THIRD CASE STUDY - POSTCLOSURE SAFETY ASSESSMENT

Report No: 06819-REP-01200-10109-R00

March 2004

Key Words: Third Case Study, safety assessment, repository, used fuel

For further information contact: Ontario Power Generation, Nuclear Waste Management Division 700 University Avenue, Toronto, Ontario, Canada, M5G 1X6

Fax: (416) 592-7336

P. Gierszewski1 J. Avis2 N. Calder2 A. D’Andrea1 F. Garisto1 C. Kitson3 T. Melnyk3 K. Wei1 L. Wojciechowski3 1Ontario Power Generation 2Intera Engineering 3AECL

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ABSTRACT

Report Number: 06819-REP-01200-10109-R00 Title: THIRD CASE STUDY - POSTCLOSURE SAFETY ASSESSMENT Author(s): P. Gierszewski1, J. Avis2, N. Calder2, A. D’Andrea1, F. Garisto1,

C. Kitson3, T. Melnyk3, K. Wei1, L. Wojciechowski3 Company: 1Ontario Power Generation, 2Intera Engineering, 3AECL Date: March 2004 Abstract Disposal in a deep geologic repository is one approach for the long-term management of used CANDU fuel. Geological disposal isolates the wastes from people and the environment. Two major Canadian postclosure safety assessments of this concept have been previously completed – the Environmental Impact Statement case study in 1994 and the Second Case Study in 1996. The Third Case Study is a new postclosure safety assessment of the deep geologic repository concept for used CANDU fuel in the Canadian Shield. It differs from the two previous studies in that it considers a revised repository design and a new hypothetical site. The updated reference container has an outer copper shell for corrosion protection, an inner steel vessel for structural support, and an increased capacity to hold 324 used fuel bundles. The Third Case Study incorporates a geostatistical approach for the creation of a geosphere model. The approach represents a significant improvement in how geosphere uncertainty could be incorporated into the safety assessment process. The TCS geosphere is considered representative of a potential Canadian Shield setting, and not necessarily representative of the best (nor worst) features of a potential site. The study is not a comprehensive safety assessment. It focuses on three key scenarios - base, defective container (groundwater transport) and inadvertent human intrusion. The study evaluates radiological doses to humans. A full safety assessment in support of a real candidate site would consider other scenarios (such as glaciation), other impacts (such as effects on biota) and other safety arguments (such as the age of groundwater at the repository level). This study is based on information typical of that available from a potential site during the site evaluation stage but prior to exploratory drilling. This report provides an overview of the Third Case Study. A full description of the study and analyses are given here and in the supporting reports referenced here. We conclude that the Third Case Study repository design and hypothetical site would meet international postclosure dose rate guidelines. If the study was for a real candidate site, these results would provide technical support to proceed to exploratory drilling. It would also lead to further safety-related studies as part of a staged evaluation program, incorporating both the results of site studies and extending the breadth of analyses. The postclosure safety of the deep repository concept for disposal of Canada's used nuclear fuel has now been illustrated for three possible combinations of engineering design and hypothetical Canadian Shield sites. This increases confidence that, from a technical perspective, a suitable repository could be designed, sited and constructed within the Canadian Shield.

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CONTENTS

Page

ABSTRACT............................................................................................................................... iii

1. INTRODUCTION................................................................................................. 1 1.1 BACKGROUND.................................................................................................. 1 1.2 PURPOSE .......................................................................................................... 3 1.3 REPORT OUTLINE............................................................................................. 3

2. ASSESSMENT CONTEXT.................................................................................. 6 2.1 OBJECTIVES ..................................................................................................... 6 2.2 REGULATORY REQUIREMENTS AND SAFETY CRITERIA............................. 7 2.3 IMPACTS OF CONCERN ................................................................................... 8 2.4 TIME SCALES OF CONCERN ........................................................................... 8 2.5 SPATIAL SCALES OF CONCERN..................................................................... 9 2.6 REPOSITORY ASSUMPTIONS........................................................................ 10 2.7 FUTURE HUMAN ACTIONS AND BEHAVIOUR ASSUMPTIONS ................... 10 2.8 DOSE RESPONSE ASSUMPTIONS ................................................................ 10 2.9 MODEL AND DATA ISSUES............................................................................ 11

3. SYSTEM DESCRIPTION .................................................................................. 13 3.1 OVERVIEW....................................................................................................... 13 3.2 USED FUEL...................................................................................................... 15 3.3 OTHER WASTES ............................................................................................. 16 3.4 CONTAINER..................................................................................................... 16 3.5 EMPLACEMENT ROOMS AND ENGINEERED SEALS................................... 17 3.6 REPOSITORY ARRANGEMENT...................................................................... 19 3.7 GEOSPHERE ................................................................................................... 20 3.8 BIOSPHERE..................................................................................................... 23 3.9 LAND USE........................................................................................................ 23

4. SCENARIO DEFINITION .................................................................................. 24 4.1 INTRODUCTION............................................................................................... 24 4.2 FEPS REVIEW.................................................................................................. 24 4.3 SCENARIOS IDENTIFIED IN RECENT STUDIES ............................................ 25 4.4 CHOICE OF SCENARIOS ................................................................................ 25

5. MODELS, CODES AND DATA......................................................................... 28 5.1 OVERVIEW....................................................................................................... 28 5.2 SOFTWARE QUALITY ASSURANCE .............................................................. 28

6. BASE SCENARIO............................................................................................. 30 6.1 SCENARIO DESCRIPTION .............................................................................. 30 6.2 CONCEPTUAL MODEL.................................................................................... 30 6.3 ANALYSIS ........................................................................................................ 32 6.4 DISCUSSION.................................................................................................... 37

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7. DEFECTIVE CONTAINER SCENARIO............................................................. 39 7.1 SCENARIO DESCRIPTION .............................................................................. 39 7.2 CONCEPTUAL MODEL.................................................................................... 40 7.3 ANALYSIS ........................................................................................................ 42 7.4 DISCUSSION.................................................................................................... 60

8. HUMAN INTRUSION SCENARIO..................................................................... 64 8.1 BACKGROUND................................................................................................ 64 8.2 SCENARIO DESCRIPTION .............................................................................. 66 8.3 CONCEPTUAL MODEL.................................................................................... 67 8.4 ANALYSIS ........................................................................................................ 69 8.5 DISCUSSION.................................................................................................... 74

9. SUMMARY AND CONCLUSIONS.................................................................... 76 9.1 SUMMARY OF APPROACH............................................................................. 76 9.2 SUMMARY OF RESULTS ................................................................................ 77 9.3 CONCLUSIONS ON THE THIRD CASE STUDY REPOSITORY SAFETY ....... 80 9.4 CONCLUSIONS ON METHODOLOGY ............................................................ 80 9.5 THE SAFETY CASE FOR THE DEEP GEOLOGIC REPOSITORY CONCEPT 81

REFERENCES......................................................................................................................... 83

APPENDIX A: LIST OF FEATURES, EVENTS AND PROCESSES (FEPs) ........................... 89

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LIST OF TABLES

Page Table 3.1: Main features of the Third Case Study repository and site and comparison with other

recent safety assessments. .................................................................................... 14 Table 3.2: Initial inventories of several radionuclides of interest for safety assessment .......... 16 Table 4.1: Scenarios explicitly analyzed in recent safety assessments ................................... 26 Table 4.2: Some potential scenarios or key factors that could define scenarios...................... 27 Table 7.1: Radionuclides identified as of potential concern in the screening analysis ............. 47 Table 8.1: Human actions that can affect a deep repository (from SKB 1999) ........................ 65 Table 8.2: Human intrusion pathways considered in recent safety assessments of used fuel

repositories. ............................................................................................................ 65 Table 8.3: Exposure specific parameters for Third Case Study human intrusion model .......... 68 Table 8.4: Dose estimates for core technician and drill crew .................................................. 70 Table 8.5: Dose estimates for resident and construction worker ............................................. 70

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LIST OF FIGURES

Page Figure 1.1: Illustration of the multi-barrier deep geologic repository concept considered in

the Third Case Study. A human figure is shown for scale....................................... 2 Figure 1.2: Report structure for the Third Case Study. .............................................................. 4 Figure 1.3: The IAEA Integrated Safety Assessment Methodology (IAEA 2004). .................... 5 Figure 2.1: Time scale for radioactivity decay in repository. ...................................................... 9 Figure 3.1: Illustration of the deep geologic repository concept, showing the container,

emplacement room, and the room and tunnel layout. A human figure is shown for scale................................................................................................................. 13

