Title: SITE BOUNDARY CONSIDERATIONS FOR NEW NUCLEAR ...€¦ · Site Boundary Considerations for...

OPG Proprietary Document Number:

NK054-REP-01210-00009 Sheet Number: Revision:

Coversheet

N/A R001 Title:

SITE BOUNDARY CONSIDERATIONS FOR NEW NUCLEAR - DARLINGTON

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.

Site Boundary Considerations for New Nuclear - Darlington

NK054-REP-01210-00009-R001

2009-09-01

Project ID: 10-27600

AMEC NSS: P0977/RP/002 R03

OPG Proprietary

Prepared by: [AMEC NSS – Per attached] Verified by: [AMEC NSS – Per attached] Joshua Guin

Crawford Morrison AMEC NSS

Date Jeremy Edward AMEC NSS

Date

Reviewed by: [AMEC NSS – Per attached] Approved by: [AMEC NSS – Per attached] Rick Russell

Richard Fluke AMEC NSS

Date Ismail Cheng 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

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AMEC NSS Limited P0977/RP/002 R03 Page 2 of 37 Form 114 R14

Revision Summary

Rev Date Author Comments

R00 November 2008 J. Guin , C. Morrison Initial issue.

R01 January 2009 J. Guin , C. Morrison Client comments addressed.

R02 July 2009 J. Guin , C. Morrison Further comments addressed.

R03 August 2009 J. Guin , C. Morrison Further comments addressed and references updated.

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TABLE OF CONTENTS

Page

1.0 BACKGROUND & PROBLEM STATEMENT .....................................................5 2.0 OBJECTIVES ................................................................................................5 3.0 SCOPE..........................................................................................................5 4.0 METHOD ......................................................................................................6 5.0 FACTORS RELEVANT TO EXTENT OF EXCLUSION ZONE / DISTANCE TO THE

SITE BOUNDARY..........................................................................................7 6.0 RELATIVE EFFECT OF DOWNWIND DISTANCE ON AIRBORNE DOSES.........9 6.1 Purpose of investigating dependence of ADF on downwind distance....................... 9 6.2 Method and limitations in calculating ADFs at near distances ................................. 9 6.3 Ratios of ADFs obtained, applicability and interpretation.......................................11 7.0 SOURCE TERM INPUTS FOR DESIGN BASIS EVENTS.................................12 7.1 Inputs required for dose calculation....................................................................12 7.2 Event categorization, frequency and acceptance criterion .....................................13 7.3 US NRC Deterministic Safety Analysis Requirements ............................................13 7.3.1 Accident Source Terms (10 CFR 50.67)...............................................................14 7.3.2 Accident Dose Criteria under Alternative Radiological Source Terms for Different

Accident Types .................................................................................................15 7.4 Available sources of useful data .........................................................................15 8.0 DOSE CONSEQUENCE EVALUATION FOR CANDIDATE REACTOR TYPES....16 8.1 AECL ACR-1000 ................................................................................................16 8.1.1 Anticipated Operational Occurrences (AOOs).......................................................17 8.1.2 Design Basis Accidents (DBAs) ...........................................................................17 8.2 AREVA EPR.......................................................................................................18 8.2.1 Anticipated Operational Occurrences (AOOs).......................................................18 8.2.2 Design Basis Accidents (DBAs) ...........................................................................20 8.3 Westinghouse AP1000.......................................................................................22 8.3.1 Anticipated Operational Occurrences (AOOs).......................................................22 8.3.2 Design Basis Accidents (DBAs) ...........................................................................23 8.3.2.1 Fuel Handling Accident ......................................................................................24 8.3.2.2 Large LOCA......................................................................................................25 8.3.2.3 AP1000 and UK EPR Rationale ...........................................................................25 9.0 CONCLUSIONS...........................................................................................25

REFERENCES.............................................................................................................28 GLOSSARY.................................................................................................................31 APPENDIX A: DARLINGTON NUCLEAR SITE SPECIFIC ADFS ....................................33 APPENDIX B: EVENT CLASSIFICATION.....................................................................36 APPENDIX C: US EPR AND AP1000 DESIGN BASIS ACCIDENTS ...............................37


LIST OF TABLES AND FIGURES

Tables

Table 7.1: Accident Dose Criteria in 10CFR50.67..................................................................15 Table 8.3-1: AP1000 Final Safety Evaluation Report TEDE for DBAs.......................................24 Table 8.3-2: AP1000 Generic Site ADFs ..............................................................................25 Table 9.0-1: RD-310 Event Classification: AOO (f ≥10-2).......................................................27 Table 9.0-2: RD-310 Event Classification: DBA (10-2 > f ≥ 10-5) ............................................27 Table A1: METPROC Results for Atmospheric Dilution Factor Using 1997-2000 Meteorological

Data for Darlington ...........................................................................................33 Table C1: US EPR Final Safety Analysis Report ....................................................................37 Table C2: AP1000 Design Control Document Rev. 16 ...........................................................37

Figures

Figure 8.2-1: Effective Short Term (7 day) Dose In The case of a Severe Accident .................21 Figure A1: Darlington Atmospheric Dilution Factors Relative to 0.914 km Using 1997-2000

Meteorological Data ..........................................................................................34 Figure A2: Atmospheric Dispersion Factors (ADFs) for EDF Flamanville EPR local cities and

DNGS Site Specific Long Term ADFs ...................................................................35 Figure B1: Event Classification for CNSC and Candidate Reactors ..........................................36


1.0 BACKGROUND & PROBLEM STATEMENT

The Government of Ontario has selected the Darlington Nuclear (DN) Site as the location for new nuclear build in Ontario. Reactor technologies now being considered for siting at Darlington represent a subset of those originally identified in the associated EA Project Description [1], which were included in an earlier report [2] looking at Exclusion Zone requirements from a safety analysis perspective.

The broader issue originally examined in that phase of work was how readily the various reactor designs under consideration might be accommodated, in terms of potential radiation exposure implications, within a conventionally sized exclusion zone at the Darlington Nuclear site. It was recognized that, in most cases, reactor types being considered had been demonstrated by vendors to be compliant with licensing processes and standards in a variety of jurisdictions, but had not necessarily been assessed against the relevant requirements applicable to a new reactor to be situated in Ontario, Canada. Now with a more limited set of candidate reactor technologies, the ongoing focus of the material presented in the current report continues to be the extent of the exclusion zone, or more precisely the specification of the site boundary, required to satisfy regulatory requirements applicable at the Darlington Nuclear site.

2.0 OBJECTIVES

The objective of the work undertaken in the initial phase of this project, documented in [2], was to determine, through comparison of applicable requirements and available safety performance data, exclusion zone requirements for reactor designs sited in Canada, specifically considering implications for the Darlington site. Safety issues considered encompassed radiological impacts arising from normal operation, anticipated operational occurrences, design basis accidents, as well as events considered beyond design basis.

The objective of the current phase of work, which is the subject of the current report, is to examine specifically the possibility of justifying a site boundary distance for new nuclear build at the Darlington site which is less than the exclusion zone distance of 914 m in place for the existing Darlington A plant, and at other CANDU plants in Canada. Now that the technology selection process has narrowed to three designs (AECL ACR-1000, Areva EPR, and Westinghouse AP1000), further assessment of the issue will be restricted to those three candidate reactor types.

3.0 SCOPE

Following the completion of the initial phase of work documented in [2] (referred to as Phase 1), a subsequent phase of work was initiated to look further at the issue of the exclusion zone/site boundary distance for the reduced set of candidate reactor designs. The intent was to take account of recent updates in regulatory guidance, as well as possibly to take advantage of greater availability of vendor specific data. It was anticipated the work for this phase would be carried out in two sub- phases, as follows.


Phase 2(a)

An indication of the relative dose distribution versus distance for an airborne release of radioactivity can be obtained by examining the variation with distance of the Atmospheric Dilution Factors (ADFs) involved. Hence, it was planned that calculation of ADFs using the METPROC calculation routine [28] previously undertaken in Phase 1 for downwind distances ranging between 0.914 km and 10 km be extended to encompass closer distances approaching 0.5 km from the release point. The results of these atmospheric dispersion calculations would then be used to infer approximate upper and lower bounds on the ratio of doses expected at an exclusion zone radius of 0.5 km relative to the dose at a distance of 0.914 km.

