The Likelihood of Criticality: Synthesis report

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Technical Report Likelihood of Criticality

The Likelihood of Criticality Synthesis Report

T.W. Hicks and T.D. Baldwin, Galson Sciences Ltd

AMEC Report Reference 17293-TR-023

Client Reference RWMD/003/001

Client Name NDA RWMD

Issue Number Version 2

Report Date 17 March 2014


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Conditions of Publication

This report has been prepared by AMEC Nuclear UK Ltd under contract to the NDA. The report has been reviewed by the NDA, but the views expressed and conclusions drawn are those of the authors and do not necessarily represent those of the NDA.

This report is made available under the NDA Transparency Policy. In line with this policy, the NDA is seeking to make information on its activities readily available, and to enable interested parties to have access to and influence on its future programmes. The report may be freely used for non-commercial purposes. However, all commercial uses, including copying and re-publication, require permission from the NDA. All copyright, database rights and other intellectual property rights reside with the NDA. Applications for permission to use the report commercially should be made to the NDA Information Manager.

Although great care has been taken to ensure the accuracy and completeness of the information contained in this publication, the NDA cannot assume any responsibility for consequences that may arise from its use by other parties.

© Nuclear Decommissioning Authority 2014. All rights reserved.

Bibliography If you would like to see other reports available from the NDA, a complete listing can be viewed at our website www.nda.gov.uk, or please write to the Communications Department at the address below.

Feedback Readers are invited to provide feedback to the NDA on the contents, clarity and presentation of this report and on the means of improving the range of the NDA reports published. Feedback should be addressed to:

Dr Elizabeth Atherton Head of Stakeholder Engagement and Communications Nuclear Decommissioning Authority Radioactive Waste Management Directorate Building 587 Curie Avenue Harwell Oxford Didcot Oxfordshire OX11 0RH

Email: [email protected]


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Project Team

Work on the NDA RWMD’s Likelihood of Criticality project has been undertaken by a team of organisations that together have substantial experience of the UK programme for management of higher-activity radioactive wastes and of criticality safety concerns associated with geological disposal of these wastes.

AMEC is responsible for overall project management. Galson Sciences Ltd is responsible for managing the technical work programme as well as leading technical activities. TerraSalus Ltd has provided technical specialist support to Galson Sciences Ltd.

The preparation of this report has been led by Galson Sciences Ltd.

TerraSalus LtdTerraSalus Ltd


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Document Issue Record

Document title Likelihood of Criticality: The Likelihood of Criticality Synthesis Report

Project Reference 17293 Galson Sciences Reference 1124-27

Purpose of Issue Version 2

Security Class Not Protectively Marked Issue Description Originator Checker Approver Date

dd/mm/yy

1 First draft (v1d1) for comment.

T.W. Hicks

T.D. Baldwin A. Guida

18/09/13

2 Final version to client (v1), issued in response to comments from Peter Wood, Robert Winsley, Lucy Bailey, Mike Poole and Simon Wisbey (RWMD).

T.W. Hicks

T.D. Baldwin A. Guida

05/11/13

3 Final version (v2) issued to client in response to peer review comments.

T.W. Hicks J.C. Smith A. Guida 17/03/14

Previous issues of this document shall be destroyed or marked SUPERSEDED


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Executive Summary

The Radioactive Waste Management Directorate (RWMD) is responsible for implementing geological disposal of the UK’s higher-activity radioactive wastes. RWMD’s research into geological disposal considers safety during waste transport to a disposal facility, during waste disposal operations, and once the facility has been closed. The wastes for disposal comprise a wide range of materials and include some fissile radionuclides. The presence of fissile materials requires an assessment of the potential for criticality.

After disposal, the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for such time as the waste packaging affords a high level of containment. However, as waste packages begin to degrade, fissile and other materials may be mobilised and this could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to lead to criticality requires consideration.

RWMD established the Likelihood of Criticality research project to develop, document and communicate the qualitative and quantitative arguments required to evaluate the probability of nuclear criticality after closure of a GDF. The evaluation of the likelihood of criticality has been underpinned by consideration of the features, events and processes (FEPs) that could affect nuclear reactivity after GDF closure. The FEP analysis led to the construction of post-closure criticality scenarios and these scenarios have been evaluated using a range of approaches, from high-level judgments about scenario credibility, to modelling of evolving conditions in a GDF to determine if critical systems could develop. Understanding the radioactive waste inventories, the GDF concepts for the different waste types and geological settings, and the expected evolution of conditions in the different GDF concepts, as well as associated uncertainties, has been of fundamental importance to the criticality FEP and scenario analysis.

The project has focused on the analysis of a GDF containing the Baseline Inventory of radioactive wastes and materials. The Baseline Inventory defined in the 2007 UK Radioactive Waste Inventory (UK RWI) has been assumed, and disposal concepts for intermediate-level waste and some low-level waste (ILW/LLW), depleted, natural and low-enriched uranium (DNLEU), spent nuclear fuel (SF), high-level waste (HLW), separated plutonium (Pu) and highly-enriched uranium (HEU) have been considered. The project has considered disposal concepts for these wastes in three generic host-rock settings (higher-strength rocks, lower-strength sedimentary rocks and evaporites). This report provides a summary of the approach taken to evaluate the likelihood of criticality for each disposal concept and the results of the analysis.

The most important fissile nuclides in LLW/ILW/DNLEU and SF/HLW/HEU/Pu are 239Pu and 235U, and their behaviour after disposal has been the focus of the Likelihood of Criticality project. The project has involved identification and assessment of GDF post-closure criticality scenarios involving rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package. These criticality scenarios could occur as a result of events and processes such as container breach due to corrosion, followed by uranium and plutonium dissolution, advection and sorption.


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The analysis of these post-closure scenarios has involved different levels of detail, depending largely on the fissile material contents of waste packages and the assumed properties of the host rock, as follows:

• For some waste types, the waste packages contain insufficient fissile material for criticality, even under the most favourable conditions that can be envisaged for criticality in the vicinity of the waste package. In this case, criticality at the package scale can be judged not to be credible and no further analysis of package-scale scenarios is required.

• Where such judgments cannot be made, more detailed analyses of waste package degradation and fissile material relocation have been undertaken. Such scenarios have been evaluated in order to determine the concentrations and masses of fissile material that could occur at potential accumulation sites. A probabilistic modelling approach using the GoldSim software package has been taken and the likelihood of criticality has been judged based on comparisons of the calculated fissile masses and concentrations with minimum values required for criticality at different locations.

Combinations of realistic and cautious assumptions have been made about waste package evolution and uranium and plutonium behaviour. These assumptions have been expressed in terms of parameters and their values, including probability distributions representing uncertainties, for the criticality scenario analysis. In the GoldSim calculations, the probability distributions have been sampled over many model runs in order to estimate the likelihood of critical concentrations or masses of fissile material developing after GDF closure.

Judgments about the conditions required for criticality in different components of a GDF (i.e., in waste packages, engineered barriers and host rock) have been key to this analysis. However, there are large uncertainties in the materials that might be involved in fissile material accumulation scenarios and the configurations of the accumulated material. Such uncertainties have been addressed by making bounding assumptions. For example, in many cases fissile material has been assumed to accumulate in optimal spherical configurations and potential neutron absorbing materials have been ignored. In some barrier components, such as a low porosity host rock, it can be shown that there is insufficient pore space for criticality irrespective of the mass of fissile material that might accumulate. However, for most barrier components, data on minimum critical masses and concentrations of fissile material have been used to judge whether critical systems develop.

The main conclusions from the analysis of LLW/ILW/DNLEU disposal in higher-strength rock, lower-strength sedimentary rock and evaporite are shown in Table ES.1. The main conclusions from the analysis of SF/HLW/HEU/Pu disposal are shown in Table ES.2. In some cases, the GoldSim calculations have shown that it is not possible to accumulate a critical mass or concentration of fissile material, conditional on the treatment of parameter value uncertainty and bounding assumptions about the requirements for criticality. In other cases, the modelling has shown that it is possible to accumulate a critical mass of fissile material with a certain probability. In these cases, it is possible to make qualitative judgments about the low likelihood of criticality, again conditional on the treatment of parameter value uncertainty. However, at no point have the different estimates of the probability of criticality by different processes at different locations in a GDF, and involving different types of waste, been combined to evaluate the overall likelihood of criticality in a GDF.


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Table ES.1: Main results of the analysis of post-closure criticality scenarios for LLW/ILW/DNLEU disposal.

Criticality Scenario Likelihood of Criticality for GDFs in Different Types of Host Rock

Higher-strength rock

Reactivity increase in a waste package

The package-scale model developed for a GDF in higher-strength rock consists of a single package of grouted waste surrounded by backfill in a disposal vault. The package may be breached by corrosion, after which time uranium and plutonium may be mobilised and migrate in flowing groundwater into the backfill. Grout may also be dissolved and removed in groundwater to the extent that solid plutonium and uranium could slump to the base of the package. Review of the UK RWI has enabled the package-scale analysis to focus on waste streams for which the waste packages contain on average the greatest amounts of fissile material. A small number of ILW and DNLEU packages may contain sufficient LEU, HEU or Pu for slumping to generate conditions that are critical. The waste packages for which in-package post-closure criticality may be possible based on the package-scale modelling results represent about 0.04% of around 202,000 unshielded ILW packages and about 1.25% of around 165,000 DNLEU waste packages in the UK RWI.

Accumulation outside a waste package

For situations in which the fissile material migrates to and accumulates in the backfill, critical conditions could occur for the ILW and DNLEU packages that contain the highest masses of LEU, HEU or Pu. However, the modelling results indicate that such accumulations are only possible in the extreme when the highest values of uranium solubility and sorption are selected from the relevant probability distributions. The waste packages for which post-closure criticality may be possible under such conditions represent about 0.15% of unshielded ILW packages and about 1.25% of DNLEU waste packages.

Accumulation from multiple waste packages

A multi-package model has been developed to evaluate the release, migration and mixing of fissile material from many waste packages in a disposal vault. The gravitational slumping of uranium and plutonium through a stack of degraded waste packages is modelled, as is the advection of fissile material from an array of waste packages and its downstream accumulation in GDF engineered barriers and host rock. The contents of each waste package in the modelled two-dimensional array of packages are determined based on random selection from the population of waste packages in the UK RWI. The analysis has found that criticality is not possible, except under conditions in which uranium and plutonium solubilities and sorption have extremely high values and the fissile material accumulates in backfill. Such conditions were achieved in about 2% of probabilistic model runs. The fissile material in all such accumulations can be traced back to the few waste packages in the UK RWI that have the highest average fissile material content.


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Criticality Scenario Likelihood of Criticality for GDFs in Different Types of Host Rock

Lower-strength sedimentary rock


For a GDF in lower-strength sedimentary rock, it has been assumed that there is no significant component of groundwater flow, so that the potential for material to be removed from waste packages in sufficient quantities for voids to be created and slumping to occur is small. Uranium concentrations in waste packages would gradually reduce by diffusion into the backfill following loss of package integrity and dissolution in groundwater, and the waste packages would remain sub-critical.