Figure 3.2: Standard 37-element CANDU fuel bundle. The bundle is 102-mm in diameter and 495 mm long................................................................................................... 15

Figure 3.3: Cutaway illustration of container showing copper outer shell with welded copper lid with handling lugs (right end), steel inner vessel, and steel inner baskets holding used fuel. ................................................................................................. 17

Figure 3.4: Cross-section of an emplacement room indicating the general geometry. The container is nested inside layers of buffer and backfill. .......................................... 18

Figure 3.5: Perspective cutaway view of emplacement room showing general arrangement of materials. ......................................................................................................... 18

Figure 3.6: Plan view of the vault. The main access shaft and the underground support facilities are at the top of the figure. Dimensions are in metres............................. 19

Figure 3.7: Surface topography and hydrology of the 84 km2 watershed containing the repository (not shown, but located near the southeastern corner). ...................... 20

Figure 3.8: Perspective view of the 84 km2 watershed area to 1.5 km depth showing: (a) intersection of major fractures with the surface; and (b) all discrete fractures and surface water features. .......................................................................................... 21

Figure 3.9: Perspective view of the area around the repository showing: (a) the entire 84 km2 watershed area at the repository horizon (670 m depth), with the major fractures intersecting this plane; and (b) more detail of the 3-D discrete fracture network in the vicinity of the repository (repository is not shown, but would be in the fracture-free volume near the intersection of the two cut-away planes). .......... 22

Figure 3.10: Illustration of a Canadian Shield surface biosphere similar to that assumed at the Third Case Study site. ..................................................................................... 23

Figure 5.1: General relationship between data and models in the Third Case Study. ............ 29 Figure 6.1: General conceptual model showing the main processes considered in the Base

Scenario. ............................................................................................................... 31 Figure 6.2: Radioactivity of key nuclides in used CANDU fuel after discharge from reactor

(220 MWh/kgU burnup). . ..................................................................................... 33 Figure 6.3: Total thermal power of used fuel after discharge from reactor (220 MWh/kgU

burnup). ............................................................................................................... 33 Figure 6.4: Some potential effects of glaciation on groundwater movement. ......................... 35 Figure 6.5: Cross-section of the Cigar Lake uranium ore body in Saskatchewan (from

Cramer and Smellie 1994). .................................................................................. 38 Figure 7.1: General conceptual model for the Defective Container Scenario. ......................... 41 Figure 7.2: Plan view of area around repository, showing the subregional watershed

boundary (thick black line), surface elevation (in metres Above Sea Level) with 5-m contour lines, the main surface water bodies (blue), and the location of the hypothetical repository (square outline). ................................................................ 43

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Figure 7.3: Groundwater flow vectors (arrows) and head contours (red lines) in a vertical cross-section through the repository (black horizontal line near 7000 to 8000 Easting), for three different permeability profiles. ................................................ 44

Figure 7.4: Plan view of particle tracks for the Case 2 geosphere (intermediate permeability) and a well water flow rate of 738 m3/a. ................................................................ 45

Figure 7.5: Vault-scale model showing: (a) vertical cross-section of room with buffer, backfill and EDZ layers, and (b) plan view of repository showing emplacement rooms, access tunnels and seals. ................................................................................... 48

Figure 7.6: Reference Case I-129 results showing: (a) concentration contours near the surface (the location of repository, near-surface fractures and surface waters are also shown), and (b) the time-varying I-129 mass flow outputs to the surface discharge areas..................................................................................................... 50

Figure 7.7: Dose rate impact for the Reference Case (Case 2 geosphere, self-sufficient farmer critical group, 738 m3/a well). ................................................................... 51

Figure 7.8: Effect of geosphere permeability on I-129 dose rates to a self-sufficient farmer critical group. ......................................................................................................... 51

Figure 7.9: Comparison of SYVAC3-CC4 and FRAC3DVS mass flow results from vault into geosphere, and from geosphere into surface biosphere, for (a) I-129 and (b) Cl-36. ......................................................................................................................... 53

Figure 7.10: Dose rate results from selected sensitivity study cases. .................................... 54 Figure 7.11: Distribution of the number of container failures from 45,000 randomly sampled

simulations. The most likely number is two failed containers. ................................ 56 Figure 7.12: Distribution of peak total dose rates for simulations with container failures. The

vertical blue line is the average peak dose rate of 3.6x10-7 Sv/a. .......................... 57 Figure 7.13: Downward cumulative distribution functions for the peak total dose rate and the

I-129 and Rn-222 peak dose rates for the Defective Container Scenario. ............. 57 Figure 7.14: Average and 95th percentile total dose rates from the probabilistic simulations. . 58 Figure 7.15: Calculated average dose rates to the critical group from fission product

radionuclides. ........................................................................................................ 59 Figure 7.16: Calculated average dose rates to the critical group from the natural U-238

decay chain nuclides present in the used fuel. ...................................................... 59 Figure 7.17: Mass distribution of long-lived radionuclides at one million years, for the

Reference Case. ................................................................................................. 61 Figure 7.18: Comparison of three safety indicators. .............................................................. 63 Figure 8.1: General sequence of events for inadvertent human intrusion................................ 66 Figure 8.2: General conceptual model for the Human Intrusion Scenario................................ 68 Figure 8.3: Calculated inadvertent exposures as a result of a borehole drilled into a

container. ............................................................................................................ 73 Figure 8.4: Dose-probability graph for inadvertent human intrusion exposures. .................... 75 Figure 9.1: Best-estimate (BE) Reference Case dose rates to a self-sufficient farmer in the

Defective Container Scenario. ............................................................................ 78 Figure 9.2: Effect of model input uncertainties on the peak total dose rate to a self-sufficient

farmer in the Defective Container Scenario. ........................................................ 78 Figure 9.3: Impact of various "what if" assumptions on calculated dose rates compared to

the Reference Case (BE-RefWell)......................................................................... 79 Figure 9.4: Impact of the repository over a one-million year time-frame evaluated using

different indicators for the Defective Container Scenario. .................................... 79 Figure 9.5: Comparison of the calculated average dose rates to a self-sufficient farmer

critical group living at the repository site from the EIS, the Second Case Study (SCS) and the Third Case Study (TCS)................................................................. 81

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1. INTRODUCTION 1.1 BACKGROUND Nuclear power stations supply about 35 to 40% of Ontario's electricity. The operation of these stations generates radioactive waste, including used CANDU fuel. Presently, this used fuel is stored in wet storage bays or in dry concrete containers at the power stations. Studies on options for storage and disposal of used fuel have been underway for approximately 30 years. As specified in the Canadian Nuclear Fuel Waste Act (2002), the Nuclear Waste Management Organization was created to compare the approaches, to consult with the public, and to recommend a long-term management approach (www.nwmo.ca). One approach is disposal in a repository located deep underground in stable rock. Deep geologic disposal has been selected as the preferred approach by several countries, and candidate sites are under evaluation in the US, Finland and Sweden. As part of its research into the geologic repository concept, Canada has developed the capability to prepare a postclosure safety assessment for such a repository. This capability is presently used to guide the development of a repository design, to identify research activities that would improve confidence in the performance of a repository, and to support technical program decisions. Two Canadian postclosure safety assessments have been completed for hypothetical repositories located in Canadian Shield rock - the Environmental Impact Statement (EIS) study (AECL 1994, Goodwin et al. 1994a) and the Second Case Study (Goodwin et al. 1996). Also, studies have been published for similar repository concepts in other countries, notably Sweden (SKB 1999), Finland (Vieno and Nordman 1999), Japan (JNC 2000), USA (BSC 2001) and Switzerland (NAGRA 2002). The previous EIS study considered titanium alloy containers with 72-fuel-bundle capacity placed vertically into boreholes along the vault rooms, and assumed the repository was located in sparsely-fractured granitic rock with very low permeability. The Second Case Study considered 72-bundle copper containers placed horizontally within the vault rooms, and assumed the repository was located in granitic rock with substantially higher permeability. The Third Case Study (TCS) is a new postclosure safety assessment of the deep geologic repository concept for used CANDU fuel in the Canadian Shield (Figure 1.1). It evaluates an updated repository design and site assumptions. The repository has sufficient capacity to hold all the used fuel from present Canadian nuclear power stations to the end of their planned life. The TCS uses a copper and steel container with 324-bundle capacity, placed horizontally within the vault rooms. The site is also different from that assumed in the previous Canadian studies. The geologic setting is hypothetical, but believed to be representative of potential sites that could exist within the Canadian Shield. The TCS assumes the repository is located in granitic rock that is characterized by an intermediate permeability and a geostatistically-generated discrete fracture network. The two previous Canadian studies have indicated repository design and site characteristics under which a deep geologic repository could be safely sited in the Canadian Shield. The Third Case Study provides a third example of the characteristics that could lead to a safe repository. In this way, the Third Case Study also contributes to the safety case for the deep geologic repository concept.

