yr old Cs-137+ 3.95E-12

Adult Cs-137+ 6.16E-12

Rural Residents 1 yr old Fe-59 1.66E-12

10 yr old Fe-59 1.26E-12


St. Marys Adult Cs-137+ 2.96E-13

West East Beach Residents 1 yr old Co-60 2.31E-12

10 yr old Co-60 1.76E-12

Adult Cs-137+ 1.83E-12

Note: “+” means the contribution from its progeny is included

From the preceding table, the Adult in Fisher group had the limiting dose from gross beta-gamma in water emission resulting from Cs-137. Therefore, the limiting radionuclide in gross beta-gamma (waterborne emission) released from ACR-1000 reactor design is assumed to be Cs-137.

Estimated doses to members of the public

Doses to members of the public, represented by potential critical groups, were calculated for three reactor designs using IMPACT 5.2.2 based on the emission data described previously. Two cooling options, including once through cooling option and cooling tower option, were considered separately for each design. The results are

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discussed below. It should be noted that the following radionuclides were not included for dose calculation.

• Co-57 (in particulate) released from AP1000 and EPR accounts for 0.02% and 0.65% of total activity, respectively. It is expected the dose associated with the emission of this radionuclide only accounts for a small fraction of total dose resulting from the emission of particulate and therefore the exclusion of Co-57 for dose calculation has a negligible effect on the total dose estimated.

• Some waterborne radionuclides including Ag-110, Br-84, Ce-143, Pr-143, Sr-91, Te-129, Te-129m, Te-131, Te-131m, W-187, Y-91m, Y-93, and the category of “all others” for which the information of individual radionuclides was not provided by vendors were not modeled in IMPACT. The sum of the activity (released from the AP1000 or EPR) from these radionuclides is 0.5% and 1% of the gross beta-gamma activity, respectively. It is expected the doses associated with the emission of these radionuclides only account for a small fraction of total dose resulting from the waterborne emission and therefore the exclusion of these radionuclides for dose calculation has a negligible effect on the total dose estimated.

• Other radionuclides released to water are Ba-137m, Pr-144, and Rh-106, in total accounting for 35% for AP1000 and 18% for EPR of the total gross beta-

gamma, and have very short half lives (Rh-106: 29.8 s; Ba-137m: 2.5 min; and

Pr-144: 17.3 min). Therefore, it is expected that these isotopes will not be available for human intake due to decay to the stable elements (Rh-106—Pd-106, Ba-137m—Ba-137, Pr-144—Nd-144) prior to release to the lake and in the environment. Therefore, these three radionuclides will not be considered in the subsequent dose calculations.

Once-through cooling option

The doses to each of eleven PCG, including three age classes in each PCG, are presented in Table 3.4.3-4 through Table 3.4.3-6 for once-through cooling operation.

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Table 3.4.3-4: Doses to Potential Critical Groups Resulting from the Operation of AP1000 Reactor (Once-Through Cooling)


Dose due to

airborne emission(mSv)

Dose due to

waterborne emission(mSv)

Total annual

dose (mSv)

Bowmanville Residents 1 yr old 5.15E-04 1.13E-05 5.27E-04

10 yr old 3.98E-04 7.62E-06 4.05E-04

Adult 3.83E-04 9.57E-06 3.92E-04

Camper 1 yr old 4.96E-04 3.76E-06 5.00E-04

10 yr old 3.91E-04 3.14E-06 3.94E-04

Adult 3.74E-04 4.58E-06 3.78E-04

Dairy Farm 1 yr old 4.90E-03 3.48E-07 4.90E-03

10 yr old 1.93E-03 3.32E-07 1.93E-03

Adult 1.40E-03 3.37E-07 1.40E-03

Fisher 1 yr old 2.43E-05 5.07E-06 2.94E-05

10 yr old 2.04E-05 9.29E-06 2.97E-05

Adult 1.87E-05 1.84E-05 3.71E-05

Farm 1 yr old 2.37E-03 6.95E-07 2.37E-03

10 yr old 1.80E-03 5.26E-07 1.80E-03

Adult 1.77E-03 5.81E-07 1.77E-03

New Industry Adult 4.54E-05 2.50E-06 4.79E-05

New Residents 1 yr old 7.28E-04 1.07E-05 7.39E-04

10 yr old 5.76E-04 7.19E-06 5.83E-04

Adult 5.46E-04 9.14E-06 5.55E-04

Oshawa Residents 1 yr old 3.38E-04 5.76E-06 3.44E-04

10 yr old 2.71E-04 4.30E-06 2.75E-04

Adult 2.56E-04 5.79E-06 2.62E-04

Rural Residents 1 yr old 7.45E-04 2.59E-06 7.48E-04

10 yr old 5.69E-04 1.78E-06 5.71E-04

Adult 5.46E-04 2.19E-06 5.48E-04

St. Marys Adult 1.62E-04 2.50E-06 1.64E-04

West East Beach Residents 1 yr old 1.66E-03 2.97E-06 1.66E-03

10 yr old 1.34E-03 2.05E-06 1.34E-03

Adult 1.28E-03 2.52E-06 1.28E-03

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Table 3.4.3-5: Doses to Potential Critical Groups Resulting from the Operation of EPR Reactor (Once-Through Cooling)


Dose due to


Dose due to


Total annual

dose (mSv)


10 yr old 1.50E-04 8.47E-06 1.59E-04

Adult 1.54E-04 1.08E-05 1.65E-04

Camper 1 yr old 1.93E-04 4.01E-06 1.97E-04

10 yr old 1.56E-04 3.16E-06 1.59E-04

Adult 1.57E-04 4.36E-06 1.62E-04


10 yr old 3.78E-04 1.60E-07 3.78E-04

Adult 3.42E-04 1.66E-07 3.42E-04

Fisher 1 yr old 1.82E-05 4.45E-06 2.27E-05

10 yr old 1.43E-05 7.56E-06 2.19E-05

Adult 1.42E-05 1.32E-05 2.74E-05

Farm 1 yr old 8.05E-04 5.18E-07 8.05E-04

10 yr old 6.46E-04 3.89E-07 6.46E-04

Adult 6.70E-04 4.56E-07 6.70E-04



10 yr old 2.64E-04 8.37E-06 2.73E-04

Adult 2.67E-04 1.07E-05 2.77E-04


10 yr old 1.37E-04 4.48E-06 1.41E-04

Adult 1.37E-04 5.93E-06 1.43E-04


10 yr old 2.66E-04 1.87E-06 2.67E-04

Adult 2.69E-04 2.36E-06 2.71E-04


West Each Beach Residents 1 yr old 1.16E-03 3.16E-06 1.16E-03

10 yr old 9.13E-04 2.15E-06 9.15E-04

Adult 9.13E-04 2.71E-06 9.15E-04

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Table 3.4.3-6: Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 Reactor (Once-Through Cooling)


Dose due to


Dose due to


Total annual

dose (mSv)


10 yr old 5.13E-04 6.44E-05 5.77E-04

Adult 5.70E-04 8.26E-05 6.53E-04

Camper 1 yr old 5.30E-04 2.59E-05 5.56E-04

10 yr old 5.04E-04 2.06E-05 5.24E-04

Adult 5.34E-04 3.17E-05 5.66E-04


10 yr old 1.24E-03 1.50E-06 1.24E-03

Adult 1.26E-03 1.54E-06 1.27E-03

Fisher 1 yr old 4.63E-05 1.33E-05 5.96E-05

10 yr old 4.30E-05 3.27E-05 7.57E-05

Adult 4.05E-05 9.00E-05 1.30E-04

Farm 1 yr old 2.70E-03 3.97E-06 2.70E-03

10 yr old 2.46E-03 3.07E-06 2.46E-03

Adult 2.72E-03 3.62E-06 2.72E-03



10 yr old 8.71E-04 5.64E-05 9.28E-04

Adult 9.09E-04 7.46E-05 9.84E-04


10 yr old 4.12E-04 3.36E-05 4.46E-04

Adult 4.29E-04 4.60E-05 4.74E-04


10 yr old 9.77E-04 1.35E-05 9.90E-04

Adult 1.03E-03 1.73E-05 1.05E-03



10 yr old 3.21E-03 1.64E-05 3.23E-03

Adult 3.29E-03 2.08E-05 3.31E-03

The largest estimated dose for the three proposed designs for the once-through

cooling option is approximately 5 µSv/y for a one year old infant in the dairy farm group. This value is far below the regulatory dose limit, representing only 0.5 percent of the limit of 1 mSv/y. Airborne emission was estimated to be the major contributor to the public dose.

Cooling tower option

The doses to each of eleven PCG, including three age classes in each PCG, are presented in Table 3.4.3-7 through Table 3.4.3-9 for cooling tower option. Mechanical draft cooling method is used in this study to illustrate the effect of cooling tower

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option. It is expected that the Natural draft cooling method will produce similar results as the only difference between two cooling tower methods, from the dose calculation viewpoint, is the effluent discharge rate which is however similar for the two methods as shown in Table 3.2-2.

Table 3.4.3-7: Doses to Potential Critical Groups Resulting from the Operation of AP1000 reactor (Cooling Tower Option)


Dose due to airborne

emission(mSv)

Dose due to waterborne

emission(mSv) Total annual dose (mSv)


10 yr old 3.98E-04 7.92E-05 4.77E-04

Adult 3.83E-04 1.00E-04 4.83E-04

Camper 1 yr old 4.96E-04 2.66E-05 5.23E-04

10 yr old 3.91E-04 2.32E-05 4.14E-04

Adult 3.74E-04 3.46E-05 4.08E-04


10 yr old 1.93E-03 2.95E-06 1.94E-03

Adult 1.40E-03 2.99E-06 1.40E-03

Fisher 1 yr old 2.43E-05 2.36E-04 2.60E-04

10 yr old 2.04E-05 4.33E-04 4.53E-04

Adult 1.87E-05 8.59E-04 8.78E-04

Farm 1 yr old 2.37E-03 5.94E-06 2.38E-03

10 yr old 1.80E-03 4.52E-06 1.81E-03

Adult 1.77E-03 4.82E-06 1.77E-03



10 yr old 5.76E-04 7.47E-05 6.50E-04

Adult 5.46E-04 9.59E-05 6.42E-04


10 yr old 2.71E-04 5.31E-05 3.24E-04

Adult 2.56E-04 8.67E-05 3.43E-04


10 yr old 5.69E-04 1.79E-05 5.87E-04

Adult 5.46E-04 2.20E-05 5.68E-04



10 yr old 1.34E-03 2.13E-05 1.36E-03

Adult 1.28E-03 2.60E-05 1.31E-03

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Table 3.4.3-8: Doses to Potential Critical Groups Resulting from the Operation of EPR reactor (Cooling Tower Option)


Dose due to


Dose due to


Total annual

dose (mSv)


10 yr old 1.50E-04 8.79E-05 2.38E-04

Adult 1.54E-04 1.12E-04 2.67E-04

Camper 1 yr old 1.93E-04 2.76E-05 2.21E-04

10 yr old 1.56E-04 2.27E-05 1.79E-04

Adult 1.57E-04 3.17E-05 1.89E-04


10 yr old 3.78E-04 1.45E-06 3.80E-04

Adult 3.42E-04 1.49E-06 3.43E-04

Fisher 1 yr old 1.82E-05 2.07E-04 2.26E-04

10 yr old 1.43E-05 3.52E-04 3.66E-04

Adult 1.42E-05 6.14E-04 6.28E-04

Farm 1 yr old 8.05E-04 4.23E-06 8.09E-04

10 yr old 6.46E-04 3.19E-06 6.49E-04

Adult 6.70E-04 3.55E-06 6.73E-04



10 yr old 2.64E-04 8.69E-05 3.51E-04

Adult 2.67E-04 1.11E-04 3.78E-04


10 yr old 1.37E-04 4.94E-05 1.86E-04

Adult 1.37E-04 7.36E-05 2.11E-04


10 yr old 2.66E-04 1.91E-05 2.85E-04

Adult 2.69E-04 2.40E-05 2.93E-04



10 yr old 9.13E-04 2.24E-05 9.35E-04

Adult 9.13E-04 2.81E-05 9.41E-04

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Table 3.4.3-9: Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 reactor (Cooling Tower Option)


Dose due to


Dose due to


Total annual

dose (mSv)


10 yr old 5.13E-04 7.12E-04 1.22E-03

Adult 5.70E-04 9.21E-04 1.49E-03

Camper 1 yr old 5.30E-04 1.83E-04 7.13E-04

10 yr old 5.04E-04 1.52E-04 6.56E-04

Adult 5.34E-04 2.42E-04 7.76E-04


10 yr old 1.24E-03 1.40E-05 1.25E-03

Adult 1.26E-03 1.46E-05 1.28E-03

Fisher 1 yr old 4.63E-05 6.96E-04 7.43E-04

10 yr old 4.30E-05 1.72E-03 1.76E-03

Adult 4.05E-05 4.73E-03 4.77E-03

Farm 1 yr old 2.70E-03 3.26E-05 2.73E-03

10 yr old 2.46E-03 2.59E-05 2.48E-03

Adult 2.72E-03 3.00E-05 2.75E-03



10 yr old 8.71E-04 6.24E-04 1.50E-03

Adult 9.09E-04 8.33E-04 1.74E-03


10 yr old 4.12E-04 3.30E-04 7.42E-04

Adult 4.29E-04 5.84E-04 1.01E-03


10 yr old 9.77E-04 1.46E-04 1.12E-03

Adult 1.03E-03 1.87E-04 1.22E-03



10 yr old 3.21E-03 1.81E-04 3.39E-03

Adult 3.29E-03 2.29E-04 3.52E-03

For the cooling tower option, the estimated bounding dose is approximately 5 µSv/y,

representing 0.5 percent of the dose limit of 1 mSv/y. The critical group who will be exposed to such a dose is a one year old infant in the dairy farm group. As for the once-through cooling option, airborne emission was estimated to be the major contributor to the public dose for the cooling tower option.