Uranium concentrations in the backfill would be lower than the concentrations in grout at the time of disposal and conditions would be expected to remain sub-critical.


As the system evolves and multiple waste packages fail and release uranium, uranium from many packages would mix and the uranium concentration will tend to reduce to an average for the inventory as a whole. Conditions would remain sub-critical.

Evaporite


For a GDF in evaporite, it has been assumed that there is no water present, so that there is no significant wasteform dissolution and no mechanism for mass transfer, other than as a result of the effects of rock creep. With limited quantities of neutron moderators present and neutron absorbing materials being retained in the waste packages, the waste packages will remain sub-critical.


In the absence of water, there is no potential for fissile material migration and accumulation. Conditions would remain sub-critical.


In the absence of water, there is no potential for fissile material migration and accumulation from multiple waste packages. Conditions would remain sub-critical.

Table ES.2: Main results of the analysis of post-closure criticality scenarios for SF/HLW/HEU/Pu disposal.

Wasteform Likelihood of Criticality for GDFs in Different Types of Host Rock


The GoldSim model developed for a GDF in higher-strength rock consists of a disposal tunnel above a number of vertical deposition holes, with each deposition hole containing a waste package. The waste packages are surrounded by bentonite and the tunnel is backfilled with a mixture of crushed rock and bentonite. The waste packages are copper canisters that contain HLW, SF, HEU or Pu. The containers may be breached by, for example, corrosion, after which time dissolution and migration of uranium and plutonium from the waste packages and into the buffer, tunnel backfill and host rock may occur. Mass transfer is assumed to be advection dominated. The uranium and plutonium released from the waste package may accumulate by sorption in the engineered barriers and host rock. However, based on the expected lifetime of a copper canister under disposal conditions, 239Pu is calculated to have almost entirely decayed to 235U before canister failure.


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HLW HLW packages include relatively small amounts of fissile material and accumulation of such material from many waste packages would be required for criticality. GoldSim calculations have indicated that, after loss of canister integrity, insignificant amounts of 235U are likely to migrate into any accumulation zone.

SF The Advanced Gas-cooled Reactor (AGR) SF and Pressurised Water Reactor (PWR) SF considered in this analysis are of sufficient burn-up that the fuel will always be sub-critical under natural-water-moderated conditions. The GoldSim calculations confirm that insufficient uranium for criticality would accumulate in the buffer surrounding a failed container. It is unlikely that large amounts of unirradiated or low burn-up fuel would require disposal. However, waste packages that include such fuel would exhibit higher reactivities than calculated in this analysis for post-closure criticality scenarios. In particular, waste packages that contain fresh or low burn-up PWR fuel may not be sub-critical when in-package post-closure scenarios are considered. However, the GoldSim calculations indicate that accumulations of uranium in the buffer surrounding the waste packages would not result in criticality even if the uranium derived from fresh fuel.

HEU/Pu Combined GoldSim and MCNP calculations indicate that failed HEU/Pu waste packages are sub-critical because of the neutron absorbing components of the wasteform. However, based on the calculated migration of 235U from HEU/Pu waste packages and its accumulation in the buffer over long periods (in excess of 107 years), criticality may be possible even based on average parameter values.


The GoldSim model developed for a GDF in lower-strength sedimentary rock consists of a horizontal disposal tunnel containing a number of waste packages surrounded by bentonite. The waste packages are carbon steel canisters that contain HLW, SF, HEU or Pu. The containers may be breached by corrosion, after which time dissolution and migration of uranium and plutonium from the waste packages and into the tunnel backfill and host rock may occur. Mass transfer is assumed to be diffusion dominated. The uranium and plutonium released from the waste package may accumulate by sorption in the backfill and host rock.

HLW Results of calculations for HLW are similar to those obtained for disposal in higher-strength rock. GoldSim calculations have found that, after loss of canister integrity, insignificant amounts of 235U are likely to migrate into any accumulation zone.

SF The GoldSim calculations confirm that insufficient uranium for criticality would accumulate in the backfill surrounding a failed container if credit is taken for burn-up. However, enough 235U for criticality could accumulate in the buffer if the uranium derives from fresh PWR fuel.

HEU/Pu Results of calculations for HEU/Pu are similar to those obtained for disposal in higher-strength rock. Criticality may be possible on a timescale of 106 years as a result of accumulation of fissile material in the backfill, even based on average parameter values.


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Evaporite

No modelling has been undertaken for disposal of SF/HLW/HEU/Pu in an evaporite host environment because the rock has been assumed to be dry. Under such conditions, there would be no significant wasteform dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. Creep of the host rock may reduce void space in waste packages, increasing fissile material concentrations, but neutron absorbing material will be retained, limiting the potential for criticality for all of the wasteforms.

The analysis of post-closure criticality scenarios has, of course, been based on many assumptions about the waste inventory, waste packaging concepts, disposal facility design, barrier system properties and GDF evolution. Key assumptions are as follows:

• Based on the assumed 2007 UK RWI, a number of waste streams will be packaged such that the packages have a high fissile material content. However, few of the waste packaging concepts where the fissile material content is high have been assessed by RWMD through application of the Letter of Compliance disposability assessment process. Also, alternative approaches to waste packaging (e.g., involving vitrification, non-encapsulation or immobilisation in polymers) have been and may be proposed by waste producers for some wastes. Other packaging concepts are still under development, such as for HEU and Pu disposal. The methodologies and tools developed in the project could be used or developed to support future assessments of waste packaging proposals based on understanding of the expected behaviour of the waste packages under disposal conditions.

• The Likelihood of Criticality analysis has been undertaken based on average waste package inventories. Large variations about the average fissile material content of waste packages for any one waste stream could affect the conclusions of the criticality scenario analysis.

• Uniform chemical conditions have been assumed in the GoldSim calculations. Spatially variable chemical conditions could result in regions in which higher concentrations of fissile material occur than have been calculated in this analysis. However, the current calculations using generic sorption parameters are considered to bound the effects of precipitation. Also, potential concentrations of natural uranium in groundwater have not been considered. Assuming groundwater to include concentrations of natural uranium would reduce the rate of release of uranium from waste packages and the enrichment of dissolved uranium in the GDF would be lower than calculated in this analysis.

• Slumping of fissile material through voids in LLW/ILW/DNLEU disposal vaults has been assumed to occur on timescales of several hundred thousand years to hundreds of millions of years following dissolution of waste encapsulation grout. However, project work to improve understanding of GDF evolution has found that there may be little potential for void formation under expected geochemical and hydrological conditions. This is a significant finding because it means that assumptions made in this project about grout behaviour may be cautious rather than realistic and post-closure criticality scenarios involving fissile material slumping on the package scale or disposal vault scale may not be credible.

• Many waste packages contain significant quantities of iron, and iron corrosion products have largely been assumed to be dissolved and removed from corroded


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waste packages, which may be cautious. However, analysis of the behaviour of iron corrosion products under disposal conditions has indicated that iron would remain in solid form for long periods. The persistence of iron may significantly reduce the likelihood of criticality within a waste package.

• The waste packages in LLW/ILW/DNLEU disposal vaults have been assumed to be randomly distributed. If similar waste packages with high fissile material content are emplaced together in a vault, then this may also affect the conclusions of the criticality scenario analysis.

• SF/HLW/HEU/Pu canisters have been assumed to contain wastes for on the order 105 years or more. Canister failure prior to significant 239Pu decay would result in higher calculated reactivities. In particular, if early failure of copper canisters is deemed to be possible, this could have an important impact on the results of the Likelihood of Criticality project for the higher-strength rock disposal concept.

In conclusion, the Likelihood of Criticality research project has provided qualitative and quantitative arguments about the probability of criticality after closure of a GDF, based on assumptions about GDF evolution and, in particular, uranium and plutonium behaviour. The project has provided results and tools that can be used to support a demonstration, as part of an environmental safety case, that the possibility of a local accumulation of fissile material in a GDF such as to produce a neutron chain reaction is not a significant concern, thereby meeting regulatory requirements on GDF post-closure criticality safety.


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List of Terms and Abbreviations

AGR Advanced Gas-cooled Reactor ALARA As Low As Reasonably Achievable ALARP As Low As Reasonably Practicable AWE Atomic Weapons Establishment BFS Blast Furnace Slag BWR Boiling Water Reactor CDF Cumulative Distribution Function CfA Conditions for Acceptance CoRWM Committee on Radioactive Waste Management CSA Criticality Safety Assessment DBE Deutsche Gesellschaft zum Bau und Betrieb von Endlagern für Abfallstoffe mbH DCI Ductile Cast Iron DCTC Disposal Canister Transport Container DECC Department for Energy and Climate Change Defra Department for the Environment, Food and Rural Affairs DfT Department for Transport DNLEU Depleted, Natural and Low-enriched Uranium DOE (United States of America) Department of Energy DSFS Disposal System Functional Specification DSSC Disposal System Safety Case DSS Disposal System Specification DSTS Disposal System Technical Specification DU Depleted Uranium EA Environment Agency (for England and Wales) EBS Engineered Barrier System EC European Commission EDZ Excavation Disturbed Zone ESC Environmental Safety Case EU European Union FEP Features, Events and Processes GCSA General Criticality Safety Assessment GDF Geological Disposal Facility GLEEP Graphite Low Energy Experimental Pile GRA Guidance on Requirements for Authorisation GSL General Screening Level


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GWPS Generic Waste Package Specification HEU Highly Enriched Uranium HLW High-level Waste HSE Health and Safety Executive IAEA International Atomic Energy Agency ICRP International Commission on Radiological Protection ILW Intermediate-level Waste ILW-HEU Intermediate-level Waste Containing Highly Enriched Uranium ILW-INU Intermediate-level Waste Containing Irradiated Natural and Slightly Enriched

Uranium ILW-LEU Intermediate-level Waste Containing Low Enriched Uranium ILW-Pu Intermediate-level Waste Containing Separated Plutonium IP Industrial Package IRF Instant Release Fraction KBS-3 Kärnbränslesäkerhet-3 (Nuclear Fuel Safety #3) LEU Low-enriched Uranium LLW Low-level Waste LLWR Low Level Waste Repository LoC Letter of Compliance LSA Low Specific Activity LSL Lower Screening Level MBGWS Miscellaneous Beta Gamma Waste Store MCNP Monte Carlo N-Particle Code MoD Ministry of Defence MRWS Managing Radioactive Waste Safely NDA Nuclear Decommissioning Authority NEA Nuclear Energy Agency (of the OECD) NIEA Northern Ireland Environment Agency NRC (United States) Nuclear Regulatory Commission NRVB Nirex Reference Vault Backfill OCNS Office for Civil Nuclear Security OECD Organisation for Economic Co-operation and Development ONR Office for Nuclear Regulation OPC Ordinary Portland Cement ORR Office of Rail Regulation OSC Operational Safety Case PCM Plutonium Contaminated Material PCSA Post-closure Safety Assessment PDF Probability Density Function