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Figure 1.1: Illustration of the multi-barrier deep geologic repository concept considered in the Third Case Study. A human figure is shown for scale.

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1.2 PURPOSE The main objectives of the Third Case Study are: • to assess key aspects of the postclosure safety of a deep geologic repository for used fuel

based on a current Canadian design concept, and • to test and evaluate current safety assessment methods and tools. The study results should be considered within the following context: • The Third Case Study focuses on three key scenarios - design base, defective container

(groundwater transport) and inadvertent human intrusion - and it evaluates radiological doses to humans. Prior studies have identified these as important aspects of postclosure safety. An assessment of a real candidate site would consider other scenarios and impacts.

• The Third Case Study is based on a specific repository design and hypothetical site

concept. The geosphere is considered representative of a potential Canadian Shield setting, and not necessarily representative of the best (nor worst) features of a potential site.

• The analyses are appropriate for the level of information that would typically be available for

a candidate site during the site evaluation stage but prior to exploratory drilling.

• The Third Case Study incorporates a geostatistical approach for the creation of a geosphere model. The safety assessment uses the same hydrogeological numerical code as was used in the development of the site model by the geoscience group. This approach represents a significant improvement in how the geosphere model and its uncertainty could be incorporated into the assessment.

The results of this study provide a basic test of postclosure safety for this repository concept and hypothetical site, and a basis for future iterations in which progressively more topics can be addressed. It is not a full safety case. The results of a similar study for a real site would provide information on whether the expected safety margins are sufficient to proceed with exploratory drilling as part of a staged site evaluation program. 1.3 REPORT OUTLINE This report is an overview of the Third Case Study postclosure safety assessment. Additional details are found in the supporting documents and references therein, as illustrated in Figure 1.2. These reports are intended for a technical audience familiar with nuclear waste disposal topics. The content of these supporting reports are: • Site and Design Description - Summarizes the hypothetical site characteristics and the

reference used fuel, container and repository concept design features.

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Figure 1.2: Report structure for the Third Case Study.

• Features, Events and Processes - Documents the various factors that might be relevant to

the assessment, and indicates which are included and in which scenario. • Reference Data and Codes - A description of the main computer models used and their

documentation, and a listing of the model input data. • Defective Container Scenario - Detailed description of the analysis for this scenario. The safety assessment methodology followed for the Third Case Study is consistent with that used in other recent international assessments, and in particular it follows the IAEA Integrated Safety Assessment Methodology (IAEA 2004), with the steps as illustrated in Figure 1.3. The structure of this report reflects these steps as follows: • Section 2 summarizes the context of the safety assessment. That is, it describes the

general principles considered in carrying out the study, including the safety criteria. • Section 3 describes the main features of the repository and site relevant to safety

assessment. • Section 4 defines the scenarios identified for analysis in this assessment. • Section 5 provides a brief description of the models and datasets used for quantitative

analyses. • Sections 6, 7 and 8 each describe one of the scenarios analysed and the main results,

including comparison with safety criteria. • Section 9 summarizes the study approach and its general conclusions.

Third Case StudyPostclosure Safety Assessment Report

OPG 06819-REP-01200-10109

TCS Site & Design DescriptionOPG 06819-REP-01200-10124

TCS Features, Events and ProcessesOPG 06819-REP-01200-10125

TCS Reference Data and CodesOPG 06819-REP-01200-10107

TCS Defective Container ScenarioOPG 06819-REP-01200-10126

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Figure 1.3: The IAEA Integrated Safety Assessment Methodology (IAEA 2004). The Third Case Study followed the steps outlined here, but without the full iteration that would occur in a real siting process.

Acceptance

Assessment Context

Describe System

Develop andJustify Scenarios

Formulate andImplement Models

Run Analyses

Interpret Results Review and Modification

Effective to ModifyAssessment?

AdequateSafety Assessment?

Compare withSafety Criteria

Rejection

Yes

Yes

No

No

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2. ASSESSMENT CONTEXT This section describes the factors that define the scope of the safety assessment, including the objectives, safety criteria, time and space scales, and assessment context assumptions. 2.1 OBJECTIVES (1) Assess key aspects of the postclosure safety of a deep geologic repository for used fuel based on a current Canadian design concept. The postclosure safety of the design concept is assessed based on specific scenarios and impacts. In particular, the Third Case Study considers three scenarios - design base, defective container (i.e., groundwater transport) and inadvertent human intrusion. The impact considered is the radiological dose rate to humans. Prior Canadian studies and other international studies (e.g., Goodwin et al. 1994a, Goodwin et al. 1996, SKB 1999) have identified these as important aspects of postclosure safety. The study will not include other scenarios such as earthquakes or glaciation, nor other impacts such as preclosure operational safety, chemical contaminants, or impacts on biota. We recognize that a complete safety assessment for any specific site would require treatment of these latter topics (and others). The Third Case Study is based on a specific repository design and site concept. Since the site is hypothetical, the philosophy guiding the definition of its parameters is that the site should be feasible and conditionally acceptable. That is, it would meet reasonable technical siting criteria, such as those outlined in Davison et al. (1994a). At the same time, the site is not based on the most favourable properties that might be found only in a few settings. The analyses are based on information that would typically be available for a candidate site during the site evaluation stage but prior to significant exploratory drilling. This affects the degree of detail in the geosphere conceptual model and in the repository engineering analyses. The analyses presented here provide a basic test of postclosure safety for this repository concept and assumed site, and a basis for future iterations in which progressively more topics can be addressed. The results of a similar study for a real site would provide information on whether the expected safety margins were sufficient to proceed with exploratory drilling as part of a staged site evaluation program. The results of this study can also contribute to a larger "safety case" for the deep geologic repository concept that would draw on results from other case studies (Canadian and international), as well as other arguments regarding safety, such as the age of deep groundwaters and natural analogs. (2) Test and evaluate current safety assessment methods and tools. Key features of the assessment approach used in the Third Case Study are listed below: • The Third Case Study incorporates a geostatistical approach for the creation of a

geosphere model. The approach represents a significant improvement in how geosphere uncertainty could be incorporated into the assessment.

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• The safety assessment uses the same hydrogeological numerical code as could be used by a site characterization geoscience group, in order to ensure accurate transfer of the geosphere model from site characterization to safety assessment.

• There is a more comprehensive documentation of relevant features, events and processes included in the assessment, based on an internationally recommended structure.

• A screening assessment is conducted to identify those radionuclides that merit more detailed analysis, starting from an exhaustive list of possible radionuclides.

• The analysis uses several safety assessment tools, ranging from simple models, to probabilistic system models, to detailed transport models.

• Cross-checking of results from the different models is carried out where possible. This helps to establish that the models are numerically accurate, and that the simpler models are acceptable (or conservative) representations of the system.

• Results are presented for various sub-scenarios of interest, including several "what if" sensitivity cases, and are explicitly stated in terms of calculated dose rate consequences for these sub-scenarios, rather than as a single risk-based value.

• The main quantitative analyses use both best-estimate and probabilistic approaches. The effect of uncertainties are addressed through both probabilistic and sensitivity analyses.

The Third Case Study is an opportunity to test this assessment approach and the corresponding analysis tools, and as a result to identify directions for improvement. 2.2 REGULATORY REQUIREMENTS AND SAFETY CRITERIA Historically, the Canadian regulatory document R-104 (AECB 1987) provided quantitative criteria for safety assessment of a deep geologic repository. However, Canadian regulations related to nuclear waste disposal are currently being revised, and R-104 has been withdrawn. A new draft policy document P-290 (CNSC 2003) has been issued by the Canadian Nuclear Safety Commission (CNSC). The Third Case Study is consistent with this draft CNSC policy statement. However, this policy does not provide quantitative safety criteria. For the Third Case Study, we compare calculated dose rates to the recommendations of the International Committee on Radiation Protection, ICRP 81 (ICRP 2000), as well as to the average Canadian natural background dose rate. Complementary indicators of safety are also derived from natural background data. The key quantitative recommendations made in ICRP 81 are as follows: • Exposure from natural processes and inadvertent human intrusion should be considered. • For natural processes, assessed doses should be compared with a dose constraint to a

member of the public of 3x10-4 Sv/a. • For human intrusion, intervention (e.g., modification of repository design) should be made to

reduce the probability of human intrusion or to limits its consequences where intrusion could lead to sufficiently high doses to those living around the site. This intervention is not likely to be justifiable for an annual dose of below 0.01 Sv/a, while it is almost always justifiable for an annual dose above 0.1 Sv/a.