It should be noted that the value of home range used to derive the average dilution factor for that area is very conservative. The sport fish such as salmon, lake trout, rainbow trout, brown trout, and walleye are so highly mobile, migrate about the entire Lake and are unlikely to be strongly attracted to any feature of the lake

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bottom diffuser. This conservative assumption could result in the overestimate of the dose to all groups who consumed fish caught locally.

3.4.4 Sensitivity Analysis

Sensitivity analyses were conducted for all three designs to account for uncertainty in public dose associated with some condition changes in the future. The factors considered in the sensitivity analysis include:

• Temperature change

• Hydrological condition (change of Lake Ontario current speed)

• Changes in human habits

The annual dose due to airborne emission is estimated to be 4.9E-3 mSv for both the once through cooling option and cooling tower option. However, the annual dose due to waterborne emission is higher for the cooling tower option (3.11E-6 mSv) than the once through cooling option (3.48E-7 mSv). Overall, the doses from the cooling tower option are slightly higher than from the once through cooling option. Therefore for the sensitivity study, the cooling tower option was used for all cases as this is the most conservative option. The results of the sensitivity analyses are summarized in the following paragraphs. It should be noted that for each of three factors considered, the discussion will focus on the critical group who is expected to receive the highest total dose, taking into account all three reactor designs.

Temperature change

To investigate the effect of temperature change on the public dose, it was assumed that the ambient air temperature would increase to 25ºC from 20ºC in the future. The estimated bounding doses are summarized in Table 3.4.4-1.

As shown in the table, when the temperature increases, the dose resulting from airborne emission will increase from 4.90E-3 mSv/y to 5.84E-3 mSv/y. This is expected because the elevated ambient air temperature will result in the decrease of the height of the air plume due to the decreased buoyancy of the plume; correspondingly, the contaminant transfer factor (from the source to receptor) will increase. However, the results show that the total dose will continue to represent a small fraction of the regulatory limit of 1 mSv/y.

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Table 3.4.4-1: Effect of temperature change on public dose

Temperature Dose due to airborne

emission (mSv/y)

Dose due to waterborne

emission (mSv/y)

Total doses (mSv/y)

20ºC 4.90E-03 3.11E-06 4.90E-03

25 ºC 5.84E-03 3.11E-06 5.84E-03

Flow rate

The current flow rates in Lake Ontario could be changed in the future. The effect of current velocity changes on public dose was investigated. For the sensitivity analysis, it was assumed that the flow rate would increase by 100% (i.e.: the velocity of flow to west and east was assumed to increase to 0.2 m/s and 0.32 m/s respectively, compared to 0.1 m/s and 0.16 m/s used in the previous analysis). The estimated bounding doses (the corresponding critical group is adult in fisher group) are summarized in Table 3.4.4-2.

Table 3.4.4-2: Effect of flow rate change on public dose

Current flow rate Dose due to

airborne emission (mSv/y)

Dose due to

waterborne emission (mSv/y)

Total doses

(mSv/y)

Current flow rate of

0.1 m/s and 0.16 m/s 4.05E-05 4.73E-03 4.77E-03

Current flow rate of 0.2 m/s and 0.32 m/s 4.05E-05 5.28E-03 5.32E-03

As shown in the table, the dose to the critical group (adult in fisher group) due to waterborne emission would increase should the flow rate increase. This is corresponding to the decreased dilution factor. However, the total dose is still well below the regulatory limit of 1 mSv/y.

Hypothetical groups

Two “hypothetical groups” were designed to ensure that conservative estimates of exposure are derived to provide an upper bound value of doses incorporating feasible changes in potential group location and habits over the next 100 years. The hypothetical groups included one dairy farm group and one urban resident group. It was assumed that both hypothetical groups lived in the vicinity of the West East Beach. This location was selected as the closest to the point of emission beyond St Marys quarry site within the eastern wind sector on the basis of assessment

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summarized in Section 3.3. It has been estimated that doses to the public will be higher in the eastern wind sector due to predominant wind direction. Although no dairy farming or urban resident group is located in this area at present and there are no existing plans that would result in such changes, this sensitivity test was designed to ensure that unknown future changes are accounted for. The two hypothetical groups were assumed to reside at this location 24 hours a day, 365 days a year. Their intake of local food and water were consistent with the results of site specific survey for this population type thus representing realistic consumption habits. The maximum doses for this assessment for the most exposed age class, taking into account all three reactor designs, are presented in Table 3.4.4-3.

Table 3.4.4-3: Doses to hypothetical groups

Potential critical

group

Age class Dose due to

airborne emission

(mSv/y)

Dose due to

waterborne emission

(mSv/y)

Total doses

(mSv/y)

Hypothetical urban resident 1 yr old 2.96E-03 1.01E-03 3.97E-03

Hypothetical dairy farm 1 yr old 1.76E-02 3.11E-06 1.76E-02

From the table, the highest dose is less than 0.02 mSv/y received by a one year old infant in the hypothetical dairy farm group. This accounts for approximately 2 % of the dose limit of 1 mSv/y. It should be noted that this dose represents a very conservative bounding estimate which is based on the assumption of having a dairy farming community in the maximum exposure location at some point in the future. It has thus been demonstrated that even under this artificial scenario individual doses to the public will be below the limit.

3.5 Radioactive Effluent Dose Consequence for Refurbishment and Decommissioning

Decommissioning of a nuclear facility involves the following broad phases:

� System drying and decontaminate

� Isolate for lay–up state

� Disassemble and remove waste

� Monitor for radiological impact, decontaminate and release to the environment in the end state as per regulatory acceptance criteria

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Decommissioning activities can have radiological safety implications for occupational radiation exposures, environmental emissions, waste generation and disposal resulting in public dose. The extent of the radiological safety implications during decommissioning depends upon the specific decommissioning phase, the facility design, operating history, and magnitude and characteristics of the radiological source term when facility was shut down at the end of plant life. In addition, it greatly depends upon the technology employed to disassemble, decontaminate and remove radiological components.

Generally, the radiological source term in a nuclear power plant during the decommissioning phase or refurbishment will be significantly less compared to operating plant conditions as the spent fuel has been removed for interim storage and no new source term is being produced resulting from fission of nuclear fuel in the reactor. Noble Gases, C-14, Radio-Iodine’s, Tritium, Fission and Activation products are all subject to decay in the period between reactor shut-down and the time when a particular decommissioning or refurbishment phase takes place.

By the end of the dismantling and site restoration phases, the site will be free of industrial hazards. All radioactive contamination in excess of the established levels and other hazardous material will be removed from the site. The site will meet the criteria established by the CNSC according to the license requirements.

A nuclear power plant design includes a planned phase in the operating life of the plant to undergo a major rehabilitation of the plant components and structures to extend plant’s operating life to enhance equipment reliability and safety.

3.5.1 Radiological Implications from Decommissioning and Refurbishment

There are two radiological safety areas that are impacted during any decommissioning and rehabilitation of a nuclear power plant.

• Public radiation exposure from effluent emissions to air and water • Individual and occupational radiation exposures from in plant activities

The information presented below covers only planned operations as per the facility design and does not address implications for accidental conditions.

The decommissioning activities are carried out at the end of the plant’s useful life and after a long period (about 50 years) to allow for radioactive decay of the source term to reduce the radiological hazards. The time selected is such that all short term radionuclides decay away leaving only long lived radionuclides. Nuclear power plant refurbishment takes place after a much shorter shut-down period than in the case of decommissioning, which nevertheless results in the decay of some short-lived radionuclides.

The table below provides a comparative analysis of radionuclide inventory during normal operations, refurbishment and decommissioning. Also, the table compares the absence of pathways during decommissioning and rehabilitation activities. Most of the

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pathways will cease to exist and overall radiological emissions will be a small fraction compared to when plant was operating. Generally, this is true for all reactor designs.

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Table 3.5-1: Comparative Analysis of Radionuclide Inventory During Refurbishment and Decommissioning

Facility Life Cycle Phases and Environmental Impact/Emission Radionuclide category

Plant

Operating

Prolonged Shut Down for

Refurbishment

Decommissioning

Activities

Tritium Yes Yes (Much reduced depending

upon the plant type and

system performance)

Significantly reduced

due to decay and

system dryness

Noble Gases Yes None None

Radio-Iodine Yes None* None*

C-14 (gaseous) Yes Relatively a small fraction None

Airborne Particulates Yes Yes Yes

Gamma exposure Yes Reduced level due to shutdown

conditions

Much reduced due to

decay

* Long lived isotopes are present but inventories are insignificant.

3.5.2 Source Term and Dose Assessment for Refurbishment and Decommissioning Operations

From Table 3.5-1 it can be seen that both Refurbishment and Decommissioning are associated with lower inventory than reactor operations. However, both refurbishment and decommissioning may involve intrusion into the reactor core thus making normally isolated contaminated components accessible as potential sources for environmental releases. ACR design presents the bounding case for refurbishment as the scope of refurbishment for PWR designs is likely to be limited to replacement of steam generators. While the design life of the PWR steam generators is 60 years, their replacement during plant lifetime cannot be completely ruled out. Steam generator replacement is a well developed process with contamination well contained within the vessels thus giving rise to only minimal releases.

Refurbishment of ACR reactors is assumed to be consistent in scope to that carried out in the past for CANDU plants. Other potential refurbishment work might involve replacement of contaminated components within the reactor vault. Environmental releases are largely dependent on the processes implemented for contamination control, decontamination and effluent processing. These have been developed significantly since the first refurbishment at Pickering NGS took place in 1980s and early 1990s and for new ACR refurbishments the resulting releases are likely to be significantly smaller than in the case of Pickering.

Refurbishment of Units 3 and 4 at Pickering A Nuclear Generating Station took place between June 1989 (shut down Unit 3) and August 9, 1991 (return to service) and between August 15, 1991 (shut down Unit 4) and March 28, 1993 (return to service).

Based on comparative analysis of environmental releases for the period between 1989 and 1993 with post-refurbishment releases (Table 3.5-2), it can be concluded that

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releases to air during refurbishment at Pickering A were 2-3 times higher than during normal operations. Metals, copper and zinc, were being eroded from the condenser tubes made of brass alloy, which, which resulted in abnormally large releases of metals to water from the Pickering Nuclear Generating Station. This abnormal event was independent of refurbishment activities. However Pickering discharges for years 1990 through 1993 provided in Table 3.5-2 include releases from operational units as well as those that were refurbished during this period. From Table 3.5-1 it can be seen that all radio-iodines and noble gas releases were due to operations of the remaining units and only H-3, particulate and C-14 releases had contribution from the units that were being refurbished.

Decommissioning operations leading to releases of radioactive substances will be consistent in scope to a comprehensive refurbishment project involving dismantling and removal of core components such as that carried out at Pickering A. However decommissioning operations involving handling of radioactive materials will be carried out after a significant shut-down resulting in substantial decay of short-lived radionuclides and smaller releases to environment compared to refurbishment.

In summary, a bounding assumption can be made that environmental releases during refurbishment will be increased by three and twenty times for releases to air and water respectively for H-3, particulates and C-14, although the actual refurbishment and decommissioning releases – taking into account modern technologies for minimizing releases – are likely to be substantially lower. The release of noble gases and radioiodines will be negligible. Under normal operations noble gases and radioiodines were estimated to contribute about 80% of the total dose to the critical group (for the bounding release scenario). Assuming two units are being refurbished and two units are in operation, the resulting effective dose during the refurbishment

can be estimated as 4 µSv/y. This is based on the 5 µSv/y bounding dose for the once-through cooling option for normal operation of four units, lack of noble gas and radioiodine releases and a bounding assumption that releases of the remaining radionuclides will increase by a factor of 3 and 20 for airborne and waterborne discharges respectively.