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PFA Pulverised Fuel Ash PHE Public Health England PWR Pressurised Water Reactor R&D Research and Development RSA 93 Radioactive Substances Act 1993 RWMD Radioactive Waste Management Directorate SAPs (Health and Safety Executive) Safety Assessment Principles SCO Surface Contaminated Object SEPA Scottish Environment Protection Agency SF Spent Fuel SFAIRP So Far As Is Reasonably Practicable SGHWR Steam-Generating Heavy Water Reactor SILW Shielded Intermediate Level Waste SKB Svensk Kärnbränslehantering AB (Swedish Nuclear Fuel and Waste

Management Company) SWTC Standard Waste Transport Container THORP Thermal Oxide Reprocessing Plant TSC Transport Safety Case UCuRC Understanding Criticality under Repository Conditions UILW Unshielded Intermediate Level Waste UK RWI UK Radioactive Waste Inventory USL Upper Screening Level WAC Waste Acceptance Criteria WAGR Windscale Advanced Gas-cooled Reactor WEP Waste Encapsulation Plant WIPP Waste Isolation Pilot Plant WVP Waste Vitrification Plant YMR Yucca Mountain Repository


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Contents

1 Introduction .......................................................................................................... 1 1.1 Background ................................................................................................................. 1 1.2 Objectives and Work Programme ............................................................................... 3 1.3 Approach to Understanding the Likelihood of Criticality .............................................. 6 1.4 Report Structure .......................................................................................................... 7

2 Assessment Methodology ................................................................................... 8 2.1 Wastes and Disposal Concepts .................................................................................. 8 2.2 Criticality FEPs and Scenarios .................................................................................. 10 2.3 Criticality Scenario Analysis ...................................................................................... 11 2.4 Overseas Approaches to Evaluating the Likelihood of Criticality in GDFs ................ 13

2.4.1 Criticality Safety Standard for SF Disposal in Germany .................................. 13 2.4.2 Criticality Safety Assessment for Transuranic Waste Disposal in the US ....... 14 2.4.3 Criticality Safety Assessments for SF/HLW/HEU/Pu Disposal in the US ........ 15 2.4.4 Criticality Safety Assessment for SF Disposal in Sweden ............................... 15 2.4.5 Summary ......................................................................................................... 16

3 UK Radioactive Waste Packaging Concepts and Inventories ........................ 17 3.1 Introduction ................................................................................................................ 17 3.2 LLW/ILW Packaging Concepts and Inventories ........................................................ 18

3.2.1 LLW/ILW Packaging ........................................................................................ 18 3.2.2 LLW/ILW Inventory .......................................................................................... 19

3.3 DNLEU Packaging Concepts and Inventories ........................................................... 28 3.3.1 DNLEU Packaging .......................................................................................... 28 3.3.2 DNLEU Inventory ............................................................................................ 28

3.4 HLW Packaging Concepts and Inventories ............................................................... 29 3.4.1 HLW Packaging ............................................................................................... 29 3.4.2 HLW Inventory ................................................................................................. 30

3.5 Spent Fuel Packaging Concepts and Inventories ..................................................... 33 3.5.1 Spent Fuel Packaging ..................................................................................... 33 3.5.2 Spent Fuel Inventory ....................................................................................... 34

3.6 Pu and HEU Packaging Concepts and Inventories ................................................... 37 3.6.1 Pu and HEU Packaging ................................................................................... 37 3.6.2 Pu and HEU Inventory ..................................................................................... 37

4 Likelihood of Criticality in a GDF in Higher-strength Rock ............................ 40 4.1 Introduction ................................................................................................................ 40 4.2 Illustrative Disposal Concept for Higher-strength Rock ............................................. 40 4.3 GDF Post-closure Criticality FEPs and Scenarios – LLW/ILW/DNLEU ..................... 43

4.3.1 GDF Evolution ................................................................................................. 43


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4.3.2 Criticality FEPs and Scenarios ........................................................................ 46 4.4 GDF Post-closure Criticality Scenario Assessment – LLW/ILW/DNLEU ................... 48

4.4.1 Single Package Model ..................................................................................... 48 4.4.2 Reactivity Increase in a Waste Package ......................................................... 50 4.4.3 Accumulation outside a Waste Package ......................................................... 55 4.4.4 Multiple Package Model .................................................................................. 56 4.4.5 Accumulation from Multiple ILW Packages ..................................................... 61

4.5 Summary of Assumptions – LLW/ILW/DNLEU ......................................................... 65 4.6 GDF Post-closure Criticality FEPs and Scenarios – SF/HLW/HEU/Pu ..................... 70

4.6.1 GDF Evolution ................................................................................................. 70 4.6.2 Criticality FEPs and Scenarios ........................................................................ 71

4.7 GDF Post-closure Criticality Scenario Assessment – SF/HLW/HEU/Pu ................... 74 4.7.1 GoldSim Model ................................................................................................ 74 4.7.2 MCNP Model ................................................................................................... 76 4.7.3 Modelling Results ............................................................................................ 77

4.8 Summary of Assumptions – SF/HLW/HEU/Pu .......................................................... 84 5 Likelihood of Criticality in a GDF in Lower-strength Sedimentary Rock ...... 89

5.1 Introduction ................................................................................................................ 89 5.2 Illustrative Disposal Concept for Lower-strength Sedimentary Rock......................... 89 5.3 GDF Post-closure Criticality FEPs and Scenarios – LLW/ILW/DNLEU ..................... 92


5.4 GDF Post-closure Criticality Scenario Assessment – LLW/ILW/DNLEU ................... 95 5.4.1 Reactivity Increase in a Waste Package ......................................................... 95 5.4.2 Accumulation outside a Waste Package ......................................................... 95 5.4.3 Accumulation from Multiple Waste Packages ................................................. 96

5.5 Summary of Assumptions – LLW/ILW/DNLEU ......................................................... 96 5.6 GDF Post-closure Criticality FEPs and Scenarios – SF/HLW/HEU/Pu ..................... 97


5.7 GDF Post-closure Criticality Scenario Assessment – SF/HLW/HEU/Pu ................. 100 5.7.1 GoldSim Model .............................................................................................. 100 5.7.2 MCNP Model ................................................................................................. 102 5.7.3 Modelling Results .......................................................................................... 102

5.8 Summary of Assumptions – SF/HLW/HEU/Pu ........................................................ 108 6 Likelihood of Criticality in a GDF in Evaporite .............................................. 109

6.1 Introduction .............................................................................................................. 109 6.2 Illustrative Disposal Concept for Evaporite .............................................................. 109 6.3 GDF Post-closure Criticality FEPs and Scenarios – LLW/ILW/DNLEU ................... 112

6.3.1 GDF Evolution ............................................................................................... 112 6.3.2 Criticality FEPs and Scenarios ...................................................................... 112


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6.4 GDF Post-closure Criticality Scenario Assessment – LLW/ILW/DNLEU ................. 113 6.5 Summary of Assumptions – LLW/ILW/DNLEU ....................................................... 113 6.6 GDF Post-closure Criticality FEPs and Scenarios – SF/HLW/HEU/Pu ................... 114

6.6.1 GDF Evolution ............................................................................................... 114 6.6.2 Criticality FEPs and Scenarios ...................................................................... 115

6.7 GDF Post-closure Criticality Scenario Assessment – SF/HLW/HEU/Pu ................. 115 6.8 Summary of Assumptions – SF/HLW/HEU/Pu ........................................................ 116

7 Summary and Conclusions ............................................................................. 117 7.1 LLW/ILW/DNLEU Disposal ...................................................................................... 117 7.2 HLW/SF/HEU/Pu Disposal ...................................................................................... 119 7.3 Key Assumptions ..................................................................................................... 121 7.4 Conclusions ............................................................................................................. 122

8 References ........................................................................................................ 123


Not Protectively Marked NDA RWMD 17293-TR-023 Page | 1The Likelihood of Criticality Synthesis Report

1 Introduction This report presents a summary of the approach and findings of a project undertaken to evaluate the likelihood of criticality in a geological disposal facility for the UK’s radioactive wastes. Detailed findings of the project are presented in separate reports on:

• The likelihood of criticality following disposal of intermediate-level waste, some low-level waste, and depleted, natural and low-enriched uranium [1].

• The likelihood of criticality following disposal of spent nuclear fuel, high-level waste, separated plutonium and highly-enriched uranium [2].

The background, objectives and approach to the project are set out below.

1.1 Background The Nuclear Decommissioning Authority (NDA), through its Radioactive Waste Management Directorate (RWMD), is responsible for implementing UK Government policy for long-term management of higher-activity radioactive wastes. Government policy for geological disposal of higher-activity radioactive waste, preceded by safe and secure interim storage, is set out in the Managing Radioactive Waste Safely (MRWS) White Paper [3]1, along with details of the radioactive waste inventory (the Baseline Inventory) that requires disposal.

At the present stage in the process of implementing geological disposal, no site has been identified for a disposal facility. In order to manage the current uncertainty concerning a geological environment for the disposal facility, RWMD is considering three generic host-rock types (higher-strength rocks, lower-strength sedimentary rocks and evaporites) and a range of generic disposal concepts. Different types of disposal concept need to be considered for different geological settings. However, all of the concepts are based on a multi-barrier approach, which involves engineered and natural barriers working together to prevent radioactivity being released to the surface environment in amounts that could cause harm to life and the environment.

The Baseline Inventory in the MRWS White Paper is based on the 2007 UK Radioactive Waste Inventory (UK RWI) [4]. The wastes identified for geological disposal in the White Paper include high-level waste (HLW), intermediate-level waste (ILW), both shielded and unshielded ILW (SILW/UILW) and some low-level waste (LLW) unsuitable for near-surface disposal. The Baseline Inventory excludes LLW that is identified in the White Paper as being disposable in near-surface facilities [5]. However, the Baseline Inventory does include some nuclear materials that have not been declared as wastes by their owners, but which might be declared as wastes in the future, namely spent nuclear fuel (SF), separated plutonium (Pu) and uranium (U). The uranium comprises both highly enriched uranium (HEU) and depleted, natural and low-enriched uranium (DNLEU). The Government’s policy [3, para.3.8] is that, pending a decision on whether these additional radioactive materials included in the Baseline Inventory should be declared as waste, RWMD should

1 Government is currently reviewing the site-selection aspects of the MRWS programme. NDA’s

programme for implementing geological disposal will be updated to reflect any changes made in the MRWS process.



factor their possible inclusion into the design and development of a geological disposal facility (GDF).