With respect to natural background, we consider the following indicators: • the Canadian average background dose rate to humans from natural sources (0.0017 Sv/a,

Grasty and LaMarre 2004), • the natural background concentration of radionuclides in Canadian surface waters, and • the natural flux of radionuclides from the geosphere to the surface biosphere.

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2.3 IMPACTS OF CONCERN Previous studies have indicated that the most significant postclosure impacts are those associated with radiological doses to potential human groups living or working near the repository site (e.g., Goodwin et al. 1994a). In keeping with the concept of a "critical group", we conservatively assume that people live near the site in the future, and have lifestyles that maximize their potential exposure doses while behaving in an otherwise reasonable manner. For example, it is reasonable to anticipate that people living at the site in the future might obtain their food from a garden. We therefore assume that the critical group does use a garden, and we further maximize their exposure by locating the garden on the most contaminated soil. By evaluating the dose rate received by this hypothetical critical group, we provide a reasonable upper bound for the potential dose rate that could be received by a variety of possible future inhabitants living at or downstream from the site. With respect to chemical toxicity, the detailed EIS assessment showed that risks from releases of chemically toxic contaminants from the used fuel bundles were small compared to radiological risks (Goodwin et al. 1994a, p.260). In addition, most of the analyses needed to assess chemical toxicity risks (e.g., failure mechanisms, transport in the geosphere and biosphere) are substantively covered by the radionuclide contaminant models, so chemical toxicity calculations would add little to our appraisal of our safety assessment methods and tools, which is one objective of the Third Case Study. Consequently, the study does not consider chemical toxicity, although we recognize that this would be required in a siting-related safety assessment. The previous EIS and SCS assessments considered the long-term impact of the repository on non-human biota, and found no significant effects (Goodwin et al. 1994a, p.261; Goodwin et al. 1996, p.58). Furthermore, the whole approach to assessing ecological impact is presently being revised by the relevant international bodies and regulators, and the recommended methodology has not yet been established. Therefore, the Third Case Study will focus on human health as an indicator of the impact of the repository. Clearly, calculation of potential impacts to non-human biota would be required in a siting-related safety assessment. 2.4 TIME SCALES OF CONCERN The CNSC P-290 draft policy (CNSC 2003) indicates that the period over which the future impacts of radioactive waste are assessed should include the period over which the maximum impacts are expected. The deep geologic repository is expected to be capable of isolating the used fuel such that the maximum impact would likely occur well beyond 10,000 years. Since 98% of the used fuel is natural uranium, as radionuclides decay, the radioactivity in the repository will eventually become similar to that of uranium ore bodies found in other locations on the Canadian Shield. This occurs on times scales of about one million years (Figure 2.1). Therefore, for the Third Case Study, future impacts will be assessed to the time of the peak dose rate in general, with a one-million-year time base. It is recognized that estimating impacts becomes increasingly uncertain at long times, but nonetheless this study will consider this time scale in order to illustrate the potential impact.

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Figure 2.1: Time scale for radioactivity decay in repository. The gamma-emitting fission products decay within about 1000 years. The remaining fuel radioactivity becomes comparable to that of the granite in the surrounding watershed after about 10,000 years. On time scales of about 1 million years, the residual used fuel radioactivity is dominated by that of the uranium in the fuel (and its decay chain products), a level that is comparable to natural uranium ore bodies.

2.5 SPATIAL SCALES OF CONCERN The Third Case Study is a postclosure safety assessment of a deep geologic repository design at a hypothetical site on the Canadian Shield. The spatial domain that will be considered is the subregional watershed area (approximately 84 km2) in which the repository is sited (see Figure 3.7). The study will specifically focus on the local area that includes the groundwater discharge points from the repository, and the nearest major water body in which these discharges are collected. In this area, the concentrations of nuclides released from the repository, and therefore the calculated dose rates, would be the highest. Further downstream from this location, radionuclide concentrations are expected to be significantly diluted, resulting in substantially lower dose rates to people or non-human biota residing there.

101 102 103 104 105 106 107103

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109

Time [a]

To

tal

Ra

dio

act

ivity

[T

Bq

]

TimeScale.syView

Used Fuel - TotalUsed Fuel - Fission ProductsUsed Fuel - Uranium Chain10 km x 10 km x 1 km Shield graniteMcArthur Lake U ore body, CanadaCigar Lake U ore body, Canada

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2.6 REPOSITORY ASSUMPTIONS The repository, including the containers, is assumed to be designed, built and sealed as described in the TCS Site and Design Description report (2004). This includes a quality assurance program to ensure that the repository meets specifications. The scope of the Third Case Study is limited to used fuel bundles from present Canadian CANDU reactors. The repository will hold 3.6 million bundles, which is the present projected lifetime output from these reactors. The repository is constructed, operated and closed according to the present OPG reference schedule, as used for cost estimate purposes. Some key time lines are: • used fuel bundles are at least 30-years old at emplacement, • repository operation (i.e., filling of vault rooms) lasts 30 years, • the post-operation monitoring period with open access tunnels lasts 70 years, and • final decommissioning and closure takes up to 25 years. Other repository assumptions are described in the specific relevant sections. 2.7 FUTURE HUMAN ACTIONS AND BEHAVIOUR ASSUMPTIONS We assume that future humans will largely resemble present day humans in terms of societal behaviour, capabilities and actions. This means that: • There is no credit taken for advances in science and technology that might reduce the risk

from the repository - e.g., no "cure for cancer" and no simple waste transmutation process. • People live in circumstances consistent with current North American practice. Specifically,

people live in a variety of environments such as urban settings or individual farms, and tend to stay in one location for periods of years.

• There is no man-made catastrophe that has significantly contaminated the biosphere

around the repository site and made it unsuitable for human habitation. • Human activities that could affect the local environment (e.g., construction, water diversion,

climate change) are consistent with present capabilities and occur for reasons that would make current sense.

• Societal knowledge of the repository will provide control for some initial period, but cannot

be relied on indefinitely to prevent inadvertent human intrusion into the site. 2.8 DOSE RESPONSE ASSUMPTIONS To convert received dose to a measure of risk to an individual or population, the International Commission on Radiological Protection (ICRP) 1990 recommendations are used (ICRP 1991). These are based on the Linear No-Threshold model, and have three steps.

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1. The concentration of a radionuclide in some environmental medium is multiplied by a dose coefficient (previously called a dose conversion factor) and other factors such as duration of exposure or amount of material ingested to obtain a dose equivalent (previously called an annual effective dose equivalent) in sieverts per year (Sv/a).

2. The total dose to an individual is the sum of the dose equivalents from all radionuclides and all exposure pathways for a given exposure scenario.

3. This total dose may be multiplied by a risk or dose-to-risk conversion factor, which converts

the total dose into a specified health effect. The factor 0.073/Sv may be used for converting a human dose rate to a risk of serious health effects, including fatal and non-fatal cancers and serious genetic effects (ICRP 1991).