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Table 3.5-2: Pickering A Releases to Air During Refurbishment of Units 3 and 4 (Bq)

Radionuclide

19901 1991*2,3,4 19925 19936 Average for 1990-

1993

Current Annual

Emissions for PNGS

A7

H-3 6.29E+14 4.83E+14 5.92E+14 5.18E+14 5.56E+14 2.35E+14

I-131 3.22E+08 1.13E+08 8.88E+07 1.33E+08 1.64E+08 6.13E+07

Particulates 2.85E+08 8.88E+07 8.88E+07 8.51E+07 1.37E+08 1.15E+08

Noble Gases, Bq-MeV

4.07E+14 4.88E+14 3.26E+14 3.70E+14 3.98E+14 1.18E+14

C-14 2.92E+12 3.80E+12 2.07E+12 1.55E+12 2.59E+12 2.13E+12

*No information available for the fourth quarter; annual release for 1991 was estimated based on

discharge data for January through September 1991. 1OPG 1991a 2OPG 1991b 3OPG 1991c 4OPG 1991d 5OPG 1993 6OPG 1994 7OPG 2007

Table 3.5-3: Pickering A Releases to Water During Refurbishment of Units 3 and 4 (Bq)

Radionuclide

19901 1991*2,3,4 19925 19936 Average

for 1990-1993

Current

Annual Emissions

for PNGS A7

Tritium Oxide 4.07E+14 3.90E+14 3.03E+15 5.18E+14 1.09E+15 8.20E+13

Gross

Beta-Gamma

4.81E+10 4.79E+10 4.81E+10 3.48E+10 4.47E+10 2.27E+09

*No information available for the fourth quarter; annual release for 1991 was estimated based

on discharge data for January through September 1991. 1OPG 1991a 2OPG 1991b 3OPG 1991c 4OPG 1991d 5OPG 1993 6OPG 1994 7OPG 2007

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3.6 Assessment of Consequences from Radioactive Waste and Used Fuel Management

The radioactive waste and used fuel management systems include

- Liquid waste treatment facility, which may include heavy water processing plant;

- Solid waste conditioning and storage facility; - Used fuel storage facility.

Operation of these facilities will result in airborne emissions to atmosphere and liquid releases to Lake Ontario. Radioactive releases to air will mainly comprise tritium, C-14 and noble gases. They will be processed through dehumidifiers, activated charcoal delay beds and filters before release. Radioactive releases to water will be mainly comprised of tritium, Carbon 14 and other beta-gamma emitters. All effluents will be collected in holding tanks and released following treatment in facilities typically including filtration, reverse osmosis and sorption. All releases to water will be monitored prior to discharge. Based on studies conducted by OPG [OPG 2009b], a preliminary conclusion can be reached that releases from radioactive waste and used fuel management systems represent a small fraction of releases due to normal operations and that accident consequences from operating such systems are negligible compared to severe accidents considered for nuclear power plant operation.

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4.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR ACCIDENTAL DISCHARGES


Accidental discharges can result from internally or externally initiated events. Externally initiated man made events are addressed in [AMEC NSS 2009d] and externally initiated natural events are addressed in [AMEC NSS 2009a; AMEC NSS, 2009b; AMEC NSS, 2009c].

The criteria for assessing and addressing the risk from accidental radiological releases for new Nuclear build projects are provided in RD-337 [CNSC 2008b].

As stated in Section 1.0, RD-337 sets Qualitative and Quantitative Safety Goals, in particular:

Small Release Frequency (SRF)

The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1015 Becquerel of iodine-131 (I-131) is less than 10-5 per reactor year. A greater release may require temporary evacuation of the local population.

Large Release Frequency (LRF)

The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1014 Becquerel of caesium-137 (Cs-137) is less than 10-6 per reactor year. A greater release may require long term relocation of the local population.

RD-337 establishes requirements regarding containment performance following a severe accident to the effect that “Containment maintains its role as a leak-tight barrier for a period that allows sufficient time for the implementation of off-site emergency procedures following the onset of core damage. Containment also prevents uncontrolled releases of radioactivity after this period.”

In order to satisfactorily demonstrate that the proposed OPG New Nuclear at Darlington will comply with RD-337, (with respect to the impact of protective measures on the public), while also demonstrating compliance with the emergency response requirements of RD-346 it is necessary to develop an accident release scenario model to provide results that can be used to assess the impacts of an accidental release.

4.1 Derivation of Accident Scenario and Release Characteristics

Final versions of the Safety Analysis and PRAs have not yet been completed for the reactors taking into consideration Canadian regulatory requirements and specific site characteristics. The EA TSD on Malfunctions and Accidents [SENES 2009c] provides details on the design features in the new reactors which will reduce the likelihood and

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minimize the consequences of any accidents. An assessment was performed to evaluate scenarios corresponding to the RD-337 safety goal release thresholds to assess the potential impact of such release in terms of radiation dose to the public. These are not design basis cases and are intended to be indicative only, to show that the intent of RD-337 with respect to the impact of protective measures on the public can be met. For this, RD-337 Safety Goal Based (RSGB) Releases to the environment were derived and used in dose calculations.

The new reactors must comply with the RD-337 safety goals. This sets limits on the performance of the reactors with respect to accident frequency and consequences of off-site releases. The release limits associated with the LRF and SRF goals can be considered to be the maximum that would be expected for a single event of a 10-6 per year or 10-5 per year frequency respectively, for a plant that just meets the RD-337 requirements.

A stylized accident radioactive release scenario was created, with a representative relative isotopic abundance based on actual reactor core behaviour. The release to the environment (also referred to as source term) was normalized to each of the SRF and the LRF threshold release values for I-131 and Cs-137, respectively. These releases, the RSGB Small Release and the RSGB Large Release respectively, were used to determine the potential dose to the public by postulating an event in these categories corresponding to the Large and Small Release thresholds. For the purposes of this analysis, the containment leak-tight period required by RD-337 has been represented as a delay in release of 24 hours after reactor shutdown.

The process by which the detailed source term used in the dose calculations was derived is described in the following sections.

4.2 Representative Source Term for Radioactive Airborne Releases

For the purposes of modeling an airborne radiological release to the public, the source term for each event is obtained from a combination of core inventory (CI) and a set of release fractions (RF).

Source = CI x RF

The RF is normally applied to a group of chemically and physically similar isotopes. For example, group 6, which is named for Ruthenium, includes not only Ruthenium isotopes, but also Rhodium isotopes.

The RSGB cases are derived on the following basis:

� The reactor with the highest burn up rate will produce the most conservative starting isotope mix for the core inventory.

� The release fractions are normalised to generate a release that meets the threshold requirements of both the RD-337 SRF and LRF releases to provide the RSGB cases.

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� Physical and chemical processes that would normally lead to retention in containment during the assumed 24-hour leaktight period and thereby reduce the quantities available for subsequent release are neglected.

It should be noted that the above basis does not directly reflect the reactor designs nor represent a realistic accident event, since the containment designs will provide for mitigation of the release (this has not been assumed for the RSGB cases). No credit is therefore taken for mitigation in this assessment, other than the initial 24 hour delay period prior to release.

Based on the above, the BDBA scenario on which to base the isotopic mix for the RSGB cases for OPG New Nuclear at Darlington has been selected on the basis that the event is the largest contributor to the Large Release Frequency Accident (as defined in [NRC 2007]) (contributing 66%). The calculated frequency of this event is well below the RD-337 LRF limit.

4.2.1 Core Inventory

The EPR core inventory corresponds to the largest physical core inventory (i.e., it experiences highest burn-up rate and uses the highest enrichment of the three proposed designs at 100% power). The 60 dose-significant isotopes in the core inventory, as used in the assessment, are presented in Table 4.2-1.

The remaining isotopes are not considered in the assessment because they do not contribute to dose for one or more of the following reasons:

� Very short half life;

� Extremely long half life, in effect, stable;

� Very small quantity.

Table 4.2-1 : EPR Core Inventory From Vendor Data (AREVA 2007)

Core Inv.

(Bq) Radionuclide Parent

MACCS27

Isotope

Group

Half-Life (s)

EPR

Co-58 None 6 6.12E+06 1.00E+00*

Co-60 None 6 1.66E+08 1.00E+00*

Kr-85 None 1 3.39E+08 7.77E+16

Kr-85m None 1 1.61E+04 1.67E+18

Kr-87 None 1 4.56E+03 3.34E+18

Kr-88 None 1 1.01E+04 4.74E+18

Rb-86 None 3 1.61E+06 2.15E+16

Sr-89 None 5 4.49E+06 5.96E+18

7 Information on MACCS2 accident consequence assessment package is provided in Section 4.4.2.

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Core Inv.


MACCS27

Isotope

Group

Half-Life (s)

EPR

Sr-90 None 5 8.87E+08 6.25E+17

Sr-91 None 5 3.41E+04 7.66E+18

Sr-92 None 5 9.76E+03 7.92E+18

Y-90 Sr-90 7 2.31E+05 6.62E+17

Y-91 Sr-91 7 5.08E+06 7.25E+18

Y-92 Sr-92 7 1.27E+04 8.66E+18

Y-93 None 7 3.64E+04 7.92E+18

Zr-95 None 7 5.66E+06 8.47E+18

Zr-97 None 7 6.05E+04 8.99E+18

Nb-95 Zr-95 7 3.03E+06 8.47E+18

Mo-99 None 6 2.38E+05 9.58E+18

Tc-99m Mo-99 6 2.17E+04 8.40E+18

Ru-103 None 6 3.42E+06 8.95E+18

Ru-105 None 6 1.60E+04 7.25E+18

Ru-106 None 6 3.19E+07 5.29E+18

Rh-105 Ru-105 6 1.28E+05 6.48E+18

Sb-127 None 4 3.28E+05 6.66E+17

Sb-129 None 4 1.56E+04 1.79E+18

Te-127 Sb-127 4 3.37E+04 6.62E+17

Te-127m None 4 9.42E+06 8.99E+16

Te-129 Sb-129 4 4.20E+03 1.77E+18

Te-129m None 4 2.89E+06 2.62E+17

Te-131m None 4 1.08E+05 7.55E+17

Te-132 None 4 2.81E+05 7.33E+18

I-131 Te-131m 2 6.95E+05 5.14E+18

I-132 Te-132 2 8.23E+03 7.44E+18

I-133 None 2 7.49E+04 1.07E+19

I-134 None 2 3.16E+03 1.18E+19

I-135 None 2 2.37E+04 9.95E+18

Xe-133 I-133 1 4.57E+05 1.07E+19

Xe-135 I-135 1 3.30E+04 3.43E+18

Cs-134 None 3 6.50E+07 2.40E+18

Cs-136 None 3 1.12E+06 5.96E+17

Cs-137 None 3 9.50E+08 9.14E+17

Ba-139 None 9 4.99E+03 9.69E+18

Ba-140 None 9 1.11E+06 9.32E+18

La-140 Ba-140 7 1.45E+05 9.40E+18

La-141 None 7 1.42E+04 8.92E+18

La-142 None 7 5.72E+03 8.70E+18

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Core Inv.


MACCS27

Isotope

Group

Half-Life (s)

EPR

Ce-141 La-141 8 2.81E+06 8.29E+18

Ce-143 None 8 1.19E+05 8.44E+18

Ce-144 None 8 2.46E+07 6.29E+18

Pr-143 Ce-143 7 1.17E+06 8.36E+18

Nd-147 None 7 9.50E+05 3.49E+18

Np-239 None 8 2.03E+05 1.41E+20

Pu-238 Cm-242 8 2.77E+09 5.40E+16

Pu-239 None 8 7.59E+11 2.27E+15

Pu-240 Cm-244 8 2.06E+11 5.18E+15

Pu-241 None 8 4.54E+08 9.36E+17

Am-241 None 7 1.36E+10 1.07E+15

Cm-242 None 7 1.41E+07 4.85E+17

Cm-244 None 7 5.71E+08 2.57E+17 * Although not present in the source term, a value of 1 is assigned to allow the code to include build-up of Cobalt during the decay chain process.

4.2.2 Derivation of Release Fractions for RSGB Releases

RD-337 stipulates the threshold releases to the environment for SRF and LRF events and that containment should prevent releases to the environment, even in the case of a severe accident, for a period of time allowing protective actions to be implemented. This time is taken to be 24 hours. Radioactive material would therefore decay before the release to the environment. This will affect the isotope mix for the RSGB SRF and LRF due to decay during the containment hold-up period.