RWMD’s research considers safety during waste transport to a disposal facility, during waste disposal operations, and once the facility has been closed (the post-closure phase). The wastes for disposal comprise a wide range of materials and include some fissile radionuclides; that is, on interaction with slowly moving neutrons, their nuclei can fission (split) into lighter nuclei. The uranium isotope 235U and the plutonium isotope 239Pu are the main fissile nuclides present in radioactive wastes. During a fission event, energy and further neutrons are released. If sufficient fissile material is present under certain conditions, a criticality excursion (i.e., a nuclear chain reaction) may occur, with neutrons released by fission leading to fission of other nuclei. Criticality can release large amounts of radiation and energy and, therefore, criticality controls need to be imposed to ensure safety during radioactive waste management operations.

RWMD has developed fissile material limits and other controls for a range of waste packages containing different types of waste to ensure that a criticality event cannot occur during waste transport and that the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during storage and emplacement operations. These limits also ensure that criticality is prevented for such time as the waste packaging affords a high level of containment after disposal. Alternative approaches to waste packaging have been and may be proposed by waste producers. These are assessed by RWMD through application of the Letter of Compliance (LoC) disposability assessment process for compliance with RWMD waste package specifications and compatibility with the eventual needs for transport to and disposal in a GDF [6, §6.1].

After disposal, as waste packages and GDF barrier materials begin to degrade, fissile and other materials may be mobilised and this could affect the potential for criticality. For example, fissile material could migrate out of waste packages in groundwater and accumulate in new configurations. Any resultant criticality event has the potential to adversely affect the performance of a GDF by affecting engineered and natural barriers. Therefore, the potential for the evolution of conditions in a GDF to lead to criticality and associated adverse effects on barrier system performance requires consideration.

The regulatory requirement to consider post-closure criticality in a GDF is set out in the environment agencies’ Guidance on the Requirements for Authorisation (GRA) for geological disposal facilities [7, §7.3.31]:

“If significant amounts of fissile material are being disposed of at the facility, the developer/operator will need to demonstrate as part of the environmental safety case that the possibility of a local accumulation of fissile material such as to produce a neutron chain reaction is not a significant concern. The environmental safety case should also investigate, as a ‘what-if’ scenario, the impact of a postulated criticality event on the performance of the disposal system.”

RWMD’s approach to addressing post-closure criticality safety is to provide a demonstration that the likelihood and consequences of post-closure criticality are low (i.e., that it is not possible or is very unlikely for a critical mass or concentration of fissile material to accumulate, but if a critical accumulation did occur, the consequences would be insignificant). Work was undertaken over many years through RWMD’s programme of research on ‘Understanding Criticality under Repository Conditions’ (UCuRC). The aim of the programme was to obtain a better understanding of the processes that would control the



nature and magnitude of criticality under the particular conditions expected in a GDF. The results of this work were summarised in 1998 in a topical report on post-closure criticality safety [8]. Complementary studies aimed to develop an understanding of the likelihood of post-closure criticality based on consideration of processes that could result in accumulation of fissile material in disposal facilities for LLW/ILW in higher-strength rock [9] and in alternative geological environments (lower-strength sedimentary rock and evaporites) [10]. Work to understand the possible consequences of post-closure criticality was also undertaken [11]. RWMD’s work on the consequences and likelihood of criticality was summarised in the 2010 Criticality Safety Status Report [12].

In 2011, RWMD initiated a project on the likelihood of criticality in collaboration with AMEC, Galson Sciences Ltd and TerraSalus Ltd. The Likelihood of Criticality project has developed the approach taken in the previous criticality likelihood studies and extended the scope of the analysis to consideration of the likelihood (or probability) of post-closure criticality associated with the disposal of LLW (unsuitable for near-surface disposal), ILW, HLW, SF, Pu and U in each of the three generic geological settings. In parallel, RWMD commissioned a programme of work on modelling the consequences of hypothetical criticality in a GDF [13], which built on the work undertaken in the UCuRC research programme.

1.2 Objectives and Work Programme The overall objective of the Likelihood of Criticality project has been to develop, document and communicate the qualitative and quantitative arguments required to evaluate the probability of nuclear criticality after closure of a GDF for the UK’s higher-activity radioactive wastes. The approach to the project was set out in the form of a Roadmap, which defined the work programme in four sequential stages with associated activities. ‘Gates’ between stages marked decision points at which decisions about the work to be undertaken in the next stage were taken and, if necessary, the Roadmap was updated. The four stages of work are summarised in Figure 1.1 and activities undertaken during each stage are indicated in Figure 1.2. The activities undertaken in the first three stages of the project led to the production of a series of technical notes, which formed the basis of the reports produced in the final stage of the work.

A main focus of the work was the need to ensure effective communication of the project results. To this end, during Stage 3, a workshop on communicating project findings was held in conjunction with the parallel project on the consequences of hypothetical criticality. The workshop included attendees from the Likelihood of Criticality project, the Consequences of Hypothetical Criticality project, RWMD, NDA, EA NWAT (Environment Agency Nuclear Waste Assessment Team), ONR (Office for Nuclear Regulation), and independent peer reviewers for both research programmes. The workshop provided an opportunity to discuss and obtain feedback on project results and presentational methods prior to the preparation of the project reports in Stage 4. The final project reports are as follows:

• A report on the results of analyses to evaluate the likelihood of criticality in a GDF for LLW/ILW/DNLEU [1];

• A report on the results of analyses to evaluate the likelihood of criticality in a GDF for SF/HLW/HEU/Pu [2]; and

• A synthesis report that summarises the work presented in the above two reports (this report).



Figure 1.1: Stages in the Likelihood of Criticality project.

Stage 1: Scope and Knowledge Base Stage 1 began with preparation of descriptions of waste packaging and GDF concepts, including the evolution of conditions after GDF closure. Work to address knowledge gaps was specified. The features, events and processes (FEPs) that could act together in an evolving disposal facility environment to lead to conditions required for criticality were considered and credible post-closure criticality scenarios were constructed. A methodology for estimating the likelihood of criticality was outlined and agreed.

Stage 2: Preliminary Scenario Assessment Stage 2 involved the development of a preliminary database of FEP values (including representations of uncertainty) in support of initial work to develop qualitative and quantitative arguments about the likelihood of criticality. Work to address FEP knowledge gaps and to develop a probabilistic GoldSim model was undertaken.

Stage 3: Further Scenario Assessment The third stage involved further targeted review of FEP values to reduce uncertainties in support of revised qualitative and quantitative assessment of scenarios, driven by the findings of Stage 2.

Stage 4: Documentation

Stage 4 involved presenting the results of the analysis in reports on: • the likelihood of criticality as a result of LLW, ILW and DNLEU

disposal; and • the likelihood of criticality as a result of spent fuel, HLW, uranium

and plutonium disposal. These reports form the basis of this concluding synthesis report.



Figure 1.2: Stages, gates and activities in the Likelihood of Criticality project.

Stag

e 3

Waste packaging and disposal

concept understanding

FEPs and criticality scenario

construction

Stag

e 1

Methodology review

and development

Stag

e 2

FEP database

Evaluate critical masses and

concentrations

Initial data

elicitation

Stakeholder

Communication

GoldSim model

development

Initial scenario

assessment

Stag

e 4 Report preparation

Gate 3

Hold

Gate 2

Hold

Gate 1

Hold

GoldSim model revision

Further scenario assessment

Stakeholder engagement

Targeted data elicitation and GDF evolution

issues



1.3 Approach to Understanding the Likelihood of Criticality

Work to understand the likelihood of criticality has been underpinned by consideration of the evolution of conditions in a GDF and the subsequent construction of criticality scenarios. Criticality scenarios have been defined in terms of those that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package. Understanding the radioactive waste inventories, the GDF concepts for the different waste types and geological settings, and the expected evolution of conditions in the different GDF concepts, as well as associated uncertainties, has been of fundamental importance to the identification and analysis of these scenarios.

The criticality scenarios have been analysed in different ways depending on the fissile material contents of waste packages and the expected evolution of conditions in a GDF:

• If a waste package contains insufficient fissile material for criticality, even under the most favourable conditions that can be envisaged for criticality in the vicinity of the waste package, simple judgments have been made that criticality at the package scale is not credible.

• Where such judgments cannot be made or where scenarios involving accumulation of fissile material from more than one waste package are considered, a more detailed analysis of waste package degradation and fissile material relocation has been undertaken in order to evaluate the likelihood of criticality. In this case a probabilistic modelling approach has been taken in which parameter value uncertainties have been accounted for in probability density functions or arbitrary assumptions have been made about the occurrence or otherwise of particular processes. These probability distributions have been sampled over many model runs in order to estimate the probability (or likelihood) of critical concentrations or masses of fissile material developing after GDF closure.

Judgments about the conditions required for criticality in different components of a GDF (i.e., in waste packages, engineered barriers and host rock) are key to this analysis. However, there are large uncertainties in the materials that might be involved in fissile material accumulation scenarios and the configurations of the accumulated material. Such uncertainties have been addressed by making bounding assumptions. For example, in many cases fissile material has been assumed to accumulate in optimal spherical configurations and potential neutron absorbing materials have been ignored. Data on minimum critical masses and concentrations of fissile material in such configurations have been used to judge whether critical systems could develop in different components of a GDF.

In some cases, the modelling has shown that it is not possible to accumulate a critical mass or concentration of fissile material, conditional on the treatment of parameter value uncertainty and bounding assumptions about the requirements for criticality. In other cases, the modelling has shown that it is possible to accumulate a critical mass of fissile material with a certain probability. In these cases it is possible to make qualitative judgments about the low likelihood of criticality, again conditional on the treatment of parameter value uncertainty. However, at no point have the different estimates of the probability of criticality by different processes at different locations in a GDF, and involving different types of waste,



been combined to evaluate the overall likelihood of criticality in a GDF using probabilistic models.

The research has focused on illustrative concepts for radioactive waste disposal in different types of host rock, but the approach could be refined when more detailed information about specific disposal sites becomes available. The methodologies and tools developed in the project could also be used to support future assessments of waste packaging proposals.

1.4 Report Structure Section 2 presents a description of the assessment methodology. Section 3 discusses LLW/ILW/DNLEU and SF/HLW/HEU/Pu packaging concepts and inventories. The analyses of post-closure criticality scenarios for disposal concepts in higher-strength rock, lower-strength sedimentary rock and evaporite are presented in Sections 4, 5 and 6, respectively. A summary and conclusions are presented in Section 7.



2 Assessment Methodology The evaluation of the likelihood of criticality in a GDF for higher-activity radioactive wastes and other radioactive materials potentially requiring disposal has been underpinned by consideration of the features, events and processes (FEPs) that could affect reactivity2 after GDF closure. The FEP analysis led to the construction of post-closure criticality scenarios and these scenarios have been evaluated using a range of approaches, from high-level judgments about scenario credibility, to modelling of evolving conditions in a GDF to determine if critical systems could develop. Understanding the radioactive waste inventories, the GDF concepts for the different waste types and geological settings, and the expected evolution of conditions in the different GDF concepts, as well as associated uncertainties, has been of fundamental importance to the criticality FEP and scenario analysis. This section provides a high-level summary of the approach taken at each step of the criticality scenario and assessment process, in terms of identification of the wastes and disposal concepts (Section 2.1), analysis of criticality FEPs and scenarios (Section 2.2) and the approach taken to assess post-closure criticality scenarios (Section 2.3).