Phase 2(b)

It was anticipated that the deterministically calculated ADFs derived in Phase 2(a) would be applied in Phase 2(b) to typical or bounding ex-plant release profiles, in order to translate the relative results obtained in Phase 2(a) into absolute terms. It was recognized from the outset that the viability of this approach was contingent on availability of appropriate source term information, with detailed time-dependent release profiles from containment needing to be specified for dose-significant nuclides.

As discussed in later sections of the current report, unavailability of suitable data in the public domain or from vendors has for the most part precluded the completion of this sub-phase on a comprehensive basis as originally planned. Instead, a somewhat more qualitative approach has been followed to reach conclusions concerning the expected dose as a function of distance for the candidate reactor types.

4.0 METHOD

The approach used in executing the scope of the work comprised the following steps:

a) Revisit those requirements and criteria relevant to siting which were previously identified [2] from examination of regulatory requirements and expectations.

b) Update the requirements to reflect any subsequent changes which have taken place since the previous review, such as revisions in 2008 to the set of CNSC Regulatory Documents regarding licensing of new reactors.

c) Undertake sensitivity studies on the dependence of atmospheric dilution factor on downwind distance, to investigate the variation of dose to be expected with change in proximity of the site boundary.

d) Compile data from vendor documentation and from public information sources on airborne releases to the environment associated with design basis events.

e) Evaluate dose to the most exposed member of the public, assumed to be located at the site boundary, with a view to establishing the minimum distance at which the site boundary must be located in order that dose acceptance criteria for design basis events are met. (For ground level releases, as


conservatively assumed in this work, dose decreases monotonically with increasing downwind distance. Hence, the assumption that the individual is located at the site boundary provides an upper bound on dose to the most exposed member of the public located at or beyond the site boundary).

f) Where explicit calculation of dose is not feasible due to unavailability or lack of sufficient detail in ex-plant release information, make use of relevant supporting information to reach conclusions regarding acceptability of the expected dose in each case with respect to the identified criteria.

5.0 FACTORS RELEVANT TO EXTENT OF EXCLUSION ZONE / DISTANCE TO THE SITE BOUNDARY

Based on review of governing documents and applicable requirements, factors relevant to the extent of the exclusion zone or the location of the property boundary were previously identified [2] as being associated primarily with controlling the distance from the plant at which the most exposed member of the population may be subjected to exposure due to radioactive emissions arising from either normal operating conditions, from operational incidents, or from accidents. Although past practice had mandated a fixed distance of 914m for the exclusion zone distance, the underlying consideration for the extent of physical separation was recognized to be the dose, as demonstrated by analysis, to the most exposed member of the public at or beyond the site boundary, arising under normal operating and accident conditions.

Definitions of the relevant terms, as applicable to licensing of reactors in Canada, are for the most part found in applicable regulations and /or regulatory documents issued by the CNSC, as follows.

� Exclusion Zone (defined per Class I Nuclear Facilities Regulations, also quoted

in RD-346 and RD-337) : “A parcel of land within or surrounding a nuclear facility on which there is no permanent dwelling and over which a licensee has the legal authority to exercise control.” � Site (defined in RD-346 glossary) : “The area within the exclusion zone where the NPP and all associated support structures and systems are located.” � Protective Zone (defined in RD-346) : “The area beyond the exclusion zone that needs to be considered with respect to implementing emergency measures. This includes consideration of such matters as population distribution and density, land and water usage, roadways, evacuation planning and consequence analysis.”

The Regulatory Document, Design of Nuclear Power Plants (RD-337, Section 4.2, November 2008 issue) sets out Technical Safety Objectives, the application of which are implemented through dose acceptance criteria for design basis events, and safety goals for beyond design basis events. Although the safety goals (intended for BDBA) involve both qualitative and quantitative goals, none of these measures are expressed


in terms of dose or quantities that are dependent on the extent of exclusion zone or location of the site boundary. Hence only the quantitative dose acceptance criteria (as specified for events within the design basis, i.e. AOOs and DBAs) have a direct influence on the question of site boundary location or exclusion zone radius. Qualitative acceptance criteria are also specified in RD-310, “Safety Analysis for Nuclear Power Plants” (February 2008 issue, section 5.3.4) for AOOs and DBAs, for which corresponding quantitative derived acceptance criteria are to be identified prior to performing analysis. In terms of analysis outputs associated with potential radiation doses to the public, it is to be expected that any derived acceptance criteria would align with the dose acceptance criteria specifically provided in RD-337. The RD-337 Dose Acceptance Criteria are specified as follows. “The committed whole-body dose for average members of the critical groups who are most at risk, at or beyond the site boundary is calculated in the deterministic safety analysis for a period of 30 days after the analyzed event. This dose is less than or equal to the dose acceptance criterion of: 1. 0.5 millisievert for any anticipated operational occurrence (AOO); or 2. 20 millisieverts for any design basis accident (DBA).”

Dose Acceptance Criteria

AOOs DBAs

0.5 mSv 20.0 mSv

The criteria are specified in terms of dose at or beyond the site boundary, rather than with reference to the exclusion zone. The classification of events, including AOOs and DBAs, is provided in RD-310 (section 5.2.3, February 2008 issue) as follows. “The identified events shall be classified, based on the results of probabilistic studies and engineering judgement, into the following three classes of events: 1. Anticipated Operational Occurrences (AOOs) include all events with frequencies of

occurrence equal to or greater than 10-2 per reactor year; 2. Design Basis Accidents (DBAs) include events with frequencies of occurrence equal

to or greater than 10-5 per reactor year but less than 10-2 per reactor year;

3. Beyond Design Basis Accidents (BDBAs) include events with frequencies of occurrence less than 10-5 per reactor year.

Other factors to be considered in the event classification are any relevant regulatory requirements or historical practices. Events with a frequency on the border between


two classes of events, or with substantial uncertainty over the predicted event frequency, shall be classified into the higher frequency class. Credible common-cause events shall also be classified within the AOO, DBA and BDBA classes.” Classification of events according to RD-310, together with classification schemes in other jurisdictions, is depicted graphically in Appendix B.

6.0 RELATIVE EFFECT OF DOWNWIND DISTANCE ON AIRBORNE DOSES

6.1 Purpose of investigating dependence of ADF on downwind distance

The prediction of off-site dose consequences downwind is dependent on atmospheric dispersion factors (ADFs), which are a direct multiplier in dose calculations. ADFs are highly simplified representations of many complex phenomena. The calculation of ADFs requires a combination of modelling of the physical processes involved in atmospheric dispersion, derivation of suitable site meteorological parameters to be used in the models, and statistical analysis to determine dilution factors suitable for application. The dependence of ADFs on downwind distance is the primary determinant of how airborne dose varies with distance from the point of release, since for the dominant exposure pathways (immersion and inhalation), dose is directly proportional to the local concentration of airborne activity, which is determined by the dilution factor.

6.2 Method and limitations in calculating ADFs at near distances

Atmospheric dispersion factors used in the current work were derived using the METPROC calculation routine [28], which implements models specified in the Canadian standard CSA N288.2-M91 [29]. METPROC was developed by Ontario Hydro (OH) in the mid-1990s to calculate site-specific ADF’s for use in deterministic safety analysis [30], and has been used since that time as the primary tool for such analysis in support of regulatory submissions, [31] for example. Gaussian models for atmospheric dispersion, as described in CSA N288.2, are relatively standard and widely used for applications involving neutrally buoyant or buoyant releases. A comprehensive review [32] of modelling approaches, relevant to the atmospheric dispersion of accidental releases of radionuclides from Ontario Hydro’s nuclear generating stations for safety analysis applications was carried out in 1991, at around the same time that the current version of CSA N288.2-M91 was being finalized. The review discussed Gaussian models and data requirements for short-term, prolonged and long term releases, together with modifying factors such as for building wake effect and for the effect of a capping inversion associated with the Thermal Internal Boundary Layer. The report looked at the impact of using different dispersion models and stability determination schemes on calculating dilution factors, and went on to recommend a preferred method. Finally, dilution factors for use in safety analysis were derived using the recommended approach, based on best available site specific meteorological data.