The effect of this can be adjusted for by reverse decay of the I-131 or Cs-137 quantity (for the SRF and LRF cases respectively) and then normalising the release to the pre-decay quantities of I-131 or Cs-137, as appropriate, to give a release to containment. These will then decay before release to the environment. The quantities (released into containment) for normalisation are shown below:

Case Normalising Isotope

RD-337 Quantity Released to Environment

Starting Quantity Pre-Decay (24 hrs)

Released to Environment

SRF I-131 1.0x1015 Bq 1.09x1015 Bq

LRF Cs-137 1.0x1014 Bq 1.00x1014 Bq

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The baseline release fractions for the chosen event are presented in Table 4.2-2 by MACCS2 fission group. Table 4.2-2 also presents the total baseline activity released, for all isotopes in the group for the BDBA event.

The baseline release fractions have then been normalised to the pre-decay8 quantities of 1.09x 1015 Bq I-131 (1) and 1.00x 1014 Bq Cs-137 (2) for the SRF and LRF cases respectively, as follows:

RFiNI-131 = Release Fraction for Isotope Group i x (1.09 x1015) (1) Release Fraction for I-131 x Core Inventory of I-131

RFiNCs-137 = Release Fraction for Isotope Group i x (1.0x1014) (2) Release Fraction for Cs-137 x Core Inventory of Cs-137

Where:

RFiNI-131 = Normalised Release Fraction for isotope i, normalised to the SRF pre-decay quantity

RFiNCs-137 = Normalised Release Fraction for isotope i, normalised to the LRF pre-decay quantity

The above equations were used to normalise the baseline release to give the isotope mix release fractions for an RSGB SRF case and an RSGB LRF case release.

The resulting RFs and release quantities are presented in Table 4.2-3. Note that the quantities released (in Bq) shown in Table 4.2-3 are for the group, thus Group 2 includes not only the quantity of I-131, but all the iodine isotopes, and hence the value is greater than 1.00x 1015 Bq.

Table 4.2-2 : Baseline Release

Fraction of initial core inventory released to environment as a total for each MACCS2 fission group

Group 1

(Xe/Kr)

Group 2

(I)

Group 3

(Cs)

Group 4

(Te)

Group 5

(Sr)

Group 6

(Ru)

Group 7

(La)

Group 8

(Ce)

Group 9

(Ba)

9.90E-01 4.10E-02 2.60E-02 2.70E-01 4.80E-04 7.20E-03 7.60E-05 2.20E-04 7.20E-03

Quantities released to environment in Becquerels

Group 1 (Xe/Kr)

Group 2 (I)

Group 3 (Cs)

Group 4 (Te)

Group 5 (Sr)

Group 6 (Ru)

Group 7 (La)

Group 8 (Ce)

Group 9 (Ba)

2.4E+19 1.8E+18 1.0E+17 3.6E+18 1.1E+16 3.3E+17 6.8E+15 3.6E+16 1.4E+17

8 The normalisation and reverse decay method is a simplification, however for the generic source term, the methodology is conservative for the SRF because it artificially increases the starting quantities of such short lived

isotopes as Tellurium and I-132. For the LRF it has a negligible impact because the half life of Cs-137 is so long that the applied factor is extremely close to unity.

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Table 4.2-3 : RSGB Releases Normalised against SRF and LRF

Fraction of initial core inventory released as a total for each MACCS2 fission group Case

Group 1

(Xe/Kr)

Group 2

(I)

Group 3

(Cs)

Group 4

(Te)

Group 5

(Sr)

Group 6

(Ru)

Group 7

(La)

Group 8

(Ce)

Group 9

(Ba)

SRF Case 5.12E-03 2.12E-04 1.34E-04 1.40E-03 2.48E-06 3.72E-05 3.93E-07 1.14E-06 3.72E-05

LRF Case 4.17E-03 1.73E-04 1.09E-04 1.14E-03 2.02E-06 3.03E-05 3.20E-07 9.26E-07 3.03E-05

Quantities Released From Containment In Becquerels Case

Group 1 (Xe/Kr)

Group 2 (I)

Group 3 (Cs)

Group 4 (Te)

Group 5 (Sr)

Group 6 (Ru)

Group 7 (La)

Group 8 (Ce)

Group 9 (Ba)

SRF Case 1.2E+17 9.5E+15 5.3E+14 1.9E+16 5.5E+13 1.7E+15 3.5E+13 1.9E+14 7.1E+14

LRF Case 1.0E+17 7.8E+15 4.3E+14 1.5E+16 4.5E+13 1.4E+15 2.9E+13 1.5E+14 5.8E+14

The RSGB Releases were modeled as a continuous plume, delayed for 24 hours after reactor shutdown, with a duration of 72 hours. This provides a representative scenario for evaluation of potential doses consistent with RD-337 requirements but is not necessarily characteristic of all potential releases for the reactor technologies.

4.3 Source Term for Accidental Radioactive Discharges to Surface Water and Groundwater

Accidental waterborne releases from Darlington Nuclear Site to the environment could originate from the following pathways:

• Accidental surface water discharge • Atmospheric fallout from a severe reactor accident • Accidental releases to groundwater

For the OPG New Nuclear at Darlington, it is assumed that surface water runoff from the nuclear power plant buildings will be collected in the storm water management ponds and then discharged to an existing drainage course or Lake Ontario. A storm water management system will be installed [CH2MHill 2009b] to:

1. Control runoff release rates to levels that may be efficiently accommodated by the management system and the receiving water body; and

2. Store and detain site runoff to ensure discharges meet applicable regulatory water quantity and quality objectives;

in line with the Ontario Stormwater Management Planning and Design Manual 2003 [Ontario 2003]. Therefore, the only direct releases to water during accidents would be from airborne releases, and are represented by the impact from fallout to Lake Ontario

AMEC NSS Limited


following a severe accident involving the OPG New Nuclear at Darlington as assessed in Section 4.4, which includes ingestion doses from both food and water.

In the event of an accident, radioactivity can be accidentally released to the groundwater. This can be mitigated through standard engineering practices for detecting and containing leaks to meet requirements. In addition to any engineered systems, there will be a local lowering of the groundwater table in the vicinity of the nuclear power plant building. This will be due to reduced recharge as well as abstraction of water. Therefore groundwater gradients will be towards the location of the OPG New Nuclear so that any contamination can be safely extracted and treated.

4.4 Offsite Public Dose Consequence for Radioactive Airborne Release Accidents

4.4.1 Dose Targets and Limits

For the Early or Emergency Phase of an event, Protective Action Levels (PALs) have been specified by the Province of Ontario Nuclear Emergency Plan [Ontario 1999], as shown in Table 4.4-1 supported by Health Canada [Health Canada 2003] who have recommended intervention levels, as shown in Table 4.4-2.

Table 4.4-1: Protective Action Levels

[Ontario 1999]

Exposure Control Measures*

Lower Level Upper Level Protective

Measure Effective

Dose Thyroid Dose

Effective

Dose Thyroid Dose

Sheltering 1 mSv

(0.1 rem)

10 mSv

(1 rem)

10 mSv

(1 rem)

100 mSv

(10 rem)

Evacuation 10 mSv

(1 rem)

100 mSv

(10 rem)

100 mSv

(10 rem)

1 Sv

(100 rem)

Thyroid Blocking - 100 mSv

(10 rem) -

1 Sv

(100 rem)

* Defined in Section 5.3

AMEC NSS Limited


Table 4.4-2: Health Canada Recommended Intervention Levels

[Health Canada 2003]

Countermeasure Intervention Level (averted dose)

Sheltering 5 mSv in 1 day

Evacuation 50 mSv in 7 days

Relocation 50 mSv in 1 year (return when <50 mSv in a year and < 10 mSv in 1 month)

Stable Iodine Prophylaxis 100 mSv to thyroid

Food Controls 1 mSv from each of 3 food groups

For the Late phase of the event, where relocation may be required, [Ontario 1999] sets out dose limits of no more than 20 mSv in 1 year for the late dose. Health Canada have a less stringent 50 mSv dose in 1 year limit for relocation, as indicated in Table 4.4-2. Care should be taken when considering results against these latter values due to the varying limits for differing time periods.

The above tables form the limits for assessing the acceptability of risk from the proposed station and the extent required for evaluation of the emergency response plan.

4.4.2 Dose Consequence Calculations

The MACCS2 (MELCOR accident consequence code system) computer code [SANDIA 1990, SANDIA 1997 and SANDIA 1998] is used to estimate the off-site radiological doses and health effects that could result from postulated accidental releases of radioactive materials to the atmosphere. The most recent validated version (Version 1.13.1.0) of the computer code known as MELCOR accident consequence code system (MACCS2) was used to assess the off-site consequences of each of the stylized releases. The development of MACCS2 is sponsored by the U.S. Nuclear Regulatory Commission and is carried out by Sandia National Laboratories in conformance with its standard software development procedures [SANDIA 1998, US DOE 2004]. MACCS2 is used for the current OPG Probabilistic Risk Assessments. Therefore, the use of MACCS2 in this assessment enables reasonable comparison of results between the proposed plant and existing OPG stations.

MACCS2 models the off-site consequences of a reactor accident that releases plumes of radioactive materials to the atmosphere. After the release, the radioactive gases and aerosols in the plume, while dispersing in the atmosphere, are transported by the prevailing wind. The effects of building wake and release height are taken into account. Radioactive decay and build-up of daughter products are modeled throughout the process and the effects of meteorological conditions, including frequency of such conditions are also modelled.

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Radioactive materials deposited from the plume can contaminate the environment and the population can be exposed to radiation through a number of pathways.

In MACCS2, the time period after the accident is divided into three phases: Early phase, Intermediate and long-term phase. The Late dose is derived from the sum of the Intermediate and Long-term phases.

The Early phase begins immediately upon the arrival of the first plume and can last up to seven days. MACCS calculates the Committed Effective Dose (CED) with distance for varying periods of residence. The dose is calculated for a member of the most critical group. Along with the dose by distance, the dose by pathway (Cloudshine, Groundshine, Inhalation and Resuspension Inhalation) is also calculated for a member of the critical group who is assumed to continuously reside at a distance of 1 km from the release point. This is a conservative approach relative to any population behaviour. Various mitigative measures can be specified for this phase, including evacuation, sheltering and dose-dependent relocation. The dose to thyroid is also calculated.

Following the Early phase, four long-term exposure pathways are modeled in the intermediate and long-term phases: groundshine, resuspension, inhalation and ingestion of contaminated food and ingestion of contaminated drinking water.

The intermediate phase is used to represent a period in which post-accident hazard evaluation can be performed. In this case, since no mitigation is being performed, the intermediate phase is set to zero.

The long-term phase represents the time period subsequent to the intermediate phase and is set to obtain the required late dose results (in this case to one year and to 50 years).

The MACCS2 code calculates the radiological dose that is not avoided by mitigative measures. In this case no mitigative actions are to be modeled, enabling calculation of the potential avoidable dose if immediate evacuation occurred.

The release must be defined in terms of the initial inventory of material and the fraction of that material that is released, along with its physical and chemical properties. Many of the input variables in these files use default values based on the reasoning supplied in the MACCS documentation [SANDIA, 1990, SANDIA, 1998]. The source term is described in Section 4.2 above. Location specific parameters have been developed in the first instance from the Darlington NGS Safety Report [OPG 2004]. The physical station parameters and other assumptions made in the model are briefly summarised below.

Key Assumptions

The mean meteorological conditions were calculated by MACCS based on hourly data collected from the site for the year 2000. The following key assumptions have been made in preparing this assessment:

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� No mitigation (i.e.: sheltering or evacuation) is taken into account.

� The doses are calculated assuming a radial grid with 16 sectors centred on each of the 16 compass points. The dose is calculated in the center of each of the elements of the grid.

� The releases are conservatively assumed to be cold ground level releases.

� The nearest building downstream of the release (for building wake effects etc.) is assumed to be the existing Darlington Nuclear Generating Station.

� The release scenario is a single continuous plume lasting 72 hours, modelled as three consecutive 24 hour plumes in MACCS.

� The CED and the equivalent dose to thyroid have been calculated for the Early RSGB SRF, and the Late RSGB LRF phases. The total CED for the LRF case has also been calculated.

� The immediate surrounding area is assumed to be low density residential and farmland for the purposes of surface roughness.

� The population in each grid element is assumed to be 1. This is adequate for determining individual doses.

The above assumptions coupled with the source term information were used to form the RSGB scenarios.

4.4.3 RSGB Dose Consequence Results

Early (Emergency) Phase Results

Figure 4.4-1 shows the 7-day Early committed effective doses to the whole body for the RSGB SRF release. The results are summarised in Table 4.4-3. This is the phase which dictates emergency response and shows the PALs for effective dose.

Figure 4.4-2 shows the same information for the equivalent dose to the thyroid over the same period, and the PAL values for dose to the thyroid. The results are summarised in Table 4.4-4.

Late Phase Results

The Late results are presented for the RSGB LRF cases.