2.1 Wastes and Disposal Concepts The wastes and packaging concepts considered in the Likelihood of Criticality project are as presented in the NDA 2010 generic Disposal System Safety Case (DSSC) [14] and in the Derived Inventory [15]. The Derived Inventory is based on the Baseline Inventory of higher-activity waste and other materials that might need to be managed through geological disposal [3] and data from the 2007 UK RWI [4]. The scope of the Likelihood of Criticality project is limited to the Derived Inventory Reference Case, which excludes any potential new nuclear-build wastes and Ministry of Defence (MoD) radioactive materials3.

The period over which the waste packages provide criticality control depends on how conditions in the GDF environment evolve, which will depend on the GDF design and geological setting. As yet, no site has been identified for a disposal facility and therefore, in order to manage uncertainty concerning the geological environment for a disposal facility, RWMD is considering three generic host-rock types encompassing typical, potentially suitable UK geologies as follows [16, §4]:

• Higher-strength rocks - these are typically crystalline igneous, metamorphic rocks or geologically older sedimentary rocks, where any fluid movement is predominantly through discontinuities in the rock.

• Lower-strength sedimentary rocks - these are typically geologically younger sedimentary rocks, where any fluid movement is predominantly through the rock mass.

2 Reactivity is a measure of the departure of a system from a just critical state (when reactivity is

zero). A system is sub-critical when reactivity is negative. Changes in an evolving GDF associated with the behaviour of fissile material are often described in terms of whether they would increase or decrease reactivity.

3 The Derived Inventory considers two inventory scenarios, in addition to the Reference Case, that address uncertainties in forecasts of future waste arisings: the Lower and Upper Inventories give indicative estimates of the minimum and maximum volumes of waste that could require geological disposal.



• Evaporites - these are typically anhydrite (anhydrous calcium sulphate), halite (rock salt) or other rocks that result from the evaporation of water from water bodies containing dissolved salts.

RWMD has set out illustrative concepts for each of the three generic geological host rock environments [16, §5] based on the international GDF concept examples listed in Table 2.1. The scope of the Likelihood of Criticality project is limited to consideration of these illustrative geological disposal concepts.

The illustrative GDF designs are based on the assumption of a single facility to accommodate all of the wastes and materials in the Baseline Inventory [3]. In such a ‘co-located’ disposal facility it is assumed that there would be two distinct disposal areas, one for ILW, LLW and DNLEU (LLW/ILW disposal area) and the other for HLW, SF, plutonium and HEU (HLW/SF disposal area), but that disposal operations would share surface facilities, access tunnels, construction support and security provision [17, §5.2.4]. The analysis of the likelihood of criticality in a GDF has considered the two disposal areas separately [1, 2].

Table 2.1: Illustrative Disposal Concept Examples [16].



2.2 Criticality FEPs and Scenarios An analysis of FEPs is usually undertaken in order to provide a systematic framework for identifying all factors relevant to the post-closure safety of a GDF [18]. Scenarios are descriptions of possible evolutions of a GDF that are specified in terms of sets of FEPs and their interactions. The Likelihood of Criticality project has focused on ‘criticality FEPs’, which are here defined broadly as FEPs that could result in changes in reactivity after GDF closure. Criticality FEPs have been identified for each geological setting based on consideration of the evolution of waste packages under disposal conditions, and criticality scenarios have then been constructed based on consideration of sequences and combinations of these criticality FEPs.

It is recognised that features such as hydraulically conductive fracture zones (e.g., large faults) in the vicinity of a GDF could provide channels for focused groundwater flow and radionuclide transport, which could influence the likelihood of fissile material accumulation in the long term after GDF closure. However, the layout of a disposal facility would be influenced by the requirement to avoid intersecting zones of high hydraulic conductivity. Even so, the possibility that an accumulation zone exists downstream from a disposal region has been included in the analysis.

Also, it is recognised that a comprehensive FEP analysis includes consideration of events and processes external to the GDF that could alter conditions in the disposal facility after closure and affect the potential for criticality. For example, in the long term, natural processes such as tectonics, uplift or subsidence and erosion, and climate change may affect the geosphere. However, as discussed in Section 2.3, it is assumed that the geosphere at the depths of a GDF is not affected significantly by natural events and processes on assessment timescales.

GDF performance (and the likelihood of criticality) could also be affected by future human intrusion into the GDF. For example, drilling or mining intrusions could result in relocation of fissile material and alteration of geochemical and hydrological conditions, and could provide pathways for long-term fissile material transfer and accumulation. However, the type, frequency and effects of such events are largely site specific, such that the likelihood of their occurrence and the likelihood for criticality to occur as a consequence are difficult to predict. Rather, consistent with the GRA requirement on the assessment of human intrusion [7, Requirement R7], the assessment of human intrusion should focus on the potential consequences of intrusion, which could include the consequences of post-closure criticality. Therefore, the likelihood of criticality as a result of human intrusion has not been assessed in this project.

Criticality scenarios have been defined broadly in terms of:

• FEPs that could result in increased reactivity inside a single waste package.

• FEPs that could result in accumulation of fissile material outside a single waste package.

• FEPs that could result in accumulation of fissile material from multiple waste packages.

Event tree diagrams depicting the different events and processes involved in the development of fissile material accumulation scenarios have been produced for radioactive waste disposal in each geological setting.



2.3 Criticality Scenario Analysis The approaches to analysing post-closure criticality scenarios involve different levels of detail, depending largely on the fissile material contents of waste packages and the assumed properties of the host rock, as follows:

• For some waste types, the waste packages contain insufficient fissile material for criticality, even under the most favourable conditions that can be envisaged for criticality in the vicinity of the waste package. In this case, criticality at the package scale can be judged not to be credible.

• Where such judgments cannot be made, more detailed analyses of waste package degradation and fissile material relocation have been undertaken. For such scenarios, the defining sequence and combination of FEPs require evaluation in order to determine the concentrations and masses of fissile material that could occur at potential accumulation sites. A probabilistic modelling approach has been taken in this project, which allows parameters relevant to each criticality scenario to be sampled a sufficient number of times to achieve required confidence levels in the results. The likelihood of criticality has been evaluated based on comparisons of the calculated fissile masses and concentrations with minimum values required for criticality at different locations (i.e., in different GDF barrier components).

In order to support assessments of the likelihood of criticality, information is required on the minimum critical masses or concentrations that would be required for criticality under the conditions of a given accumulation scenario. Data on minimum critical masses and concentrations in different components of a GDF (i.e., waste packages, engineered barriers and host rock) have been obtained from criticality handbooks, criticality safety assessments, criticality calculations undertaken as part of RWMD’s project on the consequences of a hypothetical criticality in a GDF and other publicly available information. In addition, in some cases, the MCNP code [19] has been used to evaluate the neutron multiplication factor for different stages in the waste package degradation process.

Note that the project has focused on understanding the likelihood of criticality for wastes that contain substantial amounts of 239Pu and 235U, which are the key fissile nuclides (fissionable nuclides that can undergo fission with low energy neutrons). Nuclides that only fission as a result of interaction with fast neutrons are not considered to present a criticality concern in a GDF, because disposal systems are expected to be moderating (and potentially over-moderating) in the presence of waste materials and water.

The quantitative assessment of post-closure criticality scenarios has been undertaken using the GoldSim Monte Carlo simulation software [20, 21]. GoldSim provides a framework for simulating system behaviour while quantitatively representing system uncertainty. The likelihood of criticality modelling approach has involved implementing in GoldSim the equations representing fissile material migration and accumulation mechanisms and associated parameter value uncertainties for each post-closure criticality scenario requiring assessment. The probabilistic modelling capability in GoldSim allows parameter value distributions to be sampled over many code runs, or realisations, to take account of uncertainties. As far as possible, parameter values have been selected that are consistent with the Baseline Inventory and the illustrative geological disposal concepts associated with the three generic geological environments considered in the 2010 Generic DSSC. However, in some instances, discussion and agreement on the selection of parameter values has been required. Workshops involving RWMD experts provided the focus for making such judgments on parameter value selection. The parameters have been complied in a



Parameter Value Data Sheet, which serves as a single traceable document of values for use in post-closure criticality scenario assessments.

GDF post-closure criticality scenarios depend on the evolution of conditions in the near field of the GDF and the migration of fissile radionuclides over long time periods. The choice of timeframes over which to undertake criticality scenario assessment calculations is not straightforward. For instance, mobilisation and dispersion of radionuclides could occur on timescales from tens of years up to billions of years, depending on the type of waste package and GDF conditions. Also, uncertainties in GDF evolution tend to increase with time and the reliability of calculations over very long timeframes is questionable.

For example, the long-term performance of a GDF will depend inter alia on the stability of the geosphere. Natural processes such as tectonics, uplift or subsidence and erosion, and climate change, particularly future glaciations, may affect the geosphere in a UK geological setting over the next million years [22, §2.3; 23]. In the absence of a disposal site, the effects of such processes on geosphere and barrier system performance are difficult to predict. However, at depths under consideration for a GDF, the geosphere is less dynamic than shallow geological or surface environments and natural processes will occur slowly. The following broad observations can be made about the likely stability of the geological setting of a GDF [23]:

• The UK is situated in a tectonically quiet region, approximately a thousand kilometres from a tectonic plate boundary, and is expected to remain in a quiet region for many millions of years [23, §4]. The next period of volcanic activity in the UK is not expected to occur for at least some millions of years.

• To minimise potential effects, waste disposal vaults or deposition holes would be constructed such that waste would not be emplaced in the vicinity of larger fractures, which are more likely to be affected by future seismic activity. These larger fractures would be identified as part of the site characterisation process.

• Uplift accompanied by erosion at the Earth’s surface would reduce the burial depth of a GDF. However, the extent of any uplift and erosion before the next glaciation would be substantially less than the depth of a GDF [23, §4].

• Estimates of the extent of erosion caused by glaciation, which is not expected in the UK for 100,000 years or more, are significantly less than the design depth of a GDF [23, §4.3]. However, glacial loading and unloading could affect fracture properties and groundwater flow in the geosphere. Also, during glaciation, recharging groundwater will be dilute, low pH and oxygenating, which could affect engineered barrier performance and radionuclide transport behaviour if the groundwater reaches GDF depths.

The extent and timing of such changes and their potential impacts on GDF performance are uncertain, especially in the absence of a disposal site. However, for a carefully chosen site, the contribution to safety from the geosphere in the long term would be achieved [23, §8].