Radiological exposures to humans are converted to dose rates using dose coefficients based on the 1990 ICRP recommendations. Human internal and external dose coefficients are from ICRP 72 (ICRP 1996) and Eckerman and Leggett (1996), respectively. The dose coefficients are selected assuming the chemical species that leads to the highest dose (Garisto 2002). 2.9 MODEL AND DATA ISSUES The Third Case Study is a postclosure safety assessment of a geologic repository for used fuel located at a hypothetical site. Various model and data issues will be handled as outlined below:

(a) Degree of conservatism. The intent is to use best-estimate (realistic) values in general for the main safety assessment calculations. For example, the initial radionuclide inventories will be based on an average fuel burnup of 220 MWh/kgU. The uncertainty in the value for a parameter can be expressed via a probability density or parameter distribution function, and included in the probabilistic analyses. Conservative values will only be used where a realistic value is unavailable, or as part of screening calculations, or as part of "what if" scenarios. Conditions corresponding to extreme values should be considered explicitly through "what if" or sensitivity studies. (b) Conceptual model uncertainty. The models used are intended to be either realistic or conservatively simple for the processes they represent. One important conceptual model uncertainty is related to the geosphere model. It is recognized that the information available at a real site will have uncertainties - for example, in the locations of fracture zones. For the Third Case Study, we will test an approach for directly linking the site geosphere model and its uncertainties into the safety assessment. This will be tested using one geostatistically-generated fracture network model, although the methodology can be extended to consider others. Within this one model, the impact of the following geosphere uncertainties will be addressed: bulk rock permeability, effective dispersivity, sorption coefficients, equivalent porosity versus dual-porosity model, and varying well water flow rate. The effects of salinity, climate change and glaciation will not be explored in this study. Another important conceptual model uncertainty is the treatment of time-varying conditions. The approach to handling this uncertainty has two parts.

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First, for any particular simulation, most models will use constant parameter values, but the probabilistic calculations will consider the range of possible values that could occur over time. This approach gives an estimate of the range of results for time-varying parameters for processes where the effects vary roughly linearly with the parameter. Subsequent sensitivity analyses will identify parameters for which the parameter value uncertainty is significant. Second, complementary safety indicators will be used in addition to the radiological dose rate. In particular, the radiotoxicity concentration in the nearby lake and the radiotoxicity flow from the geosphere will be calculated. These indicators should be less sensitive to the large scale changes that would take place on the surface over long time scales due to, for example, glaciation. (c) Parameter value uncertainty. The effects of parameter value uncertainty will be addressed by a combination of probabilistic analyses and deterministic sensitivity analyses. The parameter value uncertainty can be defined using parameter distribution functions to describe the range of values and their likelihood. The probabilistic analyses will include random sampling and defined sampling approaches to characterise the effects of the parameter value uncertainty. (d) Scenario uncertainty. We do not intend to provide a complete scenario analysis for the Third Case Study, but rather to focus on a base scenario (in which the repository is built as designed) and two safety-assessment scenarios that have consistently been identified as important in previous Canadian and other international studies, i.e., the defective container and human intrusion scenarios.

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3. SYSTEM DESCRIPTION 3.1 OVERVIEW The Canadian concept for deep geologic disposal of used nuclear fuel consists of used fuel bundles placed in durable containers, and the containers surrounded by self-sealing clay-based materials and emplaced within rooms excavated deep in a stable geological formation. This concept is similar to that considered by other countries for disposal of their high-level wastes, and provides multiple, independent, passive barriers for isolating and retaining the radionuclides. The range of engineered designs under consideration is described in Russell and Simmons (2003). For the Third Case Study, a specific disposal concept and site is assumed as shown in Figure 3.1. The specific concept is based on used fuel containers, with an external copper shell and an internal steel structural vessel, and emplaced horizontally in an in-room configuration. Table 3.1 provides a summary of the design features of this concept, along with those considered in other similar safety assessment studies. The remainder of this section describes the key characteristics of the design and site. More details are available in the TCS Site and Design Description report (2004).

Figure 3.1: Illustration of the deep geologic repository concept, showing the container, emplacement room, and the room and tunnel layout. A human figure is shown for scale.

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Table 3.1: Main features of the Third Case Study repository and site and comparison with other recent safety assessments.

Parameter Third Case Study

AECL Second Case Study

AECL EIS Case Study

SKB SR 97

Reference This report Goodwin et al. 1996

Goodwin et al. 1994

SKB 1999a

Total used fuel, MgU

69,100 (CANDU type)

82,000 (CANDU type)

162,000 (CANDU type)

8,000 (BWR type)

Fuel burnup, MWh/kgU

220 200 190 920

Fuel cooling time, a

30 10 10 40

Used fuel, MgU/container

6.2 1.4 1.4 2

Number containers

11,200 60,100 119,000 4,000

Container design

Copper shell, steel inner vessel

Copper shell, glass bead internals

Titanium shell, glass bead internals

Copper shell, cast steel insert

Emplacement Geometry

Two rows, in-room

Two rows, in-room

Three rows, in-floor

Single row, in-floor

Buffer thickness, m

0.5-1.1 radial 0.5-0.8 radial 0.25 radial 0.35 radial

Vault area, km2

2.0 3.4 3.2 1.7 to 2.6 (site-dependent)

Depth, m 670 500 500 Approx. 500 Geosphere Canadian Shield

granite, hypothetical site

Canadian Shield granite, Whiteshell site but permeable rock

Canadian Shield granite, Whiteshell site

Fennoscandian Shield granite, 3 sites

Bulk rock permeability at vault level, m2

7x10-19 1x10-19 1x10-17 1x10-16 to 3x10-15

Surface topography

Hypothetical site. Granite extends close to surface, shallow topography.

WRA1 site. Granite extends close to surface, shallow topography.

WRA1 site. Granite extends close to surface, shallow topography.

3 sites

Surface biosphere

Temperate inland boreal

Temperate inland boreal

Temperate inland boreal

Coastal or near-coastal boreal

1 Whiteshell Research Area, near Pinawa, Manitoba.

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3.2 USED FUEL The TCS repository is designed to hold about 3.6 million CANDU used fuel bundles, which is the amount of used fuel expected to be discharged from CANDU reactors in Canada, given the planned lifetime of the various existing and closed OPG, Bruce Power, AECL, New Brunswick Power and Hydro-Québec nuclear power stations. The characteristics of natural-uranium-based CANDU used fuel bundles are described in detail in Tait et al. (2000a). A representative fuel bundle is shown in Figure 3.2. Although the physical characteristics of the used fuel have varied slightly over time and from station-to-station, they are all similar. The standard-length 37-element fuel bundle containing 19.2 kgU (initial) is adopted here as the reference used fuel bundle, as recommended in Tait et al. (2000a). The fuel element consists of pellets of sintered UO2 enclosed in a 0.2-mm-thick Zircaloy-4 sheath with a thin layer of graphite-based CANLUB on the inside surface. The most important used fuel parameter, in terms of the used fuel behaviour in the repository, is the burnup. Tait et al. (2000a) defined 455 kW/bundle average power and a median burnup of 220 MWh/kgU as reference values, with 290 MWh/kgU as the 95th percentile burnup and 320 MWh/kgU burnup as the 99th percentile burnup. The corresponding radionuclide inventories, thermal power, and radiation characteristics were calculated using the ORIGIN-S code (Hermann and Westfall 1995) and tabulated in Tait et al. (2000b). The inventories for some radionuclides of interest are listed in Table 3.2. At the time of emplacement, the used fuel will have been in storage for a minimum of 30-years after discharge (up to 60 years in some cases). This time period will have included at least 10 years storage in water, with the balance either in the water pools or in dry storage. The fuel elements will be mostly intact, with less than 1% having sheath (cladding) defects.

Figure 3.2: Standard 37-element CANDU fuel bundle. The bundle is 102-mm in diameter and 495 mm long.

Zirconium AlloyFuel Sheath

Ceramic Pellets ofUranium Dioxide

AA_003, v.01AJCP, 96 MAR 06

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Table 3.2: Initial inventories of several radionuclides of interest for safety assessment

Radionuclide Half-life [years] Inventory [mol/kgU] 1 C-14 5,730 6.1x10-6 C-14 (in Zircaloy) 5,730 2.2x10-6 Ca-41 103,000 2.3x10-6 Cl-36 301,000 9.8x10-6 Cl-36 (in Zircaloy) 301,000 1.3x10-6 I-129 15,700,000 4.2x10-4 Np-237 2,140,000 1.7x10-4 Pu-239 24,100 1.1x10-2 Se-79 1,100,000 1.8x10-5 Sr-90 29 7.6x10-4 Tc-99 213,000 2.4x10-3 U-238 4,470,000,000 4.1 Zr-93 1,530,000 1.4x10-3 Zr-93 (in Zircaloy) 1,530,000 1.6x10-4

1 220 MWh/kgU burnup, 30 years cooling 3.3 OTHER WASTES If the federal government selects geologic disposal as the approach for long-term management of nuclear fuel waste, a Canadian deep geologic repository may also be considered for disposal of other Canadian high-level wastes, such as specialty used fuels from research reactors. However, the amount of such additional wastes would be small compared to the amount of CANDU used fuel. If these wastes were to be included, it is likely that they would be placed in containers similar to those for used fuel and would not materially change the repository design or performance. These additional wastes are not considered further within the scope of this assessment. This study also assumes that there are no low or intermediate level wastes in the used fuel repository. 3.4 CONTAINER The container design is described in the TCS Site and Design Description report (2004), and illustrated in Figure 3.3. This container is similar in size and concept to that being considered as part of the Swedish and Finnish repository programs. The outer corrosion-resistant barrier is a 25-mm thick shell of oxygen-free phosphorous doped copper. This material has excellent corrosion resistance properties under the low-oxygen conditions that naturally occurs deep underground. The inner vessel is a 96-mm thick shell of carbon steel that provides mechanical support during both handling and emplacement. It is also designed to handle the additional loads imposed if a 3-km thick glacier covered the repository site. After the fuel is loaded and the air inside the container is replaced with helium, the steel vessel is bolted shut and the copper shell is welded closed. At least two independent methods would be used to inspect the weld to ensure that there were no defects. The container would hold 324 fuel bundles, and would weigh about 24 Mg loaded.