Figure 4.4-3 shows 1-year and 50-year late mean committed effective doses for a 7-day Early phase, representing the additional dose received (after the first 7 days of the Early phase is complete), over the next 1 year and 50 years respectively, following the event. The results are summarised in Table 4.4-5. In addition, Table 4.4-6 presents the dose by pathway to a receptor at a distance of 1 km from the point of release for the same period.

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Total Event Results

The Total results are presented for the RSGB LRF cases.

Table 4.4-7 presents the 1-year and 50-year mean total committed effective doses for a 7-day Early phase for the RSGB LRF releases. Again, Table 4.4-6 presents the dose by pathway to a receptor at a distance of 1 km for the Total event.

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Table 4.4-3: Variation of Committed Effective Early Whole Body Doses with Distance for RSGB SRF

Distance

from point of

release (km)

RSGB SRF

Early Effective

Whole Body

Dose (mSv)

0.2

2.8

E+

02

0.5

7.5

E+

01

0.7

4.5

E+

01

0.8

5

3.3

E+

01

1

2.5

E+

01

1.5

5

1.3

E+

01

3

4.1

E+

00

6

1.6

E+

00

10

9.7

E-0

1

14

7.4

E-0

1

20

5.8

E-0

1

28

4.7

E-0

1

36

4.2

E-0

1

50

2.9

E-0

1

70

2.2

E-0

1

90

1.8

E-0

1

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Table 4.4-4: Variation of Early Equivalent Dose to the Thyroid with distance for RSGB SRF Release

Distance

from point of

release (km)

RSGB SRF

Equivalent

Dose to the

Thyroid (mSv)

0.2

1.8

E+

03

0.5

4.7

E+

02

0.7

2.8

E+

02

0.8

5

2.0

E+

02

1

1.6

E+

02

1.5

5

7.8

E+

01

3

2.5

E+

01

6

9.8

E+

00

10

5.9

E+

00

14

4.5

E+

00

20

3.5

E+

00

28

2.9

E+

00

36

2.5

E+

00

50

1.7

E+

00

70

1.3

E+

00

90

1.1

E+

00

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Table 4.4-5: Variation of Late Committed effective Whole Body Doses with Distance for RSGB LRF Release

Distance

from point of

release (km)

RSGB LRF 1

Year Late

Whole Body

Dose (mSv)

RSGB LRF 50

Year Late

Whole Body

Dose (mSv)

0.2

1.9

E+

02

6.4

E+

02

0.5

5.0

E+

01

1.7

E+

02

0.7

3.0

E+

01

9.9

E+

01

0.8

5

2.2

E+

01

7.2

E+

01

1

1.7

E+

01

5.5

E+

01

1.5

5

8.2

E+

00

2.7

E+

01

3

2.6

E+

00

8.8

E+

00

6

1.0

E+

00

3.4

E+

00

10

6.0

E-0

1

2.0

E+

00

14

4.5

E-0

1

1.5

E+

00

20

3.4

E-0

1

1.1

E+

00

28

2.6

E-0

1

8.8

E-0

1

36

2.2

E-0

1

7.4

E-0

1

50

1.5

E-0

1

5.1

E-0

1

70

1.1

E-0

1

3.5

E-0

1

90

8.3

E-0

2

2.8

E-0

1

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RD

337 S

RF

3 d

ay

Rele

ase, 24 h

our

dela

y w

ith 7

day

EA

RLY

Pha

se

0.11

10

100

1000

010

20

30

40

50

60

70

80

90

100

Dis

tan

ce f

rom

rele

ase p

oin

t (k

m)

Dose (mSv)R

D337 S

RF

3 d

ay

Rele

ase, 24 h

our

dela

y w

ith 7

da

y E

AR

LY

Phase

EV

AC

UA

TIO

N O

PT

ION

AL/S

HE

LT

ER

ING

MA

ND

AT

OR

Y P

AL 1

0 m

Sv to 1

00 m

Sv

SH

ELT

ER

ING

OP

TIO

NA

L P

AL 1

mS

v to 1

0 m

Sv

EV

AC

UA

TIO

N M

AN

DA

TO

RY

PA

L G

RE

AT

ER

TH

AN

100m

Sv

Figure 4.4-1: Variation of Committed Effective Dose with Distance for RSGB SRF Release – EARLY 7 Day

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Table 4.4-6: Dose (mSv) by Pathway at 1 km from Release Point for RSGB LRF

LATE DOSESi

Committed Effective Dose by Pathway

Thyroid Dose by Pathway

Case

Groundshine

Resuspension

Inhalation

Totalii

Groundshine

Resuspension

Inhalation

Totalii

RD

-337 R

SG

B L

RF L

ate

Dose

(

1 y

ear)

15

2.2

17

15

0.8

2

16

RD

-337 R

SG

B L

RF L

ate

Dose

(

50 y

ears

) 53

2.8

55

56

0.8

9

57

TOTAL DOSESi

Committed Effective Dose by Pathway

Thyroid Dose by Pathway

Case

Groundshine

Resuspension

Inhalation

Remainingiii

Pathways

Total

Groundshine

Resuspension

Inhalation

Remainingiii

Pathways

Total

RD

-337 R

SG

B L

RF T

ota

l D

ose

(1 y

ear

Late

Phase

) 20

2.7

14

37

22

4.3

118

140

RD

-337 R

SG

B L

RF T

ota

l D

ose

(50 y

ear

Late

Phase

) 58

3.3

14

76

62

4.3

118

180

Note

s:

i. D

ose

s are

norm

ally

rounded t

o 2

sig

nific

ant

figure

s; t

here

fore

sum

of

path

ways

may n

ot

exact

ly e

qual “T

ota

l”.

ii. I

ncl

udes

ingest

ion d

ose

, although t

he c

ontr

ibution is

very

sm

all.

iii. Rem

ain

ing p

ath

ways

incl

ude C

loudsh

ine, In

hala

tion a

nd I

ngest

ion d

ose

s.

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1

10

100

1000

100

00

01

020

30

40

50

60

70

80

90

100

Dis

tan

ce f

rom

rele

ase p

oin

t (k

m)

Dose (mSv)

RD

33

7 S

RF

3 d

ay

Rele

ase, 24

hour

dela

y w

ith 7

day

EA

RLY

Phase

SH

ELT

ER

ING

OP

TIO

NA

L P

AL 1

0m

Sv to 1

00 m

Sv T

HY

RO

ID D

OS

E

EV

AC

UA

TIO

N O

PT

ION

AL/S

HE

LT

ER

ING

MA

ND

AT

OR

Y P

AL 1

00 m

Sv to 1

Sv T

HY

RO

ID D

OS

E

EV

AC

UA

TIO

N M

AN

DA

TO

RY

PA

L G

RE

AT

ER

TH

AN

1S

v T

HY

RO

ID D

OS

E

Figure 4.4-2: Variation of Dose to Thyroid with Distance for the RSBG SRF Release – EARLY 7 Day

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0.0

1

0.11

10

100

1000

010

20

30

40

50

60

70

80

90

100

Dis

tan

ce f

rom

rele

ase p

oin

t (k

m)

Dose (mSv)

RD

337 L

RF

3 d

ay

Rele

ase, 24 h

our

dela

y w

ith 7

day

EA

RLY

Phase a

nd 1

year

late

phase

RD

337 L

RF

3 d

ay

Rele

ase, 24 h

our

dela

y w

ith 7

day

EA

RLY

Phase a

nd 5

0 y

ear

late

phase

Figure 4.4-3: Variation of Committed Effective Dose with Distance for RSGB LRF Release - LATE

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Table 4.4-7: Total Event Committed Effective Doses for RSGB LRF

Distance

from point of

release (km)

RSGB LRF

Case (mSv)

TOTAL 1 Year

RSGB LRF Case

(mSv)

TOTAL 50 Years

0.2

4.2

E+

02

8.7

E+

02

0.5

1.1

E+

02

2.3

E+

02

0.7

6.6

E+

01

1.4

E+

02

0.8

5

4.8

E+

01

9.9

E+

01

1

3.7

E+

01

7.6

E+

01

1.5

5

1.9

E+

01

3.8

E+

01

3

6.0

E+

00

1.2

E+

01

6

2.3

E+

00

4.7

E+

00

10

1.4

E+

00

2.8

E+

00

14

1.1

E+

00

2.1

E+

00

20

8.1

E-0

1

1.6

E+

00

28

6.5

E-0

1

1.3

E+

00

36

5.6

E-0

1

1.1

E+

00

50

3.9

E-0

1

7.4

E-0

1

70

2.8

E-0

1

5.3

E-0

1

90

2.3

E-0

1

4.3

E-0

1

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4.5 Offsite Public Dose Consequence for Radioactive Waterborne Release Accidents

As discussed in Section 4.3, consequences from airborne release to water will be bounded by consequences from severe accidents scenarios assessed in Section 4.4.

4.6 Impact of Mitigation on Dose Consequence for Accidents

The avertable dose is the Early phase dose which would be avoided if the affected population were evacuated. For the Early phase, the RSGB SRF event is used to evaluate the emergency response, since the RD-337RD-337 SRF event is designed to address short term evacuation cases. For an event where the release is held within the reactor containment for 24 hours or more, the evacuated population would not be exposed. In this case, the avertable dose would be the full 7 day early phase dose. In the event of a failure of containment isolation, the time for evacuation and sheltering measures would be shorter, and the avertable dose would be proportional then to the time of exposure, allowing for decay over time, and the time to travel out of the affected area. For the purposes of assessment, the dose calculations extend to a minimum distance of 500 m from the reactor building.

From Figure 4.4-1, showing the RSGB SRF Early dose, the lower PAL of 10 mSv for evacuation is met for people within approximately 2 km of the point of release. The PAL for mandatory evacuation is 100 mSv, which would only occur within the site boundary. The same is true if the thyroid dose and corresponding PALs are considered, as shown in Figure 4.4-2.

The dose that could be avoided by evacuation would be approximately 25 mSv in the 1 km evacuation range and up to around 5 mSv out to 3 km.

The results indicate that sheltering for an event such as that modeled could be mandatory to around 2 km from the point of release, while the option of sheltering may be considered for out to 10 km.

The relocation limit for dose over 1 year is not exceeded beyond 1 km from the point of release, and as such, no permanent residents would require relocation.

It is possible that ingestion dose control measures, such as control of milk, livestock and produce would be invoked for a period to minimize long term dose to the population as provided for in the Provincial Nuclear Emergency Plan’s Ingestion Control Phase for post-accident mitigation.

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5.0 CONSIDERATION OF THE FEASIBILITY OF AN EMERGENCY PLAN


From Figure 4.4-1, it can be seen that a dose of around 25 mSv might be expected at 1 km from the point of release during the Early phase if no mitigation measures were taken. This section provides an assessment of how the Emergency Plan for Darlington is impacted by the calculated dose consequences. The intent of this section is two-fold:

• To analyze compliance of the existing emergency plans against the IAEA guidance on emergency planning.

• To provide a judgement regarding the likely impacts of the OPG New Nuclear at Darlington on these plans.

5.1 Objectives for Emergency Planning

According to IAEA GS-R-2 (“Preparedness and Response for a Nuclear or Radiological Emergency”), the primary objectives for protection and safety as part of emergency planning are as follows:

• “Protection objective: to prevent the occurrence of deterministic effects in individuals by keeping doses below the relevant threshold and to ensure that all reasonable steps are taken to reduce the occurrence of stochastic effects in the population at present and in the future.”

• “Safety objective: to protect individuals, society and the environment from harm by establishing and maintaining effective defences against radiological hazards from sources.”

• “Radiation protection objective: To ensure…mitigation9 of the radiological consequences of any accidents.”

• “Technical safety objective: To take all reasonably practical measures to prevent accidents in nuclear installations and to mitigate their consequences should they occur; to ensure with a high level of confidence that, for all possible accidents taken into account in the design of the installation, including these of very low probability, any radiological consequences would be minor and below prescribed limits...”

IAEA GS-R-2 also states that measures taken to achieve these objectives (undertaking interventions) are governed at all times by the following principles:

• “Justification of intervention: Any proposed intervention shall do more good than harm.”

9 This is not to imply that the effects would be entirely mitigated through the use of emergency response. Instead, the intent is to maintain doses within the regulatory limits.

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• “Optimization of intervention: The form, scale and duration of any intervention shall be optimized so that the net benefit is maximized.”

The objectives of emergency planning are most likely to be achieved in accordance with the principles for intervention by having a sound programme for emergency preparedness in place as part of the infrastructure for protection and safety. Emergency preparedness also helps to build confidence that an emergency response would be managed, controlled and co-ordinated effectively.

A detailed description of the general, functional and infrastructure requirements for effective emergency planning and response as listed in this document are available in Appendix B.