A timeframe of the order one million years has been chosen for the assessment of criticality scenarios for a GDF for LLW/ILW and DNLEU and as a focus for the assessment of criticality scenarios for a GDF for SF/HLW/HEU/Pu. It is assumed that the geosphere is not affected significantly by natural processes over such a timeframe. This timeframe allows for the decay of 239Pu to negligible amounts (about 241,000 years, representing ten half-lives of 239Pu).



The timeframe is also intended to ensure that the release of fissile material from corroded containers and its migration through the near field are represented. However, the corrosion rate distribution for the copper containers assumed in the higher-strength rock disposal concept is such that the containers would retain their integrity for in excess of 107 years. The possibility of failure of the copper containers by a disruptive event on a timescale of 105 years has been included, but, in order to provide an indication of the potential behaviour of wastes in a copper canister after failure by corrosion, the calculational timeframe has been extended to 108 years. There can be little confidence in the results of calculations this far into the future and they should be regarded as indicative of performance beyond one million years. However, because of the long half-lives of 235U (7x108 years) and 238U (4x109 years), the reactivity of a water-saturated waste package would change little from 105 years to 108 years and, therefore, the timing of container failure beyond 105 years is not important from a likelihood of criticality perspective.

Finally, it is recognised that future safety assessments may not include detailed quantitative analysis beyond a few hundred thousand years because of increases in uncertainty associated with the potential effects of climate change. Indeed, future safety assessments may rely on qualitative and simple quantitative arguments about the likelihood of criticality in a GDF in the very long term in combination with probabilistic modelling on timescales of a few hundred thousand years.

2.4 Overseas Approaches to Evaluating the Likelihood of Criticality in GDFs

To support development of the likelihood of criticality assessment methodology described above, a review of approaches to assessing the likelihood of criticality in overseas radioactive waste disposal programmes has been undertaken. The review focused on the following analyses because they aim to assess the probability of GDF criticality safety based on consideration of a comprehensive range of post-closure criticality scenarios:

• The recently published Criticality Safety Standard for nuclear fuel disposal in Germany [24].

• Criticality safety assessments for radioactive waste disposal in the US, covering transuranic waste, SF, HLW, HEU and Pu [25, 26, 27, 28].

• A criticality safety assessment for spent fuel disposal in Sweden published in the 1970s [29].

Details of this review are presented in the main report on SF/HLW/HEU/Pu disposal [2, Appendix B] and in the discussion of LLW/ILW/DNLEU disposal in evaporites in the main report on LLW/ILW/DNLEU disposal [1, §6.4]. Key findings are discussed in the following sub-sections.

2.4.1 Criticality Safety Standard for SF Disposal in Germany

The criticality safety standard for the final disposal of nuclear fuel in Germany sets out the approach to be taken in addressing criticality safety for the operational and post-closure phases of a geological disposal facility [24]. The standard requires an approach to evaluating the likelihood of criticality that is similar to that taken in the Likelihood of Criticality project. That is, criticality safety assessments must identify scenarios involving changes in the concentration, composition, geometry, moderation and/or neutron reflection of fuel units



that result in increases in reactivity, and estimate the likelihood of occurrence of a criticality event based on assessment of these scenarios.

However, the approach proposed in the German Criticality Safety Standard is novel in that it allows for a time-dependent tolerable probability of criticality in the post-closure phase, with a continuous transition from 10-6 in the operational phase to 10-4 at a time in the post-closure phase when a reasonable estimation of the probability of criticality is not possible. The increase in the tolerable probability of criticality is intended to compensate for increasing uncertainty. In this approach, the possible occurrence of a criticality event for times when estimation of the probability of criticality is not possible can be deemed acceptable if the consequences of the event can be shown not to disrupt waste containment and increase radiological or other risks.

The approach set out in the criticality safety standard has not yet been implemented in Germany, and an approach to constructing post-closure criticality scenarios and calculating their probability of occurrence remains to be documented. Evaluation of the time at which estimation of the probability of criticality is not possible is a key step in the German approach, but it is not clear how this time should be estimated. That is, it is not clear how the point at which uncertainties are too large to merit evaluation of scenario probability will be ascertained.

2.4.2 Criticality Safety Assessment for Transuranic Waste Disposal in the US

An assessment of the potential for criticality when disposing of transuranic waste at the Waste Isolation Pilot Plant (WIPP) disposal facility in New Mexico [25] found that, even if brine is assumed to be present in an evaporite, post-closure criticality would be unlikely. This assessment argued that the probability of criticality within the repository is low because mechanisms to concentrate the fissile radioisotopes dispersed throughout the waste are absent. In addition, following an inadvertent human intrusion into the repository (an event that must be considered according to US Environmental Protection Agency safety regulations), the probability of nuclear criticality away from the repository is low because:

• The amount of fissile mass transported over 10,000 years (the regulatory time frame assumed in the analysis for considering criticality) is predicted to be small.

• There is insufficient pore space (e.g., macroscopic fractures) for fissile material to precipitate in a large enough mass for criticality to occur.

• There is a natural tendency for the fissile material to disperse; there is no credible mechanism to concentrate the fissile material in a small enough volume for it to form a critical concentration.

To determine whether criticality could occur:

• Neutron transport modelling was used to develop physical constraints on fissile mass and geochemical constraints on fissile concentration.

• Process modelling was used to evaluate the feasibility of exceeding these bounds by examining whether physical, hydrogeological or geochemical constraints (e.g., absorption, colloid filtering and precipitation) exist.

Models were used to simulate the evolution of the disposal system and to determine whether conditions exist in which a criticality could occur at different locations. Without human



intrusion, it was judged that the hydrogeological constraints (lack of brine flow) mean that insufficient fissile material can be mobilised and accumulated for criticality. The argument for low probability of criticality in the far field following borehole intrusion was supported by the view that that 239Pu adsorption on rock would be accompanied by adsorption of other materials from the brine (potentially neutron absorbers) and by examination of the results of performance assessments for the WIPP that showed that only small amounts of 239Pu are transported from the disposal facility over 10,000 years.

2.4.3 Criticality Safety Assessments for SF/HLW/HEU/Pu Disposal in the US

Some of the most detailed assessments of the probability of post-closure criticality have been undertaken in the context of radioactive waste disposal in the US [25, 27, 28]. Again, these assessments reflect the approach followed in the Likelihood of Criticality project in that they involve:

• The development of a comprehensive set of scenarios describing the accumulation of fissile material that could result in the formation of critical configurations.

• The development of deterministic models of these criticality scenarios in order to evaluate a performance measure (i.e., a fissile material concentration or mass) for judging criticality safety.

• A statistical analysis of uncertainties associated with the performance measure obtained by deterministic models. Such statistical analysis generally requires many runs of a computer code developed for the deterministic transport model. Parameter values are then sampled and the mass of fissile material at accumulation locations is calculated using the deterministic model. By repeating this process, a probability density function for fissile material at the accumulation point may be produced.

• Judgment of disposal facility criticality safety by comparing the calculated performance measure (including uncertainty) with a safety standard. Calculated minimum critical masses for various configurations of fissile material are used as safety standards or to screen criticality scenarios.

The criticality safety assessment for the Yucca Mountain Repository (YMR) Licence Application [28] generally did not go so far as to evaluate the probability of critical configurations occurring for each derived criticality scenario. Instead, the probabilities of occurrence of barrier system defects and initiating events underpinning each critical scenario were evaluated based on mean values of probability distributions. These probabilities were then combined to evaluate the criticality potential of the scenario, but the probability that critical configurations would be generated as a result of these events was not evaluated. Generally, this approach was found to be sufficient to demonstrate that the probability screening criterion for the YMR is satisfied (i.e., less than one chance in 10,000 of occurrence over 10,000 years) [28]. However, the approach was supported by qualitative judgments regarding the probability of occurrence of the processes required for the formation of critical configurations, and neutron transport calculations that showed that representative configurations are sub-critical.

2.4.4 Criticality Safety Assessment for SF Disposal in Sweden

The critically assessment undertaken for the Swedish spent fuel disposal concept in the 1970s involved a detailed consideration of criticality scenarios [29]. These scenarios comprised events and processes that could lead to criticality in a disposal canister and



criticality in components of the engineered barrier system and host rock. The assessment involved deterministic analysis of the processes that could lead to criticality and neutron transport calculations to evaluate critical masses and concentrations in different configurations and media. However, the assessment did not go so far as to evaluate the probability that the criticality scenarios would occur. Instead, judgments were made that the critical systems were extremely unlikely to develop.

2.4.5 Summary

The German and US programme approaches to evaluating the likelihood of criticality based on assessment of post-closure criticality scenarios are similar to that taken in the Likelihood of Criticality project. In particular, the probabilistic approach to treating uncertainty and the key processes modelled (dissolution, advection, diffusion and sorption) in the US programme are reflected by the Likelihood of Criticality project, although the single fracture model configuration with a downstream accumulation zone assumed in the US analysis is simpler. Also, the possibility of using low consequence arguments when low probability arguments cannot be made in the German approach is similar to the approach being developed by RWMD through the Likelihood of Criticality project and the work on modelling the consequences of hypothetical criticality in a GDF [13].



3 UK Radioactive Waste Packaging Concepts and Inventories

3.1 Introduction Radioactive wastes are produced in the UK as a result of electricity generation in nuclear power stations and from the associated production and processing of the nuclear fuel, from the use of radioactive materials in industry, medicine and research, and from military nuclear programmes. These wastes are classified in terms of the nature and quantity of radioactivity they contain and their heat-generating capacity [4, §2.2]:

• Low-level waste (LLW) has a radioactive content not exceeding 4 GBq per tonne of α, or 12 GBq per tonne of β/ɣ activity. Most LLW requiring the environmental protection provided by near-surface vault disposal is assigned to the LLW Repository (LLWR) in West Cumbria. A small fraction of LLW will need to be disposed of in a GDF due principally to the concentration of specific radionuclides in the waste.

• Intermediate-level waste (ILW) exceeds the upper boundaries on radioactive content for LLW, but is limited in heat output. ILW arises mainly from reprocessing of spent nuclear fuel, general operations and maintenance as well as decommissioning and clean-up at nuclear sites.

• The temperature of high-level waste (HLW) may rise significantly as a result of its radioactivity. HLW is a by-product of spent fuel reprocessing and arises initially as a highly radioactive liquid. HLW is converted into a solid form using a vitrification treatment process [3, §3.3].

UK uranium and plutonium stocks are not currently classified as waste but, if it were decided at some point that these materials had no further use, they may need to be managed through geological disposal, as discussed in Section 1.1. Stocks of uranium derive mainly from refining uranium ore to make fuel, although separated uranium also derives from reprocessing spent fuel, as does separated plutonium. Natural, depleted, low-enriched uranium, highly-enriched uranium and separated plutonium are defined as follows [30, §2.5]:

• Natural uranium (NU) has a 235U content of 0.711 wt%. NU can be used in its metallic form in Magnox reactor fuel (although more recently some Magnox fuel is slightly enriched to offset ageing reactor effects).