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Figure 3.3: Cutaway illustration of container showing copper outer shell with welded copper lid with handling lugs (right end), steel inner vessel, and steel inner baskets holding used fuel. The container is about 1.2 m in diameter and 3.9 m long.

3.5 EMPLACEMENT ROOMS AND ENGINEERED SEALS The containers are emplaced into rooms excavated from the host rock at the repository depth. The reference layout is shown in Figures 3.4 and 3.5. The containers are placed horizontally in two rows ("in-room" configuration), and surrounded by layers of clay-based sealing materials. These sealing materials provide a stable chemical and mechanical environment for the containers. The innermost clay or "buffer" layer is composed of dense bentonite, a clay that swells as it saturates with water to fill any spaces or gaps, and so provides a tight, low-permeability layer around the container. This dense layer inhibits chemical reactions and microbial activity that could cause corrosion of the container. It is surrounded by a 50:50 bentonite-sand buffer that provides further self-sealing capability, as well as better mechanical strength and thermal conduction. The remaining space between the buffer and the rock walls is filled with a mixture of dense backfill (placed as blocks) and light backfill (blown in pneumatically). A typical emplacement room would be about 300 m long, and contain 108 containers. The elliptical cross-section is stable under anticipated Canadian Shield conditions, and allows the rock walls to be self-supporting. As emplacement rooms are filled with containers, they would be sealed from the access tunnels by mechanical plugs using clay- and/or concrete-based materials.

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Figure 3.4: Cross-section of an emplacement room indicating the general geometry. The container is nested inside layers of buffer and backfill.

Figure 3.5: Perspective cutaway view of emplacement room showing general arrangement of materials. Containers are inserted with a dense bentonite jacket pre-installed to ensure good contact and to provide shielding during handling.

Light Backfill

Dense Backfill

7.14 m

4.20 m

Bentonite/ Sand Buffer

Dense Bentonite

Buffer

ContainerContainer

High PerformanceConcrete

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3.6 REPOSITORY ARRANGEMENT The emplacement rooms would be laid out in the repository as shown in Figure 3.6. A square layout has been adopted since this is convenient for engineering purposes, and since space was available (see Section 3.7) to locate this square geometry between the major fracture zones of the reference geosphere (i.e., there are no major fracture zones intersecting the repository). There are 104 emplacement rooms in the repository, each about 315 m long. For structural and thermal reasons, they are placed 45 m apart. The repository is located at a depth of about 670 m below ground surface. It would be accessed by a vertical shaft. In addition to the emplacement rooms and access tunnels, there would also be additional excavations for maintenance and testing rooms, and ventilation shafts.

Figure 3.6: Plan view of the vault. The main access shaft and the underground support facilities are at the top of the figure. Dimensions are in metres.

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3.7 GEOSPHERE The Third Case Study assumes that the repository is constructed in the crystalline rock of the Canadian Shield. The previous two Canadian studies used the geology of a Canadian Shield site near Lac du Bonnet, Manitoba as a basis for characterizing the site. The Third Case Study, however, considers a new but hypothetical site, as described in the TCS Site and Design Description report (2004). The site is assumed to be composed of granitic rock, intersected by a complex set of interconnecting major fractures and discontinuities. Figure 3.7 illustrates the surface topography and hydrology for the 84 km2 watershed area in which the repository is located. The watershed is drained by several streams flowing south or southwest, into the river along the south border. Figure 3.8 shows the discrete fracture model for this watershed area, while Figure 3.9 shows a close-up of the area around the repository location. The derivation of the fracture network is based on Canadian Shield fracture statistics, and was generated as described in Srivastava (2002). For a real site, the fracture locations would not be known with certainty, especially at the beginning of the site evaluation process. The Third Case Study uses just one conceptual fracture model for the site, recognizing that the effect of the fracture network uncertainty could be considered during a site evaluation by repeating the present analyses for a range of fracture networks that represent the model uncertainty.

Figure 3.7: Surface topography and hydrology of the 84 km2 watershed containing the repository (not shown, but located near the southeastern corner). Elevation is in units of mASL, metres Above Sea Level. Surface waters and wetlands are marked in blue.

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(a)

(b)

Figure 3.8: Perspective view of the 84 km2 watershed area to 1.5 km depth showing: (a) intersection of major fractures with the surface; and (b) all discrete fractures and surface water features. The model is consistent with a Canadian Shield setting, and shows an interconnected near-surface fracture network, with relatively fewer deep fractures.

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(a) (b)

Figure 3.9: Perspective view of the area around the repository showing: (a) the entire 84 km2 watershed area at the repository horizon (670 m depth), with the major fractures intersecting this plane; and (b) more detail of the 3-D discrete fracture network in the vicinity of the repository (repository is not shown, but would be in the fracture-free volume near the intersection of the two cut-away planes).

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3.8 BIOSPHERE The Third Case Study biosphere is based on central Canadian Shield conditions. The topography is relatively flat, with a general slope from north to south. The granitic rock of the Canadian Shield extends close to surface throughout most of the site (Figure 3.10), with some sedimentary layers in the low-lying areas such as near lakes and rivers. The local watershed is bounded by streams on the east and west, and a larger river on the south (Figure 3.7). There are small lakes and wetlands on the surface above the repository, notably a lake near the eastern corner. The water table is fairly close to the surface, and there are no aquifers. The site is not directly adjacent to either Hudson Bay or the Great Lakes. The present climate is typical of a cool continental location. The annual average surface temperature is +5oC. Winds and precipitation are typical for the Canadian Shield, with 2.36 m/s annual average wind speed and 0.8 m/a annual average precipitation. The surface ecosystem is temperate boreal. This is a relatively stable ecosystem for present climatic conditions. It is possible that global warming could cause changes to the climate over the next 1000 years. On a longer time scale (after 10,000 years), another glaciation cycle could begin with a period of around 120,000 years, but returning during interglacial periods to climate conditions similar to present-day (Peltier 2002). In the Third Case Study, only present day climate conditions are considered. 3.9 LAND USE The Third Case Study site is hypothetical. For the present study, we assume that there are no land uses, endangered ecosystems or mineral resources that would make the site inappropriate for a repository.

Figure 3.10: Illustration of a Canadian Shield surface biosphere similar to that assumed at the Third Case Study site.

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4. SCENARIO DEFINITION 4.1 INTRODUCTION From the reference system description (Section 3), the next step in a safety assessment is to define the scenarios to be considered. Each scenario is a hypothetical sequence of processes and events; a set of such scenarios are defined for the purpose of illustrating the range of future behaviours and states of a repository system (IAEA 2003, NEA 2001). For the Third Case Study, a limited set of important scenarios and their characteristics was selected based on: 1. the features, events and processes (FEPs), applicable to the selected design and site; and 2. scenarios identified as important in previous Canadian and international studies. 4.2 FEPS REVIEW Previous repository studies have identified many factors that may need to be considered as part of an assessment scenario. For example, the Canadian EIS study identified a set of about 300 factors (Goodwin et al. 1994b). Subsequently, an international "FEP" database was prepared that provides a comprehensive set of these factors based on studies in several countries (NEA 2000). For the OPG Third Case Study, a specific database has been developed that follows the international FEP database. It is structured as follows: 1. External factors - factors outside the repository system. 2. Repository factors - factors associated with the repository and its surroundings, including

the local geosphere and biosphere. 3. Radionuclide and contaminant factors - factors associated with the release, migration and

impact of radionuclides and other contaminants. The analysis of this set of potential factors provides an organized method for thinking about scenarios. First, the factors are reviewed against the information available for the Third Case Study repository and site, and a judgement made as to which should be included as part of the safety assessment, taking into account their likely impact on safety. At this point, many factors could be excluded from further consideration because they do not apply to this site, or are unlikely to occur over relevant timeframes, or would have little consequence. Also, some factors are excluded from consideration by the scope of work for the Third Case Study assessment. Those factors that have been identified for inclusion in the safety assessment must then be allocated to a specific scenario. In some cases, the factors are common among all scenarios (e.g., radionuclide inventories), while in other cases they are specific to one scenario and, in some cases, actually define a scenario. (For example, the presence of manufacturing defects in the container defines the Defective Container Scenario.) The full set of factors in the database is listed in Appendix A, which also indicates which were included explicitly in the Third Case Study assessment models. The documentation of these factors is in the TCS Features, Events and Processes report (2004).