5.2 Existing Emergency Plans

The Province of Ontario Nuclear Emergency Plan (PNEP) (“Province of Ontario Nuclear Emergency Plan: Part I – Provincial Master Plan”, Interim Plan, 2nd Edition, Ontario, March 1999) provides the basis upon which nuclear emergency planning, preparedness and response shall be undertaken to safeguard the health, safety, welfare and property of the inhabitants of the province, and to protect the environment in the event of a nuclear emergency. The plan also includes a Darlington specific section (“Province of Ontario Nuclear Emergency Plan: Part IV – Darlington Nuclear Emergency Plan”, Interim Plan, 2nd Edition, Ontario, December 1998).

The PNEP Part I specifies the overall principles, policies, basic concepts, organizational structures and responsibilities. PNEP Part I has been revised and is now called the Provincial Nuclear Emergency Response Plan (PNERP). The PNERP received Cabinet approval at the end of January 2009 and was issued by an Order of Council on February 11th, 2009. Subsequently, OPG is now working with the province on implementation of the PNERP. The information used in this report is not affected by the revision.

The PNEP divides the area surrounding the plant into zones and sectors, with the level of planning and preparedness being the highest in the regions nearest to the plant. In the event of an accident, projected doses are determined for the most exposed individual in the most critical group. The response to the accident is determined by comparing the projected dose to a set of Protective Action Levels (PALs), for which protective measures have been identified. The type of accident also determines the timing of radioactive releases from the plant to the public, which in turn affects the emergency response to the accident.

The area around the boundary of a nuclear installation for which a nuclear emergency plan is made is divided into the following zones:

• Contiguous Zone (~3 km radius): The zone immediately surrounding the nuclear installation. An increased level of emergency planning and preparedness shall be

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undertaken within this area because of its proximity to the source of the potential hazard.

• Primary Zone (~10 km radius): The zone around the nuclear installation within which detailed planning and preparedness shall be carried out for measures against exposure to a radioactive plume, including evacuation. (The Primary Zone includes the Contiguous Zone).

• Secondary Zone (~50 km radius): A larger zone within which it is necessary to plan and prepare measures against exposure from ingestion of radioactive material. (The Secondary Zone includes both the Primary and Contiguous Zones).

The specific sector boundary map for the Primary Zone around the Darlington Nuclear Generating Station is shown in Figure 5.2-1. The Primary Zone is itself subdivided into 3 sub-zones called “rings” as follows:

• Inner Ring (Contiguous Zone) - Sector D1 • Middle Ring - Sectors D2, D3, D4, D5, and lake sectors D14 and D15. (D14 and

D15 are not shown in the Figure).

• Outer Ring - Sectors D6, D7, D8, D9, D10, D11, D12, D13, and lake sectors D16 and D17. (D16 and D17 are not shown in the Figure).

• Sectors D6 and D8 are each divided into subsectors A and B for evacuation planning.

Figure 5.2-1: Darlington Specific Response Sectors in Primary Zone

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The Secondary Zone encompasses areas of Durham Region, the City of Toronto, York Region, the City of Kawartha Lakes, and the counties of Northumberland and Peterborough beyond the Primary Zone and within a 50 km radius of Darlington NGS.

The following table provides the boundaries for each of the land and lake emergency response sectors in the Primary Zone.

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Table 5.2-1: Response Sector Boundaries for Darlington [Ontario 1998]

5.3 Protective Action Levels

Protective Action Levels (PALs) serve as aids in planning and decision-making during an emergency. Expressed in terms of projected radiation doses, they provide technical guidance on the need to take specific protective measures, such as

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evacuation, sheltering, and thyroid blocking. PALs for banning the consumption of affected foods and water are expressed as levels of radionuclide concentrations.

PALs for a protective measure which are laid down as a lower and an upper level carry the following significance:

• Lower Level o Below this level, the protective measure would not normally be justified. At

or above this level, the protective measure should be applied unless valid reasons exist for deferring action.

• Upper Level o At or above this level, the protective measure shall be implemented, unless

implementation clearly entails greater risks for the people involved than those from the projected radiation dose.

The complete set of specific Protective Action Levels (including ingestion) that have been established for Emergency Planning in Ontario are provided in Table 5.3-1.

Table 5.3-1 Protective Action Levels (PALs)

Exposure Control Measures

Lower Level Upper Level Protective

Measure Effective

Dose Thyroid Dose

Effective

Dose Thyroid Dose

Sheltering 1 mSv

(0.1 rem)

10 mSv

(1 rem)

10 mSv

(1 rem)

100 mSv

(10 rem)

Evacuation 10 mSv

(1 rem)

100 mSv

(10 rem)

100 mSv

(10 rem)

1 Sv

(100 rem)

Thyroid Blocking - 100 mSv

(10 rem) -

1 Sv

(100 rem)

Ingestion Control Measures

Radionuclide Concentration Level Banning

Food/Water

Consumption

Cs-134, Cs-137,

Ru-103, Ru-106, Sr-89

I-131 Sr-90

Am-241, Pu-238,

Pu-239, Pu-240, Pu-242

Foods for General

Consumption

1 kBq (27 nCi)

per kg

100 Bq (2.7 nCi)

per kg

10 Bq (270 pCi)

per kg

Milk, Infant Foods,

Drinking Water

1 kBq (27 nCi) per

kg

100 Bq (2.7 nCi)

per kg

1 Bq (27 pCi)

per kg

The PNEP [Ontario 1999] also states a dose limit of no more than 20 mSv (2 rem) in 1 year from an accident; otherwise, relocation may be required.

In addition, Health Canada also recommends the Intervention Levels as set out in Table 4.4-2, for countermeasures following a nuclear emergency.

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5.4 Comparison with IAEA Requirements and Recommendations from Other Agencies

Provisions identified in the existing Province of Ontario Nuclear Emergency Plan and the Darlington Nuclear Emergency Plan were compared against the specific requirements listed in IAEA GS-R-2, which are derived from the high-level objectives and principles noted in the introductory section. The detailed comparison against requirements can be found in Appendix B.

Based on a review of the emergency preparedness provisions laid out by the existing Plans, it is judged that the IAEA requirements are met. Furthermore, the Protective Action Levels specified by the Plan are either at a lower threshold than or similar to the equivalent IAEA intervention levels.

An exhaustive review of the comparison of the PNEP and Health Canada PALs/Intervention Levels with recommendations from other agencies concludes that much similarity exists [Health Canada, 2003]. The following table is reproduced from that report as a summary of this comparison.

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Table 5.4-1 Comparison of Intervention Levels

Agency Sheltering (mSv)

Evacuation (mSv)

Relocation (mSv)

Stable iodine (mSv committed

equivalent dose)

Food Control (mSv)

Health

Canada*

5 in 1 day 50 in 7 days 50/y 100 1(3)**

Ontario 1-10 10 - 100 20/y 100 - 1000 5

Quebec* 5 during the

early phase of the

accident

10 during the

early phase of the

accident

20 for the

first year

50 2 per food group

in the first year; 1 per food group

in following years

New

Brunswick

50 100 5

USA (EPA) 5-50 10-50 50 lifetime 250 5

UK (NRPB)*

3-30 30 - 300 30 - 300 3

IAEA

(1994)*

10 in 2 days 50 in 7 days 30/mo 100 5

ICRP

(1993)*

5-50 50 - 500 5 - 15/mo 50 - 500 10

Notes

* Values are doses averted

** Apportioned equally among each of three food groups, or 1 mSv per food group (fresh milk,

other commercial foods and beverages, and public drinking water)

5.5 Potential Impact on Emergency Planning of OPG’s New Nuclear at Darlington

Based on the findings of this study, the existing Darlington Nuclear Emergency Plan will suffice for the OPG New Nuclear at Darlington as well. A recent study of evacuation time estimates was carried out to assess the adequacy of the emergency plan, including evacuation during periods of inclement weather [OPG 2009c]. The study also addresses the overall adequacy of the Emergency Plan for New Nuclear.

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6.0 QUALITY ASSURANCE PROGRAMME

All work is performed in compliance with the project Quality Assurance Plan [AMEC NSS 2008c] and meets the AMEC NSS Quality Assurance program.

The AMEC NSS Quality Assurance program is ISO 9001 registered, and is compliant with CSA Z299.1-85 as well as the applicable portions of CSA N286.2-00, Design Quality Assurance for Nuclear Power Plants and CSA N286.7-99, Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants.

The scope of the AMEC NSS ISO 9001:2001 registration covers “consulting to nuclear and other industries to support design and operations by providing specialized: software, integrated analytical and engineering solutions and services”. The program has also been audited by CANPAC and confirmed to meet the requirements of the applicable sections of CSA N286.2-00 and N286.7-99.

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7.0 CONCLUSIONS

This report presents an outcome of the analysis of how the requirements and recommendations set forth in the IAEA Safety Standard NS-G-3.2 are addressed with regards to the project to construct a new nuclear power plant at the Darlington Nuclear Site. Furthermore it addresses the requirements from CNSC Site Evaluation Guide RD-346 [CNSC 2008a] and the IAEA Requirements Document NS-R-3 in as far as they pertain to dispersion of radioactive material in air and water and analysis of population distribution in site evaluation for nuclear power plants.

A summary of the findings and conclusions is presented further:

1. Monitoring data required to assess radiological impact from radioactive releases into the environment and the baseline information on environmental contamination, hydrogeology, meteorology and population distribution and habits have been collated. The Darlington Nuclear Site provides comprehensive datasets surpassing10 the requirements of IAEA and CNSC guides with data collated over several decades to monitor performance of the operational nuclear generating station. These have been recently complemented with additional information as part of the Environmental Assessment which is being conducted by Ontario Power Generation.

2. Representative releases of radioactive materials have been estimated for both normal operations and accidents. These estimates were made on the basis of information provided by reactor vendors and, in the case of abnormal releases, on the basis of the Canadian regulatory requirements specifying frequency and credible11 threshold releases. It is assumed that any reactor design licensed for construction in Canada will satisfy these requirements.

3. An assessment of atmospheric, surface water and groundwater dispersion of radioactive materials in the environment was conducted for both normal operations and accidental releases. Doses to the most exposed members of the public have been estimated.

4. Estimated maximum annual doses for normal operations are 5 µSv for both once-through cooling option and cooling tower option, which represent a small fraction of the dose limit of 1 mSv. This demonstrates that both options are feasible and can satisfy Canadian and IAEA requirements. Additional sensitivity studies were conducted to gauge the impact on dose if in the future population groups were to be located in locations with maximum exposure. Even such a conservative scenario resulted in estimated doses significantly below the regulatory limit.

10 IAEA guide states a minimum of one years worth of meteorological and other data be recorded prior to construction. Site data for several decades are available due to the presence of the existing Darlington Station. 11 The CNSC establishes a threshold of credibility for a nuclear accident for the purposes of Environmental Assessment as an event frequency of greater than 1x10-6 occurrences per year.

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5. Stylized accident release scenarios based on the RD-337 Small Release and Large Release Frequency safety goals were modeled to predict possible doses to the public. These results are below the Protective Action Levels and Intervention Levels for evacuation, relocation, and sheltering within distances established in the existing emergency plans. The overall impact from the New Nuclear at Darlington on emergency planning from an accident event has been considered in [OPG 2009c].

The results indicate that sheltering for an event such as that modeled could be mandatory to around 2 km from the point of release, while the option of sheltering may be considered for out to 10 km. Evacuation however would only require consideration for less than 2 km from the point of release. The mandatory evacuation dose is only exceeded within 500 m from the point of release.

The results indicate that if early phase evacuation takes place, approximately half of the total predicted dose for the release over 1 year is avertable. The relocation limit for dose over 1 year is not exceeded beyond 1 km from the point of release, and as such, no local residents would require permanent relocation. It is possible that temporary relocation measures will be required within this 1 km area for a time immediately following the release and for as long as 1 year. This relocation would only apply to permanent local residents and not to workers or businesses.

Based on the RSGB analyses, the existing Darlington Nuclear Emergency Plan is adequate for the OPG New Nuclear at Darlington as well. An evaluation of the Emergency Planning at Darlington with respect to New Nuclear has been carried out in [OPG 2009c].

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8.0 REFERENCES

AMEC NSS 1996. P0305/CD/005 OPG Raw Meteorological Data Sets for Processing 1996





AMEC NSS 2005. Radiation and Radioactivity Technical Support Document - Bruce A, Refurbishment for Life Extension and Continued Operations, Environmental Assessment. December 5, 2005.

AMEC NSS 2008a, Raw Data from meteorological towers located at Darlington NGS (records between 1991 and 2008 inclusive).