The dilution factors derived in the case of Darlington were compared against the results of an earlier field study [33] conducted by OH Research Division during 1985 and 1986, involving the use of a sulphur hexafluoride tracer gas released from the vacuum building and powerhouse. The highest hourly dilution factor as measured at the station boundary 1520m east of the release point, was approximately 6 times smaller than the value calculated for safety analysis applications at 914 m. When the latter value was re-evaluated for a distance of 1520 m (comparable to the measurement), the observed value was 3 times smaller than the calculated dilution factor, indicating that the calculated value using the recommended approach was adequately conservative. Two further important aspects related to atmospheric dispersion at the Darlington site were identified by the tracer study: - broadening of the lateral plume was observed frequently, supporting the

implementation of the building wake effect in the dispersion models, - infrequent occurrence of the Thermal Internal Boundary Layer, indicating that a

correction for capping inversion is not required.

While the CSA standard CSA N288.2 provides methodology for Gaussian plume modelling with a range of applicability extending from 100m to 100km, it goes on to say that the values of the parameters in the models and equations have an inherent level of uncertainty. Appendix A states that:

"A1.2 For distances less than 1km under "ideal conditions" of uniform flat terrain, steady meteorology, etc, the prediction of the maximum downwind ground-level time-integrated concentration is only accurate to within 20% for a ground level release and to within 40% for an elevated release.

A1.5 When the model is applied for low wind speeds, less than 2m/s the plume meander and wind directional changes associated with these conditions can be grossly underestimated, thus overestimating the time integrated concentration."

In addition, the method for correcting for downwash and building wake effects is conservative, and when plume rise and buoyancy are factored in, then the accuracy of the model in the near field diminishes, as calculated results become increasingly conservative.

Traditionally, ADFs have been assessed for a site boundary distance of 0.914 km. It has been proposed that the site boundary for New Nuclear - Darlington may be located to as close as 0.5 km. Overall uncertainties associated with the calculation of the ADF at distances of approximately 1 km are considered to be about a factor of 2, on the basis of information such as that presented above. At distances less than 1 km, the uncertainties will likely be greater than a factor of 2, although it is considered that error in the predicted dilution factors will be in the conservative direction. The inherent uncertainty is a limitation of dispersion modelling, and reflects the sensitivity to assumptions concerning the geometry of the release point and surrounding area, among other factors as discussed above.


6.3 Ratios of ADFs obtained, applicability and interpretation

ADFs for the current study have been obtained with METPROC, using the same general approach and assumptions as have been applied in calculating ADFs for DNGS-A safety analysis applications. This includes the specification, for building wake effect purposes, of a cross sectional area consistent with the existing DNGS-A plant. The range of distances for which ADFs are calculated has been extended inward to a distance of 500 m from the release point.

An indication of the change in overall dose with downwind distance for a given release scenario can be seen from the variation with distance in each of the contributing ADFs. Appendix A, Table A1 presents Darlington site specific ADFs as a function of downwind distance ranging from 0.5 km to 5.0 km for three mutually exclusive atmospheric dispersion modelling regimes, which reflect the meteorological averaging period over which the release continues:

a) short term - releases occurring over a period less than about 1 hour, b) prolonged - releases over a period of about 1 day (excluding the short term

portion), and c) long term - releases extending beyond a day (excluding the prolonged and

short term portions).

In deterministic dose consequence analysis, the short term ADF is applied to the hour during the entire release sequence with the highest activity release (which is typically, though not always, the first hour of an extended release). Similarly, the prolonged term ADF is applied to the “worst” continuous 24 hour period of release, excluding the worst hour. The long term ADF is applied to the remainder of the release sequence. The overall dose consequence can therefore be considered to comprise three separate chronological components, where each respective ADF is included as a factor over the applicable period.

ADFs as a function of downwind distance, normalized to that of the historic distance of 0.914 km, provide the relative difference in dose consequence for the portion of the release occurring over that respective time period. ADF data normalized to 0.914 km for short term, prolonged term, and long term can be found in Appendix A, Figure A1. In comparing results presented in Appendix A for a downwind distance of 0.5 km relative to 0.914 km, it can be seen that the ADFs generally increases by a factor of ~2 for short and prolonged term releases, and by a factor of ~2.6 for long term releases.

The overall effect on airborne doses, when looking at a downwind distance of 0.5 km versus the traditional exclusion zone radius of 0.914 km, depends to some extent on the contribution to dose from each of the release periods, which is a function of the specific accident scenario and the time profile of releases. However, a general indication of the effect on dose of distance to the site boundary can be inferred from the band within which the three curves shown in Figure A1 are seen to vary. ADFs at a downwind distance of 0.5 km are higher than at 0.914 km by a factor of between 2 and 3. Therefore, an available margin of between 2 and 3 (excluding any allowance


for uncertainty in the ADF calculation) would be needed in dose results at 0.914 km relative to acceptance criteria, in order for doses at 0.5 km to meet the same criteria for airborne emissions. For a site boundary distance intermediate between 0.5 km and 0.914 km, the required margin would be progressively less, as can be seen from the band of curves shown in Figure A1 in Appendix A.

7.0 SOURCE TERM INPUTS FOR DESIGN BASIS EVENTS

7.1 Inputs required for dose calculation

In determining the dose consequence of an airborne release of radioactivity, three distinct types of input information are required – characterization of the radioactivity release, meteorological conditions which govern the dispersion of the airborne material, and dosimetric data which describes the effect of a given concentration of a radionuclide in terms of the resulting dose to an individual. Dosimetric information comprises primarily dose coefficients or conversion factors (DCFs) for exposure pathways including immersion (cloudshine), groundshine, inhalation and potentially ingestion. Breathing rates and intake rates are required respectively in conjunction with the latter two internal pathways. Essentially all required dosimetric data is available from well documented compilations, although there are differences from one national or international standard to another. Meteorological information and the atmospheric dilution factors (ADFs) derived from those weather conditions have already been discussed in Section 6. Clearly, this type of information is site specific. In design applications, stylized or hypothetical bounding ADFs are sometimes applied to provide an upper bound on dose consequences, which can then be tailored at a later stage to reflect characteristics of a specific site. Release data needed consists of a detailed breakdown both by radionuclide and by time, which describes the time-wise release rate to the atmosphere over the duration of the release. In cases where there are multiple release pathways or mechanisms, potentially with different characteristics, each of these needs to be separately specified. Releases may extend typically over a period of time ranging from 0 to 720 hours for a DBA.

The primary input required to the dose consequence evaluation contemplated for this study therefore comprises the detailed ex-plant source term calculated in the safety analysis carried out for AOOs and DBAs for each of the three candidate reactor types. With only a few limited exceptions, however, this type of information at the level of detail required was found not to be available from public domain sources, or directly from vendors at this stage. In lieu of that raw data, recourse was made where possible to already calculated dose consequences for relevant events for each of the candidate reactor types, as prepared for a variety of licensing, design certification, and screening type applications. Most of those safety cases have however been submitted to agencies operating under somewhat different safety and licensing frameworks than those applicable to new nuclear build in Canada, as prescribed by the CNSC regulatory documents RD-310, RD-337 and RD-346. It is necessary therefore to translate between the different jurisdictions, in terms of event characterization, safety analysis rules and practices, as well as acceptability criteria.


7.2 Event categorization, frequency and acceptance criterion

In the “modern” Canadian licensing context, as noted in Section 5.0, events are classified according to probabilistic studies and engineering judgement into classifications based upon the frequency of occurrence. These classes correlate directly with the acceptance criterion for radiological consequences. To aid in making equivalences against categorization approaches under other jurisdiction, a slightly more complete description of the events expected to fall within each category is reproduced here from an earlier draft version of RD-337, Design of New Nuclear Power Plants, released in October 2007 by the Canadian Nuclear Safety Commission for internal review and stakeholder consultation, which in section 7.2.2 describes the three classes of events as follows; - Anticipated Operational Occurrences (AOOs): An operational process deviating

from normal operation that is expected to occur once or several times during the operating lifetime of the NPP but which, in view of the appropriate design provisions, does not cause any significant damage to items important to safety nor lead to accident conditions. An AOO plant state includes all events with frequencies of occurrence equal to or greater than 10-2 per reactor year.

- Design Basis Accidents (DBAs): Accident conditions against which an NPP is

designed according to established design criteria, and for which the damage to the fuel and the release of radioactive material are kept within authorized limit. Design Basis Accidents (DBAs) include events with frequencies of occurrence equal to or greater than 10-5 per reactor year but less than 10-2 per reactor year.