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4.3 SCENARIOS IDENTIFIED IN RECENT STUDIES Table 4.1 lists the scenarios explicitly analyzed in the previous Canadian EIS and Second Case Studies, and in the recent Swedish SR 97 study. All these postclosure safety assessments were for deep geologic repositories in granitic rock. In these studies, the most important consequences were those from the groundwater transport scenario with early container failures and from human intrusion. 4.4 CHOICE OF SCENARIOS Table 4.2 lists a number of potential scenarios, including several that were identified by reviewers of the EIS safety assessment (SRG 1995). The reasons for considering these (or not) within the Third Case Study are briefly stated in Table 4.2 (see also the TCS Features, Events and Processes report, 2004). The main scenarios identified for analysis in the Third Case Study are:

1. Base Scenario, in which the repository is built according to design, and the overall system behaves as expected.

2. Defective Container Scenario, in which some containers are assumed to be emplaced

with small undetected defects, resulting in release of radionuclides into the groundwater. Several "what if" sensitivity cases will be considered within this scenario, including defective barriers, all containers failing, and zero solubility and sorption.

3. Human Intrusion Scenario, in which the engineered and natural repository barriers are

bypassed by a borehole that is inadvertently drilled through a container, bringing used fuel debris directly to the surface.

These scenarios are consistent with those identified as important in previous assessments. They are expected to provide a good test of the safety of the hypothetical Third Case Study. For a safety assessment of a real site, more scenarios would be analyzed.

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Table 4.1: Scenarios explicitly analyzed in recent safety assessments1

EIS, Canada (Goodwin et al. 1994a): - Groundwater transport scenario - Open-borehole scenario - Human intrusion scenario Second Case Study, Canada (Goodwin et al. 1996): - Groundwater transport scenario (container defect) SR 97, Sweden (SKB 1999): - Base scenario - Canister defect scenario - Climate scenario - Earthquake scenario - Intrusion scenario TILA 99, Finland (Vieno and Nordman 1999) - Base scenario - Canister defect (small hole) scenario - Disappearing container scenario H12, Japan (JNC 2000) - Base scenario - Uplift and erosion scenario - Climate and sea-level change scenario - Construction defect scenario - Human intrusion scenario Project Opalinus Clay, Switzerland (NAGRA 2002) - Reference scenario (groundwater transport) - Gas transport scenario - Human intrusion scenario - "What if?" scenarios - Alternative design scenarios - Alternative biosphere scenarios

1 Note that many of these scenarios also included variants or subscenarios.

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Table 4.2: Some potential scenarios or key factors that could define scenarios.

Possible Scenario

Included in TCS?

Comments TCS FEP# 1

Reference repository

Yes In the Base Scenario, the repository behaves as designed. 1.1

Earth-quakes

No The Canadian Shield is relatively quiescent, and analyses of earthquake scenarios as part of the similar SKB repository design have not indicated that the risk is significant. It is recognized that this would be subject to more detailed analysis in any site-specific assessment.

1.2.03

Global warming

No The effect of global warming on repository performance is not expected to be strong because of the repository depth.

1.3.01

Glaciation No Glaciation affects surface properties but geological evidence indicates that its influence on conditions at depth in the Canadian Shield has been limited. Specific analysis of glaciation effects is out of the scope of the present study.

1.3.05

Deliberate human intrusion

No We assume that any society both interested and capable of recovering materials from the repository would also be able to understand and avoid the hazards.

1.4.02

Inadvertent human intrusion

Yes This scenario is included since it cannot be ruled out and could lead to large consequences at least to a small number of people.

1.4.04

Meteor impact

No The meteor impact scenario was assessed in Wuschke et al. (1995). These results are still applicable; no further work is needed in the present study.

1.5.01

Defective Container

Yes This is the scenario considered most likely to cause some human impact.

2.1.03B

Container failure

Yes Failure of all containers will be considered as a "what if" sensitivity case within the Defective Container Scenario.

2.1.03E

Criticality No CANDU fuel is not enriched; used CANDU fuel cannot go critical.

2.1.14

Open borehole

No This scenario is out of the scope of the present study, but has not been found to be important in other studies.

1.1.08 1.4.04

Defective barriers

Yes These will be considered as part of the “what if” sensitivity studies for the Defective Container Scenario.

1.1.08

Hunter/ gatherer diet

No The characteristics of the self-sufficient farmer considered in the TCS are selected so as to maximize dose impact. Previous studies have shown that realistic lifestyle variations do not result in significantly larger dose impacts. It is recognized that this factor may require analysis in any real siting study.

2.4.03B

Alternative ground-water flows

Partial The scope of the present study is limited to consideration of one fracture network. However, the effect of a range in bulk rock permeability values is considered.

2.2.02 2.2.04

Urban community

No Urban residents would be less exposed to radionuclides from the repository because, for example, the food and water they consume would likely come from uncontaminated sources.

2.4.10

1 TCS Features, Events and Processes (2004)

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5. MODELS, CODES AND DATA 5.1 OVERVIEW The Third Case Study postclosure safety assessment considers three main scenarios: Base Scenario, Defective Container Scenario and Human Intrusion Scenario. Each of these can be described by a conceptual model, which is then analyzed or supported by a number of computer models or "codes". Details on each scenario, and its representation by mathematical and computer models, is provided in subsequent sections and the references therein. In this section, we provide an overview of the computer models, data and the quality assurance approach. Figure 5.1 summarizes the main computer models used in the Third Case Study. More details on these codes (theory, capabilities, limitations) and their documentation is provided in the TCS Reference Data and Codes report (2004). In brief, as illustrated by Figure 5.1, used fuel inventory, engineering design and site information are used in conjunction with computer models to develop a site-specific system description. In particular, the site characterization data are converted into a site hydrogeological model using the FRAC3DVS code. With this system description, the issue of postclosure safety can be addressed first using simple system models. In particular, the RSM model uses a simple repository model to assess the importance of all radionuclides present in used fuel, and screen out all those that would have negligible impact in the event of radionuclide released from a (defective) container (Defective Container Scenario). Also, the HIM model addresses the inadvertent Human Intrusion Scenario using stylized exposure cases. A more detailed postclosure safety assessment of the repository is then provided using the FRAC3DVS transport model and the SYVAC3-CC4 system model, specifically for the Defective Container Scenario in which groundwater transport is the key pathway for release of radionuclides to the surface. FRAC3DVS provides three-dimensional finite-element or finite-difference modelling of radionuclide transport, in either equivalent-porous-medium or dual-porosity mode, in the vault and geosphere. It is interfaced with SYVAC3-CC4, which provides the radionuclide source term, i.e., the radionuclide release rate from a defective container, and calculates biosphere consequences based on the radionuclide fluxes into the biosphere. SYVAC3-CC4 includes submodels that represent the container, buffer/backfill seals, geosphere, and biosphere, with the coupling as indicated in Figure 5.1. In addition, the assessment is supported by specialized models, such as MPM for calculation of UO2 dissolution rates and MOTIF for coupled thermal-hydraulic-mechanical analysis, that provide background information on specific processes. 5.2 SOFTWARE QUALITY ASSURANCE The software for use in the safety assessment of a deep geologic repository has been acquired, developed and maintained by OPG in a manner consistent with OPG's software quality assurance system.

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In addition to the main safety assessment software, the reference datasets, software tools (e.g., pre- and post-processors) and supporting research codes are also included within the software quality assurance procedure, although differing levels of documentation and testing apply to the different categories of software, depending on their impact on the final results. A more complete description of the software quality assurance system is provided in the TCS Reference Data and Codes report (2004).