AMEC NSS 2008c, Quality Assurance Plan, P1093 PL 003

AMEC NSS 2009a, Evaluation of Meteorological Events, P1093 RP 004

AMEC NSS 2009b, Evaluation of Flood Hazards, P1093 RP 005

AMEC NSS 2009c, Evaluation of Geotechnical Aspects, P1093 RP 006

AMEC NSS 2009d, Summary Report: Site Evaluation of External Human Induced Events, P1093 RP 011

AMEC NSS 2009e, Radiation & Radioactivity Environment - Existing Environmental Conditions Technical Support Document, New Nuclear – Darlington Environmental Assessment, SN002/RP/001; NK054-REP-07730-00008

AREVA 2007. U.S. EPR Tier 2 Final Safety Analysis Report Chapter 19 - PRA and Severe Accidents.

Barry Myler 2008. Personal communication: Rainbow Trout residence time. Email from Barry Myler to Mark Gerchikov on October 24, 2008.

Canadian Nuclear Safety Commission 2000 (CNSC). Regulatory Guide: Computer Programs Used in Design and Safety Analyses of Nuclear Power Plants and Research Reactors. CNSC G-149

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Canadian Nuclear Safety Commission. 2008a, November – Regulatory Document - Site Evaluation for New Nuclear Power Plants. No. RD 346

Canadian Nuclear Safety Commission. 2008b, November – Regulatory Document - Design of New Nuclear Power Plants. No. RD 337

Canadian Standards Association (CSA). 1999. Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants, CSA N286.7-99

Canadian Standards Association. 2008. Guidelines for calculating derived release limits for radioactive material in airborne and liquid effluents for normal operation of nuclear facilities, CSA-N288.1-08.

CH2MHill 2009a, Geological and Hydrogeological Environment - Existing Environmental Conditions Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00005

CH2MHill 2009b, Geological and Hydrogeological Environment - Assessment of Environmental Effects Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00015

Durham 1993, (November 1993) Official Plan for the Regional Municipality of Durham

Ecometrix Inc. 2006a. IMPACT Professional Version 5.2.2. Released December 12, 2006

Ecometrix Inc. 2006b. IMPACT 5.0 Tool Qualification Report. Report prepared for CANDU Owners Group Inc. Ecometrix 04-1123.Ecometrix Inc. 2008. CANDU Owners Group document COG-06-3090-R2-I; Derived Release Limits Guidance; Hart, D.; November 2008

Golder & SENES, 2009a, Aquatic Environment, Existing Environmental Conditions, Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00003

Golder, 2009b, Surface Water Environment, Existing Environmental Conditions Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00002

Health Canada 2003. Canadian Guidelines for Intervention During a Nuclear Emergency. November. Accessed on November 13, 2008 at http://www.hc-sc.gc.ca/ewh-semt/alt_formats/hecs-sesc/pdf/pubs/radiation/guide-03/interventions-eng.pdf

International Atomic Energy Agency. 2002a, External Human Induced Events in Site Evaluation for Nuclear Power Plants. No. NS-G-3.1

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International Atomic Energy Agency. 2002b, Dispersion of Radioactive Material in Air and Water and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants. No. NS-G-3.2

International Atomic Energy Agency. 2002c, Evaluation of Seismic Hazards for Nuclear Power Plants. No. NS-G-3.3

International Atomic Energy Agency. 2002d, Meteorological Events in Site Evaluation for Nuclear Power Plants. No. NS-G-3.4

International Atomic Energy Agency. 2002e, Flood Hazard for Nuclear Power Plant on Coastal and River Sites. No. NS-G-3.5

International Atomic Energy Agency. 2002f, Geotechnical Aspects of Site Evaluation and Foundation for Nuclear Power Plants. No. NS-G-3.6

International Atomic Energy Agency. 2002g, Preparedness and Response to Nuclear or Radiological Emergency Safety Requirements. No. GS-R-2 (Used in Appendix B)

International Atomic Energy Agency. 2003, IAEA Safety Standard Series – Site Evaluation for Nuclear Installations. No. NS-R-3

Kunz C. 1985. Carbon-14 discharge at three light-water reactors. Health Phys. 1985 Jul;49(1):25-35.

MOE 2006, Approved Environmental Assessment Terms of Reference, http://www.durhamyorkwaste.ca/pdfs/study/Annex_A_(Approved_EA_Terms_of_Reference).pdf

MSC 2008 – Canadian Climatological data digitally archived and quality controlled by the Meteorological Service of Canada (MSC), emailed to Arindam Mukherjee [[email protected]] by Ontario Climate Center [[email protected]] on 3rd July, 22nd July and 30th September 2008.

NRC 2007. NUREG-0800, Section 19, “Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors,” SRP. U.S. Nuclear Regulatory Commission, June 2007.

Ontario 1998. Province of Ontario Nuclear Emergency Plan: Part IV – Darlington Nuclear Emergency Plan, Interim Plan, 2nd Edition, Ontario, December 1998

Ontario 1999. Province of Ontario Nuclear Emergency Plan: Part I – Provincial Master Plan, Interim Plan, 2nd Edition, Ontario, March 1999

Ontario 2003. Province of Ontario: Ministry of the Environment – Stormwater Management Planning and Design Manual 2003.

Ontario Hydro, 1984, A Study of the Dispersion Climatology of the Pickering Area, Report 84-137-K

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Ontario Hydro, 1991, Lake Ontario Water Temperatures in the Vicinity of Darlington NGS 1987-1989, Report 91243

Ontario Hydro, 1992, Lake Ontario Nearshore Currents in the Vicinity of Darlington NGS 1987-1989, Report 92184

Ontario Power Generation (OPG) 1991a. Quarterly Technical Report Fourth Quarter 1990. PQR-4-90A, R-PGS-09050-03 90-04.

OPG 1991b. Quarterly Technical Report First Quarter 1991. PQR-1-91-A, NA44-09050-3

OPG 1991c. Quarterly Technical Report Second Quarter 1991. PQR-2-91-A, R-PGS-09050-3

OPG 1991d. Quarterly Technical Report Third Quarter 1991. PQR-3-91-A, R-PGS-09050-3

OPG 1993. Quarterly Technical Report Fourth Quarter 1992. PQR-4-92-A, NA44-09050-3

OPG 1994. Quarterly Technical Report Fourth Quarter 1993. PQR-4-93-A, NA44-09050-3

OPG 2002. Guidance For Calculation Of Derived Release Limits For Radionuclides In Airborne And Liquid Effluents From Ontario Power Generation Nuclear Facilities. N-REP-03482-10000-R00

OPG 2004. Darlington Safety Report, NK38-SR-03500-10001

OPG 2006. Review of the Darlington Nuclear Site Specific Survey. OPG document NK38-REP-03481-10002-R00. December 12, 2006

OPG 2007. Radiation and Radioactivity Technical Support Document. Refurbishment and Continued Operation of Pickering B Nuclear Generating Station Environmental Assessment. Report NK30-REP-07701-00004

OPG 2008a. A Letter from AECL to OPG (Ms. Laurie Swami): Response Regarding Emission Information for Environmental Assessment. Oct 31, 2008

OPG 2008b. New Nuclear Site Preparation Option Studies, OPG number N-REP-01210-0249632, prepared by NSS. May 30, 2008.

OPG 2008c. Darlington Radiological Environmental Monitoring Program (REMP) Review. OPG document NK38-REP-03481-10003-R000. June 30, 2008

OPG 2008d. 2007 Results of Radiological Environmental Monitoring Program. OPG document N-REP-03481-10006 R000. April 23, 2008

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OPG 2009a. Land Use, Assessment of Environmental Effects Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00016

OPG 2009b. Nuclear Waste Management System, Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00027

OPG 2009c. Emergency Preparedness Evaluation for OPG New Nuclear At Darlington License Submission. OPG document NK054-REP-03490-00001-R00. 2009-04-01.

OPG 2009d. Use of Plant Parameters Envelope to Encompass the Reactor Designs Being Considered for the Darlington Site, N-REP-01200-10000-R002, March 11 2009

Regional Municipality of Durham 2008, Growing Durham Study, Scenario Evaluation and Recommended Preferred Growth Scenario Working Paper and Addendum. May.

SANDIA 1990. NUREG/CR-4551, Vol.2, SAND86-1309 Rev 1 Part 7 – Evaluation of Severe Accident Risks: Quantification of Major Input Parameters: MACCS2 Input.

SANDIA 1997, NUREG/CR-6547; SAND97-2776 DOSFAC2 Users Guide, December.

SANDIA 1998. NUREG/CR-6613, Vol.1; SAND97-0594 Code Manual for MACCS2 Users Guide, May.

SENES 2009a. Memorandum: Distribution of Fish in the Vicinity of DNGS. Feb 17, 2009

SENES 2009b. Memorandum: Dry Matter Consumption Rates by Animals. Feb 10, 2009

SENES 2009c. Malfunctions, Accidents and Malevolent Acts, Assessment of Environmental Effects Technical Support Document, New Nuclear – Darlington Environmental Assessment; NK054-REP-07730-00024

US DOE 2004. DOE-EH-4.2.1.4-MACCS2-Code Guidance - MACCS2 Computer Code Application Guidance for Documented Safety Analysis – Final Report.

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Appendix A: Abbreviations

ACR (1000) - Advanced CANDU Reactor

ADF - Atmospheric Dispersion Factor

AECL - Atomic Energy of Canada Ltd

AOO - Anticipated Operational Occurrences

AP (1000) - Advanced Passive Reactor

BDBA - Beyond Design Basis Accident

CI - Core Inventory

CNSC - Canadian Nuclear Safety Commission

DBA - Design Basis Accident

DCF - Dose conversion factor

DF - Dilution Factor

DNGS - Darlington Nuclear Generating Station

EA - Environmental Assessment

ECD - Effective committed dose

EPR - Areva's US EPR Reactor

GIS - Geographical Information System

IAEA - International Atomic Energy Agency

IMPACT - Integrated Model for Probabilistic Assessment of Contaminated Transport

LOCA - Loss of coolant accident

LRF - Large Release Frequency as defined in Canada (by RD-337RD-337)

MACCS2 - MELCOR accident consequence code system

MSC - Meteorological Services Canada

NGS - Nuclear Generating Station

NND - New Nuclear at Darlington

OBT - Organically Bound Tritium

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OPG - Ontario Power Generation

PAL - Protective Action Level

PCG - Potential Critical Groups

PHR - Pressurized Hybrid Reactor

PNEP - Province of Ontario Nuclear Emergency Plan

PWR - Pressurised water Reactor

REMP - Radiological Environmental Monitoring Program

RF - Release Fraction

RSGB - RD-337 Safety Goal Based (relating to release or source term)

SRF - Small Release Frequency

TIBL - Thermal Inversion Boundary Layer

US LRF - Large Release Frequency as defined in the US Licensing Process

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Appendix B: EXTRACTS FROM IAEA GS-R-2 [IAEA 2002g]

NOTE

Extracts from IAEA are listed below, along with the Section # of this document they can be found under in italics.

PRINCIPLES

EMERGENCY RESPONSE

The Safety Fundamentals publication on Radiation Protection and the Safety of Radiation Sources [Radiation Protection and the Safety of Radiation Sources, Safety Series No. 120, IAEA, Vienna (1996)] presents the primary objectives for protection and safety as follows:

• “Protection objective: to prevent the occurrence of deterministic effects in individuals by keeping doses below the relevant threshold and to ensure that all reasonable steps are taken to reduce the occurrence of stochastic effects in the population at present and in the future.” (Section 2.1)

• “Safety objective: to protect individuals, society and the environment from harm by establishing and maintaining effective defences against radiological hazards from sources.” (Section 2.1)

The Safety Fundamentals publication on The Safety of Nuclear Installations [The Safety of Nuclear Installations, Safety Series No. 110, IAEA, Vienna (1993)] presents the primary objectives for nuclear installations12 as:

• “Radiation protection objective: To ensure… mitigation of the radiological consequences of any accidents.” (Section 2.2)

• “Technical safety objective: To take all reasonably practical measures to prevent accidents in nuclear installations and to mitigate their consequences should they occur; to ensure with a high level of confidence that, for all possible accidents taken into account in the design of the installation, including these of very low probability, any radiological consequences would be minor and below prescribed limits...” (Section 2.2)

In a nuclear or radiological emergency, the practical goals of emergency response are (Section 2.3):

(a) To regain control of the situation;

(b) To prevent or mitigate consequences at the scene;

(c) To prevent the occurrence of deterministic health effects in workers and the public;

12 Any authorized facility that is part of the nuclear fuel cycle except for radioactive waste management facilities.

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(d) To render first aid and to manage the treatment of radiation injuries;

(e) To prevent, to the extent practicable, the occurrence of stochastic health effects in the population;

(f) To prevent, to the extent practicable, the occurrence of non-radiological effects on individuals and among the population;

(g) To protect, to the extent practicable, property and the environment;

(h) To prepare, to the extent practicable, for the resumption of normal social and economic activity.