- Beyond Design Basis Accidents (BDBAs): Accident conditions less frequent and

more severe than a design basis accident. A BDBA may or may not involve core degradation. Beyond Design Basis Accidents include events with frequencies of occurrence less than 10-5 per reactor year.

Of the three event classes only AOOs and DBAs have defined dose consequence requirements which must be met for the most limiting event.

Figure B1 in Appendix B presents the RD-310 event categorization in parallel with the event categorizations adopted in other licensing jurisdictions from which information on accident source terms and doses has been utilized.

7.3 US NRC Deterministic Safety Analysis Requirements

The US regulations deal with accident scenarios by the identification of an exclusion and a low population zone. The exclusion zone is defined as being large enough such that a person standing on its boundary for the two hours immediately following the onset of the release would not receive a dose in excess of 250 mSv (whole body) or 3000 mSv (thyroid) [19]. The low population zone is defined such that an individual standing on its boundary would not exceed the above doses if they remained there for the entire period of the passage of the plume.


The NRC’s traditional methods for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and selected chapters from the Standard Review Plan (SRP), NUREG-0800 [20] that the NRC staff uses when reviewing applications for the construction and operation of new nuclear power plants. The guidance found in these documents was developed to be consistent with the TID-14844 source term [21] and the whole body and thyroid dose guidelines stated in 10 CFR 100.11 [19].

Since the publication of TID-14844, significant advances have been made in understanding the timing, magnitude and chemical form of fission product releases from severe nuclear power plant accidents. This research has resulted in the publication of new estimates for the accident source terms that are more physically based and that could be applied to the design of future light water power reactors. The NRC staff considered the applicability of these Alternate Source Terms (AST) to operating reactors and determined that the current analytical approach based on the TID-14844 source term would continue to be adequate to protect public health and safety. Consequently, operating reactors licensed under that approach would not be required to re-analyze accidents using the AST. However, the NRC staff also determined that some licensees might wish to use an AST in analyses to support cost-beneficial licensing actions. Consequently, actions were initiated by the NRC to provide a regulatory basis for operating reactors to use an AST in design basis analyses. These initiatives resulted in the development and issuance of 10 CFR 50.67 [22] and a new regulatory guide [23].

7.3.1 Accident Source Terms (10 CFR 50.67)

The NRC may issue or amend an operating license only if the applicant's analysis demonstrates with reasonable assurance that:

1. An individual located at any point on the boundary of the exclusion area boundary (EAB) for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 250 mSv total effective dose equivalent (TEDE).

2. An individual located at any point on the outer boundary of the low population zone (LPZ), who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 250 mSv total effective dose equivalent (TEDE).

3. Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 mSv total effective dose equivalent (TEDE) for the duration of the accident.

The NRC staff has determined that there is an implied synergy between the Alternate Source Terms and total effective (TEDE) criteria and between the TID-14844 source terms and the whole body and thyroid dose criteria (as listed in 10 CFR 100.11), and therefore they do not expect to allow the TEDE criteria to be used with TID-14844 calculated results.


7.3.2 Accident Dose Criteria under Alternative Radiological Source Terms for Different Accident Types

When using the Alternate Source Term, the radiological criteria for the EAB, the outer boundary of the LPZ and for the control room are listed in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation (e.g. a large break LOCA). The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 7.1.

Table 7.1: Accident Dose Criteria in 10CFR50.67

Accident or Case EAB and LPZ Dose Criteria (TEDE)

Analysis Release Duration

LOCA 250 mSv 30 days for containment, ECCS, and MSIV (BWR) leakage

BWR Main Steam Line Break Instantaneous puff Fuel Damage or Pre-incident Spike

250 mSv

Equilibrium Iodine Activity 25 mSv

BWR Rod Drop Accident 63 mSv 24 hours

PWR Steam Generator Tube Rupture Affected SG: time to isolate; Unaffected SG(s): until cold shutdown is established

Fuel Damage or Pre-incident Spike

250 mSv

Coincident Iodine Spike 25 mSv

PWR Main Steam Line Break Until cold shutdown is established Fuel Damage or Pre-incident Spike

250 mSv

Coincident Iodine Spike 25 mSv

PWR Locked Rotor Accident 25 mSv Until cold shutdown is established

PWR Rod Ejection Accident 63 mSv 30 days for containment pathway; until cold shutdown is established for secondary pathway

Fuel Handling Accident 63 mSv 2 hours

7.4 Available sources of useful data

Review of vendor documentation and information in the public domain yielded a number of documents which include relevant supporting documentation. These documents include:

Canadian Regulatory Documents - CNSC RD-310, Safety Analysis for Nuclear Power Plants [3] - CNSC RD-337, Design of New Nuclear Power Plants [4] - CNSC RD-346, Site Evaluation for New Nuclear Power Plants [5] ACR-1000 Documents


- Vendor letters providing specific information to support new nuclear site licensing and environmental assessment

- ACR-1000 Assessment Documents EPR Documents - EDF/AREVA UK EPR – Generic Design Assessment

- Volume 1: Overview - Volume 2: Design and Safety - Volume 3: Environmental Impact

- US EPR Final Safety Analysis Report (FSAR) - UK Health and Safety Executive (HSE) and Environmental Agency Assessment

Reports of the Electricité De France (EDF) and Areva United Kingdom European Pressurised Reactor (UK EPR) submission.

AP1000 Documents - UK AP1000 Generic Design Assessment

- UK Compliance Document for AP1000 Design - UK AP1000 Safety, Security and Evaluation Report - UK AP1000 Probabilistic Risk Assessment

- NRC AP1000 Final Safety Evaluation Report (FSER) - AP1000 Design Control Document (DCD) Rev. 16 (Amended) - AP1000 Design Control Document (DCD) Rev. 15 (NRC approved) - Exelon Generation Company (EGC), LLC, Clinton Early Site Permit Application - Southern Nuclear Operating Company Vogtle Early Site Permit Application - UK Health and Safety Executive (HSE) and Environmental Agency Assessment

Reports of the Westinghouse Electric Company LLC AP1000 design

8.0 DOSE CONSEQUENCE EVALUATION FOR CANDIDATE REACTOR TYPES

The following subsections provide evidence, and in some instances deterministic results, to support the acceptability of locating the site boundary for New Nuclear – Darlington at a distance closer than the traditional exclusion zone radius of 914 m. Vendors of candidate reactor types have indicated that the plant designs will allow an exclusion area boundary in some cases as close as 500 m [6]. The objective here then is to make use of those supporting analyses or assessments to determine the extent to which the dose acceptance criteria specified in RD-337 for both AOOs and DBAs are likely to be satisfied over such a range of distances for the site boundary.

8.1 AECL ACR-1000

AECL [9] has adopted a 500 m exclusion zone distance in its ACR-1000 dose assessments. This value was chosen based upon the needs of international clients, and has now become the ACR-1000 standard. It is therefore to be expected that the available information will demonstrate that the RD-337 dose acceptance criteria can be satisfied at 500 m for an ACR-1000 plant located at the Darlington site.


8.1.1 Anticipated Operational Occurrences (AOOs)

The frequency classification for an Anticipated Operational Occurrence (AOO) is equal to or greater than 10-2 occ/yr. The vendor has noted that based on bounding events considered for ACR-1000, “AOO's are not expected to contribute any additional releases beyond the normal operation” [11]. This has been interpreted as meaning that releases and dose consequences from AOOs are low enough that they would not be distinguishable from the levels associated with normal operations. Annual dose due to airborne and waterborne emissions from ACR-1000 operation is estimated [12] for the bounding critical group to be < 2% of the 1mSv/y regulatory limit for members of the public. Based on an envelope of emissions [6] for all three candidate reactor types, an upper bound estimate [12] of annual dose for members of hypothetical groups residing in the immediate vicinity of the plant, 24 hours a day, 365 days a year, and consuming all food and water from local sources only, is less than 0.02 mSv. As such it is anticipated that dose consequences at the site boundary arising from any individual AOO for ACR-1000 will readily meet the criterion of 0.5 mSv per event specified in RD-337.