Figure 5.1: General relationship between data and models in the Third Case Study. The models ranged from simple models (RSM), to detailed transport models (FRAC3DVS), to integrated system models (SYVAC3-CC4). These were supported by specialized codes such as ORIGEN-S, MPM and MOTIF that were used separately from this study to understand processes and develop data.

COMPASS,CCM...

FRAC3DVS(flow)

Groundwaterflow

Site characterization- geosphere data

Site characterization- biosphere data

Repository layout,canister design data

ORIGEN-S

Initialnuclideinventories

RSM, HIM, simple models

SYVAC3-CC4

vault geosphere biosphereGW flow

Calculatedimpacts on

humans and biota

Nuclide flow

Well demand

Nuclide flow

MOTIF...

Screened listof nuclides

MPM,CCM,PHREEQE...

Corrosion rates,solubilities

Fuel and burnupdata

FRAC3DVS (transport)

fuel/container

GW chemistry

Nuclide flow

Saturation time,redox conditions

Temperatures,glaciation effects

Nuclide flowNuclide flow

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6. BASE SCENARIO 6.1 SCENARIO DESCRIPTION A deep geologic repository differs from most engineered systems in that it is designed to perform its function over timeframes that will exceed that of recorded history and will do so without the need for institutional control. Various changes will occur to such a system during its lifetime, in some cases rapidly and in other cases very slowly. The design of the repository takes into account these processes insofar as they are expected to be important to the robustness of the repository long-term safety. The Base Scenario describes what are thought to be reasonable expectations for the evolution of the Third Case Study repository. In this Base Scenario, the various components of the repository perform as expected. In particular, the wastes are emplaced in containers that have been carefully engineered, as part of the repository system, to remain intact over timeframes in excess of 100,000 years. It is important to understand the Base Scenario since it represents the design basis. That is, the Base Scenario describes what would likely happen to the majority of the containers. The other scenarios described in later chapters consider alternatives where things don't go as planned, but which likely affect only a small number of containers. The description provided by the Base Scenario puts into perspective the overall behaviour of the repository and the surrounding geosphere, and is of value to these other scenarios because it describes the expected background conditions. 6.2 CONCEPTUAL MODEL The conceptual model for the Base Scenario is illustrated via a simple process influence diagram in Figure 6.1. Essentially, the model considers the used fuel, the container, the buffer/backfill seals, the geosphere and the surface biosphere. The main components of the model and the processes considered are summarized in Figure 6.1. Other smaller effects are also considered but not shown for clarity in Figure 6.1.

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Figure 6.1: General conceptual model showing the main processes considered in the Base Scenario.

SurfaceBiosphere

Geosphere

EngineeredBarriers

Containers

Used FuelBundles

- Radioactive decay and ingrowth- Fission gas production- Degradation of fuel and cladding

- Heat transfer- Radiation- Corrosion of fuel

- Stress and strain- Metal creep

- Corrosion of container- Microbial activity near container- Heat transfer- Swelling and hydrostatic loads on container

- Saturation and moisture redistribution- Swelling- Porewater composition changes- Oxygen consumption- Mineralogical changes

- Heat transfer- Groundwater interaction with seal materials- Initial drawdown effects into unsaturated vault

- Surface water recharge and discharge- Heat transfer- Glaciation

- Fracture development- Consumption of oxygen from repository and from surface waters

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6.3 ANALYSIS The following summarizes the main events in the evolution of the repository. This summary is taken from the description of the Base Scenario provided in McMurry et al. (2003), which includes justification, references and discussion of uncertainties. 0-100 years The repository is expected to be open and actively monitored for a period of about 100 years. This consists of about 30 years of operation, during which containers are emplaced and rooms are sealed, and 70 years of post-emplacement monitoring, during which access tunnels are kept open. In the operation period, approximately 3.6 million bundles of used fuel, within approximately 11,000 containers, are emplaced over a repository area of about 2 km2. The initial thermal output of the repository as a whole is about 13 MW, and the initial radioactivity is 1020 Bq. The repository contains about 70,000 Mg of uranium and 8,000 Mg of zirconium metal (in used-fuel bundles) in addition to 107,000 Mg of metallic iron and 45,000 Mg of copper (in containers). During the first 100 years after emplacement of the containers: • Radioactivity drops by a factor of ten; radionuclides with short half-lives such as tritium (H-3)

and Co-60 decay to negligible levels (see Figure 6.2). • Container thermal power drops by a factor of four (see Figure 6.3). • Peak temperatures are reached within the repository (values less than or equal to 100oC at

the container outer surface and less than 200oC inside the containers). • The copper container surface reacts with oxygen to form a thin corrosion layer. • Any residual oxygen and water inside the container react, mostly with the steel inner vessel,

to form a thin layer of iron oxide on the steel surface. • The buffer material near the containers dries out due to the heat emitted by the containers,

forming shrinkage cracks in the clay. The moisture driven outwards by this desiccation process recondenses in a cooler region at some distance from the container, possibly in the backfill near the rock.

• Mineral salts from the groundwater precipitate by evaporation in the region of desiccated clay around the containers.

• Microbes in the buffer material near the containers die or become dormant because of heat, desiccation, and lack of nutrients.

• Thermal expansion and contraction of the rock and concrete combine to create near-field stresses within the low-permeability rock and the concrete bulkhead at the ends of the emplacement rooms, and a limited amount of microcracking occurs.

• In the rock around the repository, groundwater flow and heads are influenced by the presence of the open tunnels and the high-suction clay, which draw water towards the repository. This is countered by thermal convection, which redistributes water away from the containers.

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Figure 6.2: Radioactivity of key nuclides in used CANDU fuel after discharge from reactor (220 MWh/kgU burnup). Natural uranium (mostly U-238 and its decay chain) dominates after about 1 million years. Figure 6.3: Total thermal power of used fuel after discharge from reactor (220 MWh/kgU burnup). The power is similar to the natural geothermal flow through the repository area after about 30,000 years. After about 1 million years, the residual power is that due to the decay of natural uranium (mostly U-238 and its decay chain).

0.1

1

10

100

1,000

10,000

100,000

10 100 1,000 10,000 100,000 1,000,000 10,000,000Time [a]

Ther

mal

pow

er [k

W].

Geothermal flux range

All containers

All containers - U-238 chain

One container

0.0001

0.001

0.01

0.1

1

10

100

1000

10000

10 100 1,000 10,000 100,000 1,000,000 10,000,000

Time [a]

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ioac

tivity

[GBq

/kgU

]

Total

Pu total

Sr-90

Tc-99 U-238 chain

Zr-93 +Nb-93m

I-129

Cs-135

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100-1000 years At the beginning of this time period, as part of the closing the repository, all access shafts and tunnels are backfilled and sealed, and all intrusive monitoring systems and deep boreholes are removed or closed. For several hundred years thereafter, distinct physical and chemical differences (e.g. temperature, porewater composition) will exist between the various components of the repository, and between the repository and the geosphere. Many of the changes that occur within this time frame are driven by these gradients. During this period: • Radioactivity drops by a factor of 30. Most fission products decay, including nuclides such

as Sr-90 and Cs-137. • Container thermal power drops to around 120 W per container. Residual heat comes from

the decay of the remaining actinides. • The oxygen initially present in the sealing materials (as trapped air) is consumed by

inorganic processes and by surviving microbes in regions away from the containers. • Groundwater from the far-field geosphere enters the repository, encountering the outermost

sealing materials first. As the clay layers become fully saturated, they start to swell and exert pressure on adjacent less-saturated components. The swelling process proceeds slowly, with the local development of some moderate non-uniform loads depending on where water access is easiest. The loads become more evenly developed across the components as saturation increases. Peak swelling loads are less than 5 MPa.

• The mechanical loads from the rock are transmitted through the expanding and swelling clays onto the container. The copper shell is compressed onto the inner steel vessel, which is rigid and maintains its shape.

• By the end of this time, the repository is fully saturated and anaerobic. • Where water is in contact with the container, some corrosion pits form in the copper at

specific locations, reaching depths of several mm at most (the copper shell is 25 mm thick). • Global warming, if it occurs, will have changed surface precipitation patterns, but deep

groundwater movement is unaffected. 1,000-10,000 years The next nine thousand years or so is a time in which the previously sharp gradients slowly diminish.