Taking measures towards achieving these goals (undertaking interventions) is governed at all times by the principles established in the Safety Fundamentals publication on Radiation Protection and the Safety of Radiation Sources [Radiation Protection and the Safety of Radiation Sources, Safety Series No. 120, IAEA, Vienna (1996)] and derived from the recommendations of the ICRP [International Commission On Radiological Protection, 1990, Recommendations of the International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford and New York (1991); International Commission On Radiological Protection, Principles for Intervention for Protection of the Public in a Radiological Emergency, Publication 63, Pergamon Press, Oxford and New York (1993).]. These principles are:

• “Justification of intervention: Any proposed intervention shall do more good than harm.” (Section 2.4)

• “Optimization of intervention: The form, scale and duration of any intervention shall be optimized so that the net benefit is maximized.” (Section 2.4)

EMERGENCY PREPAREDNESS

The goals of emergency response are most likely to be achieved in accordance with the principles for intervention by having a sound programme for emergency preparedness in place as part of the infrastructure for protection and safety [International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series No. 115, IAEA, Vienna (1996)]. (Section 2.5)

Emergency preparedness also helps to build confidence that an emergency response would be managed, controlled and co-ordinated effectively. The practical goal of emergency preparedness may be expressed as follows: “To ensure that arrangements are in place for a timely, managed, controlled, co-ordinated and effective response at the scene, and at the local, regional, national and international level, to any nuclear or radiological emergency.” (Section 2.6)

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REQUIREMENTS

General Requirements

Adequate preparations shall be established and maintained at local and national levels and, where agreed between States, at the international level to respond to [nuclear or radiological] emergencies. (Section 3.1)

The arrangements for emergency response actions both within and outside facilities, if applicable, or elsewhere under the control of the operator, are dealt with through the regulatory process. (Section 3.2)

It is presumed that the State will have determined in advance the allocation of responsibilities for the management of interventions in emergency exposure situations between the [regulatory body], national and local [response organizations] and [operators]. (Section 3.3)

Legislation shall be adopted to allocate clearly the responsibilities for preparedness and response for a nuclear or radiological emergency, including establishing or identifying an existing governmental body or organization to act as a national co-ordinating authority, which will ensure that the functions and responsibilities of operators and response organizations are clearly assigned and are understood by all response organizations, and that arrangements are in place for achieving and enforcing compliance with the requirements. (Section 3.4, 3.5)

For the purposes of the requirements, nuclear and radiation related threats are grouped according to five threat categories. In the present context, a Threat Category of I (one) is applicable, which applies to Facilities, such as nuclear power plants, for which on-site events13 (including very low probability events) are postulated that could give rise to severe deterministic health effects14 off the site, or for which such events have occurred in similar facilities. (Section 3.6)

For a facility in threat category I, appropriate emergency preparedness and response arrangements shall be established from the time that nuclear fuel or significant amounts of radioactive or fissile material is brought to the site, and complete emergency preparedness as described here shall be ensured/tested before the commencement of operation. (Section 3.8)

In designing a threat category I facility (Section 3.13, 3.14)

• a probabilistic safety analysis of the facility shall be carried out in order to assess the adequacy of the operator’s emergency response arrangements

• a comprehensive safety analysis is carried out to identify all sources of exposure and to evaluate radiation doses that could be received by workers at the facility and the public, as well as potential effects on the environment. The safety analysis examines event sequences that may lead to an emergency. On the basis of this analysis, requirements for emergency preparedness and response can be established.

13 Involving an atmospheric or aquatic release of radioactive material or external exposure (such as due to a loss of

shielding or a criticality event) that originates from a location on the site. 14 Doses in excess of those for which intervention is expected to be undertaken under any circumstances.

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In the threat assessment (Section 3.15-3.18),

• Emergencies involving a combination of a nuclear or radiological emergency and a conventional emergency such as an earthquake shall be considered. (Ontario is in a region of low seismic activity.)

• Any threat associated with nuclear facilities in other States shall also be considered. • Any populations at risk shall be identified and, to the extent practicable, the likelihood,

nature and magnitude of the various radiation related threats shall be considered. The threat assessment shall be so conducted as to provide a basis for establishing detailed requirements for arrangements for preparedness and response.

• Facilities, sources, practices, on-site areas, off-site areas and locations shall be identified for which a nuclear or radiological emergency could warrant:

o Precautionary urgent protective action to prevent severe deterministic health effects by keeping doses below those for which intervention would be expected to be undertaken under any circumstances;

o Urgent protective action to prevent stochastic effects to the extent practicable by averting doses, in accordance with international standards;

o Agricultural countermeasures, countermeasures to ingestion and longer term protective measures, in accordance with international standards; or

o Protection for the workers responding (undertaking an intervention), in accordance with international standards.

• Non-radiological threats to people on and off the site shall be identified.

Operators, the national co-ordinating authority and other appropriate organizations shall periodically conduct a review in order to ensure that all practices or situations that could necessitate an emergency intervention are identified, and shall ensure that an assessment of the threat is conducted for such practices or situations. (Section 3.16)

Functional Requirements (Section 4.1-4.100)

This is a summary of the highlights of the Functional Requirements listed in IAEA GS-R-2 [IAEA 2002g] and is not intended to be exhaustive.

• Emergency response shall be executed both on and off-site. • Coordination between various response organizations shall be ensured. • Operators shall determine the emergency class (see below) the accident falls in or the

level of emergency response required. o See list of emergency classes in Item 4.19 of IAEA GS-R-2 [IAEA 2002g]

• All Emergency Response Organizations (ERO) shall be notified including transborder EROs.

• Preparedness includes notifying first responders of the nature of the contamination, and its emergency class.

• An individual shall always be present on site who can classify the emergency, initiate without consultation the appropriate on-site emergency response, and provide

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information for effective off-site emergency response to the appropriate points of contact.

• Operator training on mitigation measures shall include beyond design basis accidents. • Protection measures shall be provided to emergency workers. • Initial response includes promptly assessing abnormal conditions at the facility;

exposures/releases of radioactive material; radiological conditions on/off-site; and any actual/potential exposure of the public to guide mitigation measures and emergency classification.

• Specification of off-site emergency zones shall be performed. Jurisdictions within these zones shall take appropriate action when notified of the emergency.

• Special population groups shall be remembered when giving out notifications (e.g. hospitalized persons, elderly, children).

• Medical response includes immediate treatment (as symptoms manifest) and tracking the incidence of cancer and appropriate treatment.

• Facilities shall make arrangements to treat overexposed or contaminated workers including arrangements for first aid, the estimation of doses, medical transport and the initial medical treatment of contaminated or highly exposed individuals in local medical facilities.

• The public shall be kept informed with useful, timely, truthful, consistent and appropriate information throughout the emergency.

• Arrangements shall be made for relocation when necessary, for monitoring of contamination levels of vehicles, personnel and goods moving into and out of contaminated areas, for radioactive waste management, and for agricultural countermeasures when necessary.

• Intervention or action levels shall be developed by the jurisdictions to guide activities such as sheltering, evacuation, and resettlement/return of evacuees

• Recovery operations and resumption of normal social and economic activities shall be planned for. Recovery workers shall continue to be subject to occupational exposure requirements as specified by IAEA GS-R-2 [IAEA 2002g].

Infrastructure Requirements (Section 5.1-5.39)

This is a summary of the highlights of the Infrastructure Requirements listed in IAEA GS-R-2 [IAEA 2002g] and is not intended to be exhaustive.

• The authority for developing, maintaining and regulating arrangements for preparedness and response for a nuclear or radiological emergency shall be established by means of acts, legal codes or statutes.

o Organizational relationships shall be defined to support the same. • Plans and procedures shall be put in place to coordinate the emergency response. • Logistical support by appropriate organizations/levels of government shall be made

available.

• Training, drills, and exercises shall be performed. • A Quality Assurance programme shall be developed to cover all aspects of the

emergency response programme.

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ADDITONAL REQUIREMENTS AND COMPARISON WITH THE PNEP

Requirements On Protection For Workers Undertaking An Intervention (Annex I)

When undertaking intervention, all reasonable efforts shall be made to keep doses to workers below twice the maximum single year dose limit, except for life saving actions, in which every effort shall be made to keep doses below ten times the maximum single year dose limit in order to avoid deterministic effects on health. In addition, workers undertaking actions in which their doses may approach or exceed ten times the maximum single year dose limit shall do so only when the benefits to others clearly outweigh their own risk.

The limit on effective dose for a Nuclear Energy Worker (NEW) in Canada is consistent with ICRP recommendations, at 50 mSv over any one year dosimetry period, or 100 mSv over a five year period (equivalent to a rolling annual average of 20 mSv). Twice the maximum single year dose limit for a NEW therefore translates to 100 mSv, while ten times the limit amounts to 500 mSv. For non-NEWs, the annual limit on effective dose is 1 mSv.

By comparison, in the PNEP, the maximum permissible radiation exposure limit for an emergency worker during initial emergency phase shall be an effective dose of 50 mSv (5 rem), and 5 mSv/y in the follow-up phase.

Workers who undertake actions in which the dose may exceed the maximum single year dose limit shall be volunteers and shall be clearly and comprehensively informed in advance of the associated health risk, and shall, to the extent feasible, be trained in the actions that may be required.

Once the emergency phase of an intervention has ended, workers undertaking recovery operations, such as repairs to the facility and buildings, waste disposal or decontamination of the site and surrounding area, shall be subject to the full system of detailed requirements for occupational exposure.

Workers shall not normally be precluded from incurring further occupational exposure because of doses received in an emergency exposure situation. However, qualified medical advice shall be obtained before any such further exposure if a worker who has undergone an emergency exposure receives a dose exceeding ten times the maximum single year dose limit or at the worker’s request.

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Dose Levels At Which Intervention Is Expected To Be Undertaken Under Any Circumstances (Annex II)

ACTION LEVEL OF DOSE FOR ACUTE EXPOSURE, BY ORGAN OR TISSUE

Organ or tissue Action level of dose: Projected absorbed dose to the organ or tissue

in less than 2 days (Gy)

Whole Body (Bone Marrow) 1

Lung 6

Skin 3

Thyroid 5

Lens of the eye 2

Gonads 3

It should be noted that the PALs listed in the PNEP and the Darlington Emergency Plan are at a lower threshold than those listed above.

Guidelines for Intervention Levels and Action Levels in Emergency Exposure Situations (Annex III)

Urgent Protective Actions: Sheltering, Evacuation, Iodine Prophylaxis

The generic optimized intervention level for sheltering is 10 mSv of avertable dose in a period of no more than 2 days. Authorities may wish to advise sheltering at lower intervention levels for shorter periods or so as to facilitate further countermeasures, e.g. evacuation. By comparison, in the PNEP, sheltering may be directed for sectors likely to receive doses above the lower PAL for sheltering of 1 mSv (Whole body).

The generic optimized intervention value for temporary evacuation is 50 mSv of avertable dose in a period of no more than 1 week. Authorities may wish to initiate evacuation at lower intervention levels for shorter periods, and also where evacuation can be carried out quickly and easily, e.g. for small groups of people. Higher intervention levels may be appropriate in situations where evacuation would be difficult, e.g. for large population groups or if there is inadequate transport. By comparison, in the PNEP, evacuation may be directed for sectors likely to receive doses above the lower PAL for evacuation of 10 mSv (Whole body).

The generic optimized intervention value for iodine prophylaxis is 100 mGy of avertable committed absorbed dose to the thyroid due to radioiodine. By comparison, in the PNEP, thyroid blocking may be directed for sectors likely to receive doses above the lower PAL for thyroid blocking of 100 mSv (thyroid).

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Generic Action Levels for Foodstuffs

Radionuclide Generic action level (kBq/kg)

Foods destined for general consumption

Cs-134, Cs-137, I-131, Ru-103, Ru-106, Sr-89 1

Sr-90 0.1

Am-241, Pu-238, Pu-239, Pu-240, Pu-242 0.01

Milk, infant foods and drinking water

Cs-134, Cs-137, Ru-103, Ru-106, Sr-89 1

I-131, Sr-90 0.1

Am-241, Pu-238, Pu-239, Pu-240, Pu-242 0.001

The radionuclide concentration levels identified under Ingestion Control Measures in the PNEP exactly match these Action Levels.

Temporary Relocation and Permanent Resettlement

The generic optimized intervention levels for initiating and terminating temporary relocation are 30 mSv in a month and 10 mSv in a month, respectively. If the dose accumulated in a month is not expected to fall below this level within a year or two, permanent resettlement with no expectation of return to homes should be considered. Permanent resettlement should also be considered if the lifetime dose is projected to exceed 1 Sv. The doses to be compared with these intervention levels are the total doses from all routes of exposure that can be averted by taking the countermeasure but usually this will exclude routes involving food and water.

By comparison, according to the PNEP, the evacuees should be permitted to return to those evacuated areas in which the projected dose from external exposure and inhalation over the next year is assessed to be under 20 mSv (2 rem).

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