8.1.2 Design Basis Accidents (DBAs)

AECL has conducted dose assessment for a range of DBA cases. Two DBA events have been identified which are considered to give rise to the limiting DBA doses for ACR-1000. These events are the Steam Generator Tube Rupture and the End Fitting Failure [13]. All other DBA events, such as Pressure Tube/Calandria Tube Rupture and Feeder Breaks (off-stagnation), as well as moderator events, will have releases from containment, and consequently doses to the public, that are lower than the two DBA events for which releases and dose consequences have been specifically evaluated. In the analysis performed for both the Steam Generator Tube Rupture and the End Fitting Failure events, fuel failures were calculated to occur. In contrast, analysis performed for the large break LOCA event indicated that fuel failures were not calculated to occur. The Steam Generator Tube Rupture event further assumed no operator action to shut down the reactor for 8 hours after the event initiation.

Doses to members of the public due to the radioactivity release to the environment from each of the two DBA events analysed were calculated using a probabilistic approach with the ADDAM code [10]. The critical individual effective dose (95th percentile) at 500m was found to be 3.57 mSv for the Steam Generator Tube Rupture Event, and 2.56 mSv for the End Fitting Failure event [13]. These dose results at 500m are well within the RD-337 dose acceptance criterion of 20 mSv at the site boundary.

Even when considering BDBAs, such as the Large Loss of Coolant Accident with Loss of Long Term Cooling and the Stagnation Feeder Break, whose frequency falls well below the frequency cut-off for DBAs, dose consequence results at 500 m supplied by the vendor remain below the regulatory limit for DBAs of 20 mSv. The critical individual effective dose (95th percentile) at 500m was determined to be 13.8 mSv and 0.482 mSv for the two events, respectively [13].


Analyses performed have indicated that deterministic calculations using site-specific ADFs for the Darlington site provide dose results which are bounded by the probabilistic results at the 95th percentile, generated using ADDAM with weather input representative of a generic site. The conclusion from these studies is that the dose consequences for the limiting DBA events for ACR-1000, and even for specific BDBA events analysed, will meet the RD-337 dose acceptance criterion for DBA events at a site boundary distance of 500 m.

8.2 AREVA EPR

The term ‘EPR’, as previously used, refers to the EPR design in general. EPR projects targeted for different countries are based on the Nuclear Island design developed during the EPR Basic Design phase. However the design of the Turbine Island and Balance of Plant may differ somewhat from project to project.

EPR designs are supported by differing safety analysis approaches, consistent with the respective licensing jurisdictions. Distinct reference to the relevant safety analysis for either the UK EPR or the US EPR is specified within subsections 8.2.1 and 8.2.2.

From a dose consequence perspective, it should be noted that both the UK and US EPR containment designs share the same principles, i.e. no venting etc. The safety analysis for EPR submitted to the UK nuclear regulators for the Generic Design Assessment process is largely based on that already in place for the Flamanville EPR, and hence it can considered to be broadly applicable to the EPR design being considered for New Nuclear – Darlington. The major difference with respect to the US EPR is the analysis approach, in that analysis for European applications uses more realistic modelling than the prescriptive assumptions mandated by NRC licensing processes.


The US EPR defines AOOs, as per Appendix A to 10 CFR Part 50, as conditions of normal operation that are expected to occur one or more times during the life of the nuclear plant unit. The US NRC standard review plan (SRP) uses the term AOO to refer to the events that are categorized in US Regulatory Guide 1.70 and in US Regulatory Guide 1.206 as incidents of moderate frequency (i.e., events that are expected to occur several times during the plant’s lifetime) and infrequent events (i.e., events that may occur during the lifetime of the plant) [24]. Incidents of moderate frequency and infrequent events are also known as Condition II and Condition III events, respectively, in the commonly used, oft cited but unofficial American Nuclear Society (ANS) standards [25]. As shown in Appendix B, Figure B1, Condition II and Condition III events as per ANS meet the CNSC frequency classifications for AOOs. For the US EPR, the limiting AOO is identified as follows: “The inadvertent opening of a Main Steam Safety Value (MSSV) is the limiting AOO event for radiological consequences” [14]. The MSSVs provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurization of the reactor


coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the condenser, is not available. The purpose of the MSSVs is to limit the secondary system pressure to ≤ 110% of design pressure during an anticipated operational occurrence (AOO) or postulated accident. The dose acceptance criteria under the US EPR approach for such events are defined in 10 CFR Part 50, Appendix I, for the summation of radioactive releases during normal operation and the annual average radioactive releases due to an AOO event based on realistic assumptions. The RCS and secondary coolant concentrations correspond to normal operating conditions, and are determined through application of the ANSI/ANS-18.1-1999 standard, “Radioactive Source Term for Normal Operation of Light Water Reactors” [26].

The radiological consequences are determined at a distance of 805m (0.5 mi.) from the site. The analysis conservatively assumes that the entire ingestion pathway is located at this distance and the exposure is continuous during the entire event. Subsequent exposure continues for an entire year thereafter, accounting for submersion, inhalation, ingestion and ground-shine pathways.

The worst-case organ dose is determined to be to the infant thyroid and to amount to 7.9E-03 mSv, mostly due to the ingestion of milk [14].

It should be noted that for the above limiting AOO as defined for the US EPR, there are no fuel failures assumed.

A more severe incident is documented for the UK EPR, called Steam Generator Tube Rupture -1 tube, Plant Condition Category 3 (PCC-3). The radiological consequences of this accident arise from the release of activity to the environment by the main steam atmospheric dump valves of the failed steam generator; the activity is due to the contamination of the secondary system by the primary system via the broken SG tube [15]. The activity peak in the primary system (called the iodine peak, or iodine spike), caused by the transfer of activity from the fuel pellet/cladding gap into the primary fluid following an automatic reactor shutdown, is taken into account for assessing the radiological consequences of this accident. The effective dose to the most critical individual at 500m, 7 days following the initiating event, would be 0.2 mSv [15]. Although this dose result does not extend to the 30 day exposure duration specified in RD-337, further dose accumulation after 7 days is expected to represent a relatively small increment (see Section 8.2.2). It should be noted that the frequency of occurrence for a PCC-3 event is between 10-2/yr and 10-4/yr which is below the frequency cut-off for AOOs. Hence this event is less probable than what needs to be considered for an AOO in demonstrating compliance with RD-337.

Nonetheless, available results for limiting AOO events addressed under both US EPR and UK EPR licensing regimes suggest that for the appropriate events, the RD-337 criteria of 0.5 mSv for an AOO would be met at a site boundary distance as close as 500m.



Analysis for the US EPR uses a prescriptive approach, evaluated by applying the methodology in U.S. RG 1.183, for the purposes of assessing a LOCA, classified as a DBA. Given the nature of the U.S. regulatory environment, dose consequences tend to be higher than with ‘realistic’ assessments. As an example supporting the previous statement, the US EPR is licensed according to provisions under RG 1.183 known as “leak-before-break” methodology. However, the LOCA radiological consequences evaluation does not credit the leak-before-break gap release phase onset of 10 minutes, available (per RG 1.183, Section 3.3) for plants licensed with leak-before-break methodology. Analysis of dose consequences for the event were found to yield a total effective dose equivalent of 122 mSv at a distance of 805 m (0.5 mi), 30 days following the event [14]. Using the U.S. prescriptive analysis approach, doses estimated for the LOCA event meet the applicable criterion of 250 mSv given in 10CFR50.67, but do not meet the requirements of RD-337 for DBA doses. For the UK EPR, there are two sets of radiological consequence analyses which provide results which are relevant to the RD-337 criteria for DBA: consequence analysis for severe accidents, and consequence analysis for reference operating conditions studies [15]. For the purposes of this assessment, the UK EPR BDBA event, which includes 100% core melt, lies beyond the classification of a DBA within RD-310, as the frequency of occurrence is less than 10-6 occ/yr. However, this event can be used to infer limiting public doses for less severe DBA events of greater probability.

The analysis for assessing the dose consequences of a Severe Accident, referred to as risk reduction category B (RRC-B) which stipulates the prevention of large releases in core melt cases [15], is based on realistic calculations of physical phenomena rather than postulates. Two of the major differing principles between the US EPR and the UK EPR are in the assessment of radiological consequences associated with Severe Accidents. The UK EPR Severe Accident assumes;

- the evaluation of the activity released is based on a reference source term,

calculated with reasonably conservative bounding assumptions, independent of the accident scenario,

- the assumptions used in calculating radiological consequences (dose evaluation) are fixed realistically to give a reasonably conservative evaluation of the radiological consequences.