• Depleted uranium (DU) is uranium with a 235U content less than in natural uranium. Some DU is a by-product of the uranium enrichment process used in the manufacture of nuclear fuels for Advanced Gas-cooled Reactor (AGR) and Pressurised Water Reactor (PWR) power stations. Such DU is currently stored as uranium hexafluoride (UF6). DU is also a product of reprocessing spent Magnox reactor fuel and this DU is stored as uranium trioxide (UO3).

• Low-enriched uranium (LEU) is uranium enriched to up to 20 wt% 235U. LEU is used as uranium dioxide (UO2) in the manufacture of AGR and PWR fuels, with a typical 235U content of between 3 and 5 wt%. LEU (with a reduced 235U content) is also a product of reprocessing these fuels, and the resulting material is stored as UO3.

• Highly enriched uranium (HEU) is uranium enriched in 235U to 20 wt% or more and is used in the manufacture of specialist nuclear fuels (e.g., for research reactors) and



for military purposes. In the past HEU has also been recovered by the reprocessing of these specialist fuels [30, §2.5].

• Plutonium (Pu) is created in nuclear reactors as a result of irradiating uranium in reactor fuel and is extracted from the spent fuel by reprocessing. Separated Pu is stored as plutonium dioxide powder (PuO2) [30, §2.4].

Spent fuel (SF) is not currently classified as a waste because it contains large amounts of U and some Pu that can be recovered through reprocessing and used to make new fuel. Most of the UK’s SF from civil reactors has been reprocessed, producing separated Pu and U, and HLW, ILW and LLW as waste by-products. SF need not be reprocessed, however, and could instead be packaged and disposed of directly in a GDF. Some SF from existing UK AGR power stations and all the SF from the Sizewell B PWR is not currently destined for reprocessing and may ultimately require direct disposal [3, §3.5].

In the next sections, the waste packaging concepts assumed for these materials, the uranium and plutonium contents of the wastes, and the waste streams of most significance to the evaluation of the likelihood of criticality in a GDF are described. Further information is provided in the more detailed project reports on the likelihood of criticality following the disposal of LLW/ILW/DNLEU [1, §3] and SF/HLW/HEU/Pu [2, §3].

3.2 LLW/ILW Packaging Concepts and Inventories

3.2.1 LLW/ILW Packaging

RWMD has defined five standardised waste containers that are considered suitable for packaging the majority of ILW and LLW in the form of unshielded and shielded waste packages [31]:

• Unshielded waste packages (500 litre drum, 3 m3 box and 3 m3 drum) are typically manufactured from stainless steel. Remote handling is usually required because of radiation levels.

• Shielded waste packages (2 m and 4 m boxes) are manufactured from stainless steel and have built-in shielding and/or contain low activity materials such that they do not require remote handling.

There is also a limited number of non-standard containers, specifically designed by waste producers for certain waste streams. However, according to assumptions made in the Derived Inventory analysis [15], almost all of the total ILW fissile mass would be packaged in 500 litre drums, 3 m3 boxes and 3 m3 drums and the packaging of ILW in these containers has formed the focus of the Likelihood of Criticality project. LLW can be packaged in 500 litre drums, and 2 and 4 m boxes, although the generic disposal facility designs [17, §2.5] assume LLW would be packaged in 4 m boxes.

Encapsulation with cementitious grout has generally been used to condition waste streams (see, for example, Figure 3.1) and the Likelihood of Criticality project has assumed such conditioning for LLW/ILW. However, alternative immobilisation and encapsulation matrices, such as glass, ceramics and polymers, may be more suitable for ensuring long term containment of certain types of waste under disposal conditions. Such matrices could include neutron absorbers in order to provide criticality control for wastes that include relatively large quantities or concentrations of fissile material, although the effectiveness of



such materials depends on how rapidly they are spent and their behaviour as the wasteform degrades.

Figure 3.1: Different means of encapsulation of ILW/LLW in cement-based encapsulants in

500 litre stainless steel drums [32, Figure 9; 33].

3.2.2 LLW/ILW Inventory

The Derived Inventory Reference Case includes packaged waste volumes of 361,692 m3 of ILW and 16,632 m3 of LLW not suitable for shallow disposal at the LLWR [34, Table 5.4]. Table 3.1 presents the disposal volumes and package numbers for the Reference Case Derived Inventory, for both SILW and UILW packages and LLW.



A summary inventory of selected fissile nuclides present in the ILW and LLW Reference Case Derived Inventory for disposal in a GDF, decayed to 2150 (an indicative closure time for the GDF), is presented in Table 3.2. Almost all of this fissile material will be emplaced in UILW packages and will be dispersed throughout thousands of waste disposal packages.

Table 3.1: LLW/ILW Derived Inventory Reference Case disposal volumes and package numbers. Waste volume values are from [34, Table 7.1 and Table 7.21].

LLW/ILW Reference Case

Conditioned Volume

(m3)

Packaged Volume

(m3)

Package Numbers

(-)

Unshielded ILW (UILW) 190,173 239,827 202,030

Shielded ILW (SILW) 84,723 121,865 6,344

Total ILW 274,896 361,692 208,374

LLW 15,656 16,632 1,106

Total ILW & LLW 290,552 378,324 209,480

Table 3.2: Key fissile nuclides in the LLW/ILW Derived Inventory Reference Case. Radionuclide activities at 2150 are from [34, Table 7.5], specific activities are from [35, Table II.1] and half-life data are from [36, Table A.1]. Calculated minimum critical masses for fissile nuclides when in the form of optimally water-moderated, water-reflected spheres are from [37, Table 5].

Nuclide Half-life (yr)

Specific Activity (Bq/g)

Minimum Critical Mass (kg)

Activity at 2150 (TBq)

Mass at 2150 (kg)

Ratio of Minimum Critical Mass to Mass in

Inventory

% of Total Fissile Mass

233U 1.59E+05 3.59E+08 5.70E-01 1.81E+00 5.04E+00 9 3.36E-02235U 7.04E+08 8.01E+04 8.20E-01 7.74E-01 9.66E+03 11,778 64.39239Pu 2.41E+04 2.31E+09 5.31E-01 1.23E+04 5.34E+03 10,049 35.58241Pu 1.44E+01 3.82E+12 2.44E-01 6.10E+02 1.60E-01 0.7 1.06E-03242mAm 1.41E+02 3.60E+11 1.90E-02 1.04E+02 2.89E-01 15 1.92E-03

The nuclides 239Pu and 235U are the most important fissile nuclides in LLW/ILW and, as indicated in Table 3.2, the inventory contains about 10,000 times the minimum mass required for a criticality under ideal conditions for each of these nuclides. However, the fissile material in ILW is generally mixed with a large amount of non-fissile material. For example, at the time of disposal, the 9.7 tonnes of 235U and 5.3 tonnes of 239Pu in ILW will be diluted with 1730 tonnes of 238U, such that the average effective enrichment of these materials in uranium will be less than 1 wt%. The 238U is also a neutron absorber in moderated systems and thus its presence increases the minimum mass of fissile material required for criticality, as does the presence of other neutron-absorbing materials in ILW.



AGR and PWR fuels typically have initial 235U content of between 3 and 5 wt% [38, §3.2.1] and it is considered that uranium enrichments in most ILW would not be greater than this. The average 235U content of uranium in the ILW inventory is about 0.6 wt%, although 235U enrichments approaching 100 wt% 235U are possible. The average 239Pu effective enrichment4 in the LLW/ILW inventory is 0.3 wt%, although some packages will contain much higher effective enrichments of 239Pu.

Other fissile isotopes are present in LLW/ILW, but only 233U and 242mAm will be present at 2150 (the assumed time of GDF closure) in quantities greater than the relevant minimum critical mass under ideal conditions for criticality. However, the 233U and 242mAm will generally be distributed amongst many waste packages such that each waste package contains substantially less than the mass required for criticality under optimum conditions. LLW/ILW will include potentially significant amounts of the fissile isotope 241Pu at the time of disposal, but 241Pu has a short half-life (14.4 years) and will have decayed to insignificant quantities by the time of GDF closure.

3.2.2.1 LLW/ILW in Shielded Packages According to the Derived Inventory for ILW and LLW [34], the conditioned LLW intended for geological disposal contains no uranium or plutonium and therefore does not present a criticality safety concern. Further, at 2150, the conditioned ILW to be placed in shielded packages is expected to include only 1 kg of 239Pu, 4 kg of 235U and a total of no more than about 100 g of all other fissile nuclides [34]. This fissile material will be distributed amongst disposal packages in amounts that do not present a post-closure criticality safety concern [1, §3.5.1] and, for this reason, the criticality safety of SILW/LLW is not considered further.

3.2.2.2 ILW in Unshielded Packages There is a broad variation in the expected fissile material and 238U content of UILW packages. This variation is evident in Figure 3.2, which shows the number of UILW packages as a function of 235U enrichment and 239Pu effective enrichment, and Figure 3.3, which shows the number of UILW packages as a function of 235U mass and 239Pu mass.

4 239Pu effective enrichment is defined here as

Mass(239Pu)/Mass(239Pu+232U+233U+234U+235U+236U+238U).



≤ 0.1%0.5% < x ≤ 1%

2% < x ≤ 3%4% < x ≤ 5%

10% < x ≤ 20%30% < x ≤ 40%

50% < x ≤ 60%70% < x ≤ 80%90% < x ≤ 100%

0

10,000

20,000

30,000

40,000

50,000

60,000

Num

ber o

f Was

te P

acka

ges

235U Enrichment (%)

Figure 3.2: Number of UILW packages as a function of average 235U enrichment and average 239Pu effective enrichment at 2150 for the Derived Inventory Reference Case. The colours indicate different 239Pu effective enrichment ranges.

= 0 g1 < x ≤ 5 g

10 < x ≤ 100 g200 < x ≤ 300 g

400 < x ≤ 500 g600 < x ≤ 700 g

800 < x ≤ 900 gx > 1000g

0

5,000

10,000

15,000

20,000

25,000

30,000

35,000

40,000

Num

ber o

f Was

te P

acka

ges

235U Mass (g)

Figure 3.3: Number of UILW packages as a function of average 235U mass and average 239Pu mass per package at 2150 for the Derived Inventory Reference Case. The colours indicate different 239Pu mass ranges.



Most waste streams to be packaged contain relatively small average masses of fissile material, which means that, for these waste streams, each waste package will contain on average less than the minimum mass of fissile material required for criticality under optimum conditions for criticality. These waste streams account for about 97% of the 202,000 waste packages. That is, 97% of waste packages will contain on average less than 0.53 kg of 239Pu (or the fissile gram equivalent of other radionuclides), which is the minimum critical mass of a water-moderated, water-reflected sphere of 239Pu (Table 3.2). Of course, for each waste stream, there will be a distribution of fissile masses in the waste packages, such that some waste packages may contain more fissile material than the above-noted minimum critical mass.