In addition to the above, the safety analysis approach for UK EPR reactors is deterministic, complemented by probabilistic analyses, based on the concept of defence in depth. Within the framework, a number of design provisions (depressurisation of the primary system, installation of recombiner units, core catcher, containment heat removal systems) are made to preserve the integrity of the containment in severe accidents and hence reduce the accident consequences [15].


Ultimately, the results of such analysis deviate from the U.S. prescriptive approach and provide a more realistic dose consequence for a Severe Accident. Figure 8.2-1, presents the effective short term dose as calculated for a typical EPR site (Flamanville in France) using the EPR reference source term for a Severe Accident. At the closest distance of 500 m, the effective dose to both adult and infant up to 7 days is within the dose acceptance criterion of 20 mSv given in RD-337 for a DBA. Although this dose estimate does not extend over the full 30 day period applicable to the RD-337 criterion, it is expected that the incremental dose up to 30 days would not substantially change conclusions using the effective short term dose given here. Leakage from containment is the primary release pathway, which is affiliated with containment pressure. A large-break LOCA presents the greatest challenge to containment pressurization, as it results in the fastest pressurization [15]. Following a 12 hour “grace period”, containment heat removal is credited which is capable of reducing containment pressure from 5 bars (500 kPa) at 12 hours following grace period, to 1.3 bars (130 kPa) after 7 days, reducing the leak rate. Based on this consideration, together with radioactive decay of the source term and the effect of natural removal processes such as deposition and sedimentation within containment, it is anticipated that the majority of the dose is obtained within the first 7 days.

Figure 8.2-1: Effective Short Term (7 day) Dose In The case of a Severe Accident (adapted from [15])

The second set of information relevant to the EPR, with respect to the requirements of RD-337 for DBA, comprises consequences evaluated for a range of accidents with a single initiating event, spanning PCC categories 2, 3 and 4. The Large Break LOCA classified by the vendor as PCC-4 (frequency of occurrence less than 10-4 occ/yr and greater than 10-6 occ/yr) was found to yield an effective dose consequence of 0.29 mSv at 500 m, up to 7 days following the event. However, the most limiting


PCC-4 event was found to be a fuel handling accident, for which the 7 day effective dose at 500 m was determined to be 5.5 mSv [15].

The results quoted here for the UK EPR thus suggest that the dose acceptance criterion for DBA of 20 mSv over the first 30 days following the event is attainable at a distance of 500 m, based on the characteristics of the site (Flamanville) for which the dose calculations were carried out. While it is not possible to directly translate these dose results to equivalent distance-dependent numbers for the Darlington site (since there is no source term info accompanying the doses), it is believed that the atmospheric dispersion factor conditions assumed in the UK EPR analyses do not differ greatly from those established for the Darlington site. This has been inferred through the comparison of Darlington NGS long term ADFs with published ADFs for normal operating releases applicable at local population centres in close proximity to EDF’s Flamanville EPR, as shown in Appendix A – Figure A2. Variations in ADFs at towns around Flamanville EPR are due to the relative location of each town with respect to wind sectors (or direction downwind from the station). In general, the Flamanville location-specific ADFs are within approximately a factor of 3 of Darlington values at comparable distances. Since Flamanville is in a coastal region, these ADFs tend to be lower on average than Darlington long term site specific ADFs. However, the difference in ADFs at Darlington relative to Flamanville is smaller than the margin available between the dose of 5.5 mSv for the most limiting PCC-4 event as calculated for Flamanville versus the RD-337 DBA dose limit of 20 mSv. With this consideration, it is expected that for relevant EPR design basis accidents, the RD-337 criteria for DBA will be met at an exclusion zone distance as close as 500 m at Darlington.

8.3 Westinghouse AP1000


The AP1000 Condition II and Condition III events as per ANSI/ANS-51.1 standard [25] are considered to be AOOs within the requirements of RD-310 based on their event frequency. The frequency of occurrence for these conditions can be found in Appendix B.

In accordance with the vendor specifications, the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The reactor is designed and performance and safety criteria requirements are established for the thermal and hydraulic design of the fuel such that;

� Fuel damage (defined as penetration of the fission product barrier, that is, the fuel rod cladding) is not expected during normal operation and operational transients (Condition I) or any transient conditions arising from faults of moderate frequency (Condition II). It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the plant cleanup system and are consistent with the plant design bases [18].


� The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged (as defined per the above description). The fraction of fuel rods damaged must be limited to meet the dose guidelines of 10 CFR 100 although sufficient fuel damage might occur to preclude immediate resumption of operation [18].

In addition, reactor core safety limits require that specified acceptable fuel design limits not be exceeded for Condition I and II events. This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur, and by that the fuel centreline temperature stays below the melting temperature. Based on the arguments above, the dose consequences for an AOO (Condition II and III events) are strongly dependent on the activity within the coolant prior to an event. Thus operating procedures, for normal operations, will ensure that in the event of an AOO the activity of the coolant will be within limits such that the dose consequences from a postulated event meet the criteria of RD-337.


In this section, a multipath approach is taken to rationalize the DBA for the AP1000. First the UK Health and Safety Executive (HSE) disposition of a DBA will be presented, then the discussion will proceed in explaining the difference between the DCD and the FSER, and finally a comparison between doses for the AP1000 and US EPR is presented.

The AP1000 Final Safety Evaluation Report (FSER) [17] presents the following design basis accidents which include the LOCA, main steam line break outside containment, reactor coolant pump seizure, rod ejection accident, fuel-handling accident, small line break, steam generator tube rupture, and spent fuel pool boiling.

Small line breaks, typically classified as small or medium break LOCAs, have been noted in the UK HSE compliance documentation [16] to have a frequency of occurrence of 10-3 to 10-4 per reactor year. Analysis results reported in DCD subsection 15.6.5.4B.4, states that no significant fuel damage is expected to occur for a postulated small or medium LOCA. The FSER calculates the Total Effective Dose Equivalent (TEDE) for such an event to be 10 mSv at the Exclusion Area Boundary (EAB), at a distance of 805 m for a 2-hour exposure duration.

In addition to small line breaks, the main steam line break, reactor coolant pump seizure, rod ejection accident, steam generator tube rupture and spent fuel pool boiling DBAs meet the DBA dose criterion in RD-337 at the AP1000 EAB (805m for a 2-hour exposure duration). The AP1000 Final Safety Evaluation Report [17] presents Total Effective Dose Equivalent values for design basis accidents at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ), as shown in Table 8.3-1.


Table 8.3-1: AP1000 Final Safety Evaluation Report TEDE for DBAs [17]

Postulated Accident EAB* LPZ**

Loss of coolant accident 190 mSv 150 mSv

Main steamline break outside containment

-With accident-initiated iodine spike 2 mSv 8 mSv

-With preaccident iodine spike <1 mSv 1 mSv

Reactor coolant pump shaft seizure

-With feedwater available <1 mSv <1 mSv

-Without feedwater available <1 mSv <1 mSv

Rod ejection accident 15 mSv 24 mSv Fuel-handling accident 24 mSv 10 mSv Small line break accident 10 mSv 4 mSv Steam generator tube rupture -With accident-initiated iodine spike 5 mSv 7 mSv -With preaccident iodine spike 10 mSv 6 mSv Spent fuel pool boiling n/a <0.1 mSv

*EAB is located 0.805 km (0.5 mi.) and is for a 2 hour exposure period **LPZ is located 2.41 km (1.5 mi.)

Of these events only the LOCA and fuel handling accident exceed the dose criterion in RD-337 at the AP1000 EAB, with TEDE of 190 mSv and 24 mSv respectively. These have been addressed separately in section 8.3.2.1 Fuel handling accident, and section 8.3.2.3 Large LOCA.

8.3.2.1 Fuel Handling Accident

The fuel handling accident is defined as the dropping of a spent fuel assembly such that every rod in the dropped assembly has its cladding breached so that the activity in the fuel/cladding gap is released [18]. The calculated offsite dose reported in the FSER exceeds the 20 mSv dose criterion in RD-337, with a TEDE of 24 mSv (at 805m for a 2-hour exposure duration). This dose analysis uses the NRC prescriptive methodology. It is expected that calculated offsite dose can be reduced to less than the target value by departing from the NRC’s prescribed assumptions, but still retaining the conservative nature of the analysis. Section 8.3.2.3, further elaborates on the impact of departure from the NRC prescribed approach. A further reduction in dose consequence for all events aforementioned is also expected through the application of site specific ADFs. AP1000 Generic Site ADFs given in Table 8.3-2 can be compared with those of Darlington site specific ADFs. The generic ADF at the EAB, also referred to as a bounding hypothetical ADF, when compared to that of Darlington at a distance of 500 m was found to be 3 times greater. Thus, dose consequence for a fuel handling accident which is analyzed for a 2-hour release period (short term) is expected to be 1/3 of that presented in Table 8.3-1 when assessed for a distance of 500 m at Darlington.


Given this basis, it is expected that all DBAs (with the exception of the LOCA) will probably meet the DBA dose criterion in RD-337 for Darlington at a site boundary distance as close as 500m.

Table 8.3-2: AP1000 Generic Site ADFs [17]

AP1000 Period (Zone)

Associated Release Period

AP1000 Generic ADF (s/m

3)

Darlington Site Specific ADFs at 500 m (s/m

3)

Ratio of Darlington Site Specific ADFs to AP1000 Generic ADFs

0 to 2 hours (EAB*) Short Term 5.10E-04 1.64E-04 0.32

0 to 8 hours (LPZ**) Prolonged Term 2.20E-04 1.27E-05 0.06

8 to 24 hours (LPZ**) Prolonged Term 1.60E-04 1.27E-05 0.08

1 to 4 days (LPZ**) Long Term 1.00E-04 6.41E-06 0.06

4 to 30 days (LPZ**) Long Term 8.00E-05 6.41E-06 0.08

*EAB is located 0.805 km (0.5 mi.) **LPZ is located 2.41 km (1.5 mi.)

8.3.2.2 Large LOCA

The identified event frequency for the large LOCA is 5.04x10-6 occ/yr [16], and therefore is considered to be a BDBA which is not required to be assessed against the quantitative dose acceptance criteria in RD-337. However, given the potential significance of LOCA events, this aspect has been addressed further in section 8.3.2.3.

8.3.2.3 AP1000 and UK EPR Rationale

At present there is no available analysis which uses the less prescriptive approach to safety analysis of design basis events for the AP1000. A comparison of dose assessment results for DBAs, as calculated using applicable NRC analysis rules and practices, for the AP1000 with equivalent information for the US EPR was undertaken and has been presented in Appendix C. The general conclusion that can be drawn from comparing Tables C1 and C2 is that for equivalent events with equivalent assumptions, AP1000 dose consequences are comparable to the corresponding results for US EPR. Thus it can be surmised that safety analysis for AP1000, when performed under the non-prescriptive licensing regime applicable in Europe (as described in Section 8.2.2), would yield dose consequences that are not dissimilar to those already evaluated for the EPR under such circumstances. Based on this supposition, and the rationale previously mentioned in section 8.3.2.1 pertaining to site specific ADFs, it is expected that the dose consequences for a LLOCA for an AP1000 will probably meet the DBA dose acceptance criteria in RD-337 at similar distances as for EPR.

9.0 CONCLUSIONS

This report presents the underlying considerations influencing the determination of site boundary distance for New Nuclear - Darlington, based on updated Canadian regulatory requirements. The primary requirements to be satisfied are identified as the dose acceptance criteria specified in RD-337 [4] for Design Basis Events, i.e. AOOs


and DBAs. The variation of atmospheric dispersion factor with downwind distance represents the primary factor which influences how doses evaluated at the site boundary vary with assumed distance between the release point and the site boundary. However, quantitative evaluation of doses as a function of distance requires detailed source term information for the requisite events, describing the breakdown both by radionuclide and by time of the release rate to atmosphere over the duration of the event.

Recognizing that insufficient source term data is currently provided by vendors, in most cases, to enable site-specific dose versus distance profiles to be evaluated on a fully quantitative basis, the available information on releases and doses as presented in a variety of licensing contexts has been assembled, for those scenarios most closely corresponding to the design basis event categories considered under RD-310 [3]. Analysis results for limiting AOOs and DBAs are reviewed for each of the three candidate reactor types in turn. Findings with respect to expectations for the site boundary distance in each case are summarized in Table 9.0-1 for AOOs and in Table 9.0-2 for DBAs, with cross-reference to the section of the report which presents the detailed discussion for the respective reactor type.

In most cases, the available safety analysis supports the contention that the dose acceptance criteria would be met for a site boundary distance at New Nuclear – Darlington as close as 500 m from the containment. In other cases, while not necessarily providing direct confirmation, the available information provides a reasonable indication that the dose acceptance criteria would be met at such a distance, if safety analysis were carried out using methods and assumptions aligned with the licensing approach applicable in Canada.


Table 9.0-1: RD-310 Event Classification: AOO (f ≥10-2)

Reactor Type Meets RD-337 AOO criterion (0.5 mSv) Basis Synopsis/Considerations

ACR-1000

At 500 m

Refer to section 8.1.1

AOO releases are interpreted to be indistinguishable from normal operations. It is therefore anticipated that dose consequences will readily meet the AOO requirements of RD-337.

EPR

At 500 m


Dose consequence results for the most limiting US EPR AOO meet the requirements of RD-337. Dose consequence analysis carried out for a more severe UK EPR accident, bounding AOOs within the context of RD-310, also meet the AOO requirements of RD-337.

AP1000

Highly likely at 500 m


Dose consequences for AP1000 AOOs are found to be primarily dependant on operational/design limits for activity in the primary coolant. It is anticipated that these limits can be controlled such that the requirements of RD-337 are met for any given AOO.

Table 9.0-2: RD-310 Event Classification: DBA (10-2 > f ≥ 10-5)

Reactor Type Meets RD-337 DBA criterion (20.0 mSv) Basis Synopsis/Considerations

ACR-1000

At 500 m


PSA dose consequence results, supported by deterministic calculations applying DN site specific ADFs, meet the DBA requirements of RD-337.

EPR

At 500 m


Dose consequence results for a Severe Accident (BDBA) as well as for DBAs, based on a "European" licensing approach largely meet the DBA requirements of RD-337.

AP1000

Probably at 500 m


Dose consequence results for DBAs (with the exception of LOCA which falls outside the RD-310 frequency cutoff for DBA) are expected to closely approach DBA requirements of RD-337, when considered in conjunction with site specific ADFs for DN site. It is anticipated that if safety analyses for AP1000, particularly for LLOCA were performed using a less prescriptive approach, then as demonstrated for EPR, it could be shown that DBA requirements of RD-337 would be met.


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14. Areva NP Inc., “U.S. EPR Final Safety Analysis Report”, Document Rev. 0, 2007.

15. Areva NP & EDF, “UK EPR Generic Design Assessment - Fundamental Safety Overview”, 2007.


16. Westinghouse Electric Company LLC, “UK Compliance Document for AP1000 Design”, Doc. # UKP-GW-GL-710 Rev. 0, May 2007.

17. Division of Regulatory Improvement Programs, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission (NRC), “Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design (NUREG-1793)”, Doc. NUREG-1793, September 2004.

18. Westinghouse Electric Company LLC, “AP1000 Design Control Document”, Document Rev. 16 (Public Version), 2007.

19. Nuclear Regulatory Commission, Part 100 Reactor Site Criteria, Sec. 100.11 Determination of exclusion area, low population zone, and population center distance, 10CFR100.11, 2002.

20. USNRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, September 1981 (or updates of specific sections).

21. J.J. DiNunno et al., Calculation of Distance Factors for Power and Test Reactor Sites, USAEC TID-14844, U.S. Atomic Energy Commission (now USNRC), 1962.

22. Nuclear Regulatory Commission, Part 50, Domestic Licensing of Production and Utilization Facilities, Sec 50.67 Accident Source Term, 10CFR50.67.

23. USNRC, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Regulatory Guide 1.183, July 2000.

24. NUREG-0800, “U.S. NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” NRC, March 2007.

25. American Nuclear Society “ANSI/ANS 51.1-R1988: Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants”, 1983.

26. American Nuclear Society “ANSI/ANS-18.1-1999: Radioactive Source Term for Normal Operation for Light Water Reactors”, 1999.

27. Electricité de France, “European Pressurized Water Reactor – Rapport Préliminaire de Sûreté de Flamanville 3”, 2006.

28. Ontario Hydro Nuclear. Quality Assurance Documentation for METPROC v1.0. File no.: N-906-03611.1-P-RSOAD-METPROC-CC Doc/Log-1.0, May 1997.

29. CSA N288.2-M91-CAN/CSA, “Guidelines for Calculating Radiation doses to the Public from a Release of Airborne Radioactive Material under Hypothetical Accident Conditions in Nuclear Reactors”, April 1991.

30. Oliverio, M. “Updated Site Specific Atmospheric dilution Factors for Use in Safety Analysis”, Nuclear Technology Services Report No. N-03611.1-965074 R0, October 1996.


31. Peabody, W.F. Letter to B.R. LeBlanc, “Calculation of Atmospheric Dilution Factors for Safety Analysis Dose Calculations from Site Specific Meteorological Data”, OH File N-03485.1 P, PRAD/N-000531 P, 5 June 1998.

32. Lam, L. “Modelling of the Atmospheric Dispersion of Accidental Releases of Radionuclides from Ontario Hydro’s Nuclear Generating Stations for Safety Analysis Applications”, OH Design & Development Report No. 91004, June 1991.

33. Ogram, G.L. et al, “Darlington Tracer Study 1985-1986: Data Report”, Ontario Hydro Research Division Report C89-10-K.


GLOSSARY

ACR Advanced CANDU Reactor

ADF Atmospheric Dilution Factor or (equivalently) Atmospheric Dispersion Factor

AECL Atomic Energy of Canada Limited

ANS American Nuclear Society

ANSI American National Standards Institute

AOO Anticipated Operational Occurrence

AST Alternative Source Term

BDBA Beyond Design Basis Accident

CANDU Canadian Deuterium Uranium

CFR Code of Federal Regulations

CNSC Canadian Nuclear Safety Commission

CSA Canadian Standards Association

DBA Design Basis Accident

DCD Design Control Document

DCF Dose Conversion Factor

DN Darlington Nuclear

DNB Departure from Nucleate Boiling

DNGS Darlington Nuclear Generating Station

EAB Exclusion Area Boundary

ECCS Emergency Core Cooling System

EDF Electricité De France

EGC Exelon Generation Company

EPR Areva’s EPR Reactor

FSAR Final Safety Analysis Report

FSER Final Safety Evaluation Report

GSCR Generic Safety Case Report

HSE Health and Safety Executive

ICRP International Commission on Radiological Protection

LCDA Limited Core Damage Accident

LLC Limited Liability Company

LLOCA Large Loss of Coolant Accident

LOCA Loss of Coolant Accident


LOE Limit of Operating Envelope

LOECC Loss of Emergency Core Cooling

LOLTC Loss of Long Term Cooling

LPZ Low Population Zone

LTC Long Term Cooling

MSLB Main Steam Line Break

MSSV Main Steam Safety Valve

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

OH Ontario Hydro

PCC Plant Condition Category

PSA Probabilistic Safety Assessment

PWR Pressurized Water Reactor

RCP Reactor Coolant Pump

RCPB Reactor Coolant Pressure Boundary

RCS Reactor Coolant System

RD Regulatory Document

RG Regulatory Guide

RIH Reactor Inlet Header

ROH Reactor Outlet Header

RRC Risk Reduction Category

RWT Reserve Water Tank

SFB Stagnation Feeder Break

SG Steam Generator

SGTR Steam Generator Tube Rupture

SRP Standard Review Plan

TEDE Total Effective Dose Equivalent


APPENDIX A: DARLINGTON NUCLEAR SITE SPECIFIC ADFS

Table A1: METPROC Results for Atmospheric Dilution Factor Using 1997-2000 Meteorological Data for Darlington

Distance (km)

Release Period Individual ADF (x 10-6 s/m3)

Short Term 164

Prolonged Term 12.7 0.5

Long Term 6.41

Short Term 123 Prolonged Term 9.35 0.65

Long Term 4.22

Short Term 97


Long Term 3.04

Short Term 83


Long Term 2.48

Short Term 75


Long Term 2.16

Short Term 57


Long Term 1.52

Short Term 45


Long Term 1.13

Short Term 37


Long Term 0.877

Short Term 31

Prolonged Term 2.13 2

Long Term 0.708


Long Term 0.493


Long Term 0.285

Short Term 10 Prolonged Term 0.729 5

Long Term 0.160

AM

EC N

SS L

imited

P0977/R

P/0

02 R

03

P

age 3

4 o

f 37

Form

114 R

14

0.0

0

0.5

0

1.0

0

1.5

0

2.0

0

2.5

0

3.0

0

0.5

11.5

22.5

33.5

44.5

5

Dis

tan

ce (

km

)

ADF Relative to 0.914km

Short

Term

- O

ne h

our

rele

ase p

eriod

Pro

longed T

erm

- 2

4 h

our

rele

ase p

eriod

exclu

din

g the s

hort

term

period

Long T

erm

- R

ele

ase p

eriod b

eyo

nd o

ne

day

exclu

din

g s

hort

and p

rolo

nged term

periods

Figure A1: Darlington Atm

ospheric Dilution Factors Relative to 0.914 km Using 1997-2000 Meteorological Data

AM

EC N

SS L

imited

P0977/R

P/0

02 R

03

P

age 3

5 o

f 37

Form

114 R

14

0.0

0E

+0

0

1.0

0E

-06

2.0

0E

-06

3.0

0E

-06

4.0

0E

-06

5.0

0E

-06

6.0

0E

-06

0.6

0.7

0.8

0.9

11.1

1.2

1.3

1.4

1.5

Dis

tan

ce

(k

m)

ADF (s/m3)

DN

GS

Site S

pecfic L

ong

Term

AD

F(1

997

-2000

ME

T D

ata

)

ED

F E

PR

- L

es H

oug

ues

ED

F E

PR

- L

a B

erq

ueri

e

ED

F E

PR

- H

am

ea

u G

uera

rd

ED

F E

PR

- L

a C

oq

uais

e

ED

F E

PR

- D

ielle

te

ED

F E

PR

- F

lam

anvill

e

Figure A2: Atm

ospheric Dispersion Factors (ADFs) for EDF Flamanville EPR local cities and DNGS Site Specific Long Term

ADFs [27]

AM

EC N

SS L

imited

P0977/R

P/0

02 R

03

P

age 3

6 o

f 37

Form

114 R

14

APPENDIX B: EVENT CLASSIFICATION

Figure B1: Event Classification for CNSC and Candidate Reactors [Yu, L., OPG 2008]


APPENDIX C: US EPR AND AP1000 DESIGN BASIS ACCIDENTS

Table C1: US EPR Final Safety Analysis Report [14]

Offsite Dose

Design Basis Accident EAB* LPZ**

LOCA 122 mSv 111 mSv

Small line break outside of reactor building 18 mSv 3.0 mSv

Pre-incident spike 11 mSv 3.0 mSv

SGTR Coincident spike 7.0 mSv 5.0 mSv

Pre-incident spike 2.0 mSv 1.0 mSv

Coincident spike 3.0 mSv 2.0 mSv

Fuel rod clad failure 53 mSv 26 mSv

MSLB Fuel overheat 58 mSv 28 mSv

RCP locked rotor/broken shaft 23 mSv 9.0 mSv

Rod ejection 57 mSv 35 mSv

Fuel handling accident 56 mSv 10 mSv *EAB is located 0.805 km (0.5 mi.) and is for a 2 hour exposure period **LPZ is located 2.41 km (1.5 mi.)

Table C2: AP1000 Design Control Document Rev. 16 [18]

Offsite Dose

Design Basis Accident EAB* LPZ**

LOCA # 115 mSv 142 mSv Small Line carrying primary coolant break outside containment 21 mSv 11 mSv

accident initiated iodine spike 11 mSv 8.0 mSv

SGTR pre-existing iodine spike 22 mSv 13 mSv

accident initiated iodine spike 11 mSv 20 mSv

MSLB pre-existing iodine spike 10 mSv 8.0 mSv

Feedwater not available 8.0 mSv 4.0 mSv Locked Rotor Accident / RCP shaft break Feedwater Available 6.0 mSv 8.0 mSv

Rod Ejection 36 mSv 69 mSv

Fuel handling accident 52 mSv 26 mSv *EAB is located 0.805 km (0.5 mi.) and is for a 2 hour exposure period **LPZ is located 2.41 km (1.5 mi.) # Differences in the doses for LOCA versus those shown in Table 8.3-1 are due to update of the AP1000 DCD [18] to Rev. 16, relative to Rev.15 upon which the FSER [17] was based.

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