RWMD has undertaken a number of deterministic criticality safety assessments (CSAs) in order to identify safe fissile masses for waste packages containing different categories of fissile material. These CSAs include the General Criticality Safety Assessment (GCSA) [39] and four generic CSAs for waste packages containing high enriched uranium (ILW-HEU) [40], low-enriched uranium (ILW-LEU) [41], irradiated natural and slightly enriched uranium (ILW-INU) [42], and separated plutonium (ILW-Pu) [43]. These CSAs derived package fissile material limits based on consideration of conditions during a GDF’s operational phase and conditions after GDF closure, including scenarios involving fissile material accumulation from multiple waste packages. As such, these CSAs provide a means of identifying waste packages that present the greatest post-closure criticality safety concerns (i.e., waste packages that exceed fissile material limits derived from deterministic post-closure criticality scenario analysis).

The GCSA [39] derived package fissile material limits for waste packages that contain pure 239Pu and 235U and was based on pessimistic assumptions about post-closure accumulations of fissile material. The GCSA supported a general screening level (GSL) for ILW packages of 50 g 239Pu (or the fissile gram equivalent of other radionuclides in terms of contribution to reactivity). Waste packages that meet the GSL are considered by RWMD to be acceptable for disposal from a criticality safety perspective. On this basis, the analysis of the likelihood of criticality for package-scale scenarios (i.e., for single waste packages) is focused on consideration of waste packages that on average contain more than the 50 g 239Pu GSL. However, the analysis of criticality scenarios involving accumulation of fissile material from multiple waste packages is based on consideration of all UILW packages.

The package-scale analysis also focuses on waste packages that contain fissile material at the highest enrichments (or effective enrichments). Broadly, if the total enrichment is less than about 1 wt%, then criticality would only be possible for uranium metal (not uranium oxide) in the presence of an efficient moderator such as graphite, as in Magnox reactors.

Details of waste streams that contain on average more than the 50 g 239Pu GSL, and where the sum of the 239Pu effective enrichment and the 235U enrichment is greater than 1 wt%, are provided in Table 3.3. These waste streams generate about 17,500 waste packages (a little under 10% of all UILW disposal packages), of which about 65% (11,440 packages) are from waste stream 2D03 (Plutonium Contaminated Materials; Drums).

UILW packages that contain more than 50 g of 239Pu, but where the 239Pu effective enrichment is less than 1 wt%, present a concern if there is potential for the plutonium to be separated from the uranium under disposal conditions, such that the effective enrichment increases as dilution and neutron absorbance by 238U is reduced. Waste streams for which waste packages contain on average more than 50 g of 239Pu, but where the total enrichment is less than 1 wt%, are listed in Table 3.4 and amount to about 14,200 waste packages.



An indication of the potential significance to post-closure criticality safety of the waste packages that contain on average more than 50 g 239Pu (or equivalent) can be obtained by comparison with the screening levels derived in the CSAs for the four generic categories of ILW (ILW-HEU, ILW-LEU, ILW-INU and ILW-Pu). The generic CSAs included derivation of waste package screening levels for different types of container based on consideration of post-closure criticality scenarios. Upper screening levels (USLs) were derived based on particular assumptions about wasteform characteristics, such as wasteform uniformity.

The post-closure USLs derived for the different categories of ILW packaged in 500 litre drums and 3 m3 boxes are listed in Table 3.5. These USLs were calculated based on the assumption of the accumulation of fissile material from seven waste packages stacked in a disposal vault. The generic waste categories to which the waste streams most closely relate and whether the average fissile mass content of the waste packages exceeds the relevant post-closure scenario USL is indicated in Table 3.3 and Table 3.4. Cases in which the average total fissile mass per package exceeds the USL are highlighted red and these waste packages have been the focus of the package-scale criticality scenario assessment.



Table 3.3: UILW streams for which waste packages are expected to contain an average of more than 50 g of fissile material at 2150 and for which the total 239Pu effective enrichment and 235U enrichment is greater than 1 wt%. Key waste package contents are shown. The generic waste categories to which the waste streams could be attributed are indicated and cases in which the average total fissile mass per package exceeds the generic USL are highlighted red.

Waste Stream

ID

Waste Stream Description Waste Package Type

Number of Waste Packages

Mass per waste package 239Pu/ (239Pu+U)

(%)

235U/U (%)

Category 239Pu (g) 233U (g) 235U (g) 238U (g) Graphite

(kg) Total

Fissile Mass (g)

5C50 Dragon Fuel Enhanced 500 litre drum (pre-cast) 9 4.5E+02 2.1E+02 9.6E+03 1.0E+04 7.2E+02 1.0E+04 2.2 48.5 ILW-Pu/

ILW-HEU

5B32 Irradiated Thorium Fuel Pin Pieces

Enhanced 500 litre drum (basket) 7 3.2E+01 2.8E+02 1.5E+00 2.3E-03 0 3.2E+02 9.4 0.5 ILW-Pu/

ILW-HEU5

5C54 Zenith Fuel Enhanced 500 litre drum (pre-cast) 4 2.3E-05 0 1.3E+04 2.4E+05 9.3E+01 1.3E+04 0.0 4.9 ILW-LEU

5B350 Uranium Recovery Plant ILW

Enhanced 500 litre drum (pre-cast) 24 0 0 1.1E+04 4.5E+03 0 1.1E+04 0.0 70.7 ILW-HEU

5B19 Uranium Contaminated Materials

Enhanced 500 litre drum (pre-cast) 46 0 0 7.2E+03 3.5E+03 0 7.2E+03 0.0 63.5 ILW-HEU

5B328 Low Alpha RHILW Silo ILW 500 litre drum 1,240 2.5E-01 2.3E-04 4.4E+02 6.9E+01 0 4.4E+02 0.0 86.5 ILW-HEU

7A29 Uranium Contaminated Operations ILW 500 litre drum 115 0 0 3.3E+02 2.2E+01 2.8E+00 3.3E+02 0.0 93.1 ILW-HEU

7A117 Decommissioning Waste Uranium Contaminated ILW 500 litre drum 839 0 0 3.3E+02 2.2E+01 0 3.3E+02 0.0 93.1 ILW-HEU

2F31 Oxide Fuel Hulls from Early Reprocessing

3 m³ Sellafield Enhanced box 61 3.8E+01 6.4E-05 2.9E+02 1.9E+04 0 3.3E+02 0.2 1.5 ILW-LEU

5C304b Radiochemical Laboratory Decommissioning RHILW & CHILW

Enhanced 500 litre drum (basket) 150 6.3E+00 2.3E-01 5.8E+01 5.2E+03 0 6.5E+01 0.1 1.1 ILW-LEU

2N01 Plutonium Contaminated Material; Drummed (Original Inventory)6

Enhanced 500 litre drum (basket) 147 4.1E+01 1.2E-06 3.4E+02 1.5E-05 0 3.8E+02 10.8 100.0 ILW-Pu/

ILW-HEU

5 Defined as ILW-HEU because of the enrichment of the fuel in 233U. 6 The high uranium enrichment reported is likely to be due to inaccuracies in the UK RWI. For example, fresh Magnox fuel is at most only slightly

enriched and THORP processes UO2 fuel originally enriched to a maximum of 4 wt% 235U [41].



Waste Stream

ID




(%)

235U/U (%)


(kg) Total

Fissile Mass (g)

2N03 Plutonium Contaminated Material; Drummed (Operational Soft Waste)6


ILW-HEU

2N100

Plutonium Contaminated Material; Drummed (Decommissioning Hard Waste)6


ILW-HEU

2N02 Plutonium Contaminated Material; Crated6


ILW-HEU 5B31 Plutonium Nitrate Waste 500 litre drum 3 2.6E+03 3.1E-04 1.1E+01 1.0E+03 0 2.6E+03 70.6 1.0 ILW-Pu

2D59 Magnox Fuel End Crops 3 m³ Sellafield Enhanced box 10 5.6E+02 1.1E-04 2.4E+00 2.6E+01 0 5.7E+02 94.9 7.8 ILW-Pu

5B30 Development Laboratory Irradiated Dissolver Liquor 500 litre drum 1 4.0E+02 5.6E-04 1.8E+01 3.2E+03 0 4.1E+02 11.0 0.6 ILW-Pu

7A36 Pyrochemical Wastes 500 litre drum 46 2.1E+02 4.5E-06 8.6E-01 2.4E-05 0 2.1E+02 99.5 79.9 ILW-Pu

2D03 Plutonium Contaminated Materials; Drum6

Enhanced 500 litre drum (basket) 11,440 1.0E+02 2.1E-05 4.1E-01 1.1E-02 0 1.0E+02 99.1 44.1 ILW-Pu

5B03 Operational RHILW 500 litre drum 333 8.4E+01 3.6E-05 2.0E+00 2.7E+02 0 8.6E+01 23.4 0.7 ILW-Pu

7A21 Operational ILW Plutonium Contaminated 500 litre drum 1,914 4.9E+01 2.9E-07 5.5E+00 3.4E-01 0 5.4E+01 89.2 92.6 ILW-Pu

2F02 Plutonium Contaminated Materials; Drums

Enhanced 500 litre drum (basket) 397 5.2E+01 1.9E-05 2.0E-01 8.0E-04 0 5.2E+01 97.8 16.5 ILW-Pu

2D58 Uranium Residues in Pile Fuel Storage Pond

3 m³ Sellafield Enhanced box 19 2.5E+03 1.2E-02 5.9E+03 6.4E+05 8.4E-01 8.3E+03 0.4 0.9 ILW-INU

5B23 DFR Breeder 500 litre drum 220 1.9E+03 1.1E-02 1.4E+03 2.2E+05 0 3.3E+03 0.8 0.6 ILW-INU



Table 3.4: UILW streams for which waste packages are expected to contain an average of more than 50 g 239Pu at 2150 and for which the total effective enrichment is less than 1 wt%. The generic waste categories to which the waste streams could be attributed are indicated and cases in which the average total fissile mass per package exceeds the generic USL are highlighted red.

Waste Stream

ID




(%)

235U/U (%)


(kg) Total

Fissile mass (g)

2D17 Cemented Wastes in Skips 3 m³ Sellafield Enhanced box 29 3.9E+03 2.5E-03 1.1E+04 2.4E+06 0 1.5E+04 0.16 0.46 ILW-INU

2D45 Magnox Fuel End Crops 500 litre drum 66 6.7E+02 4.1E-04 1.7E+03 3.8E+05 0 2.3E+03 0.17 0.43 ILW-INU

2D24 Magnox Cladding and Miscellaneous Solid Waste

3 m³ Sellafield Enhanced box 1,123 3.8E+02 3.2E-04 6.9E+02 1.8E+05 1.6E-02 1.1E+03 0.21 0.38 ILW-INU


3 m³ Sellafield Enhanced box 853 2.9E+02 2.4E-04 5.4E+02 1.4E+05 0 8.3E+02 0.21 0.39 ILW-INU

2D16 Magnox Fuel Storage Pond Sludge



3 m³ Sellafield Enhanced box 2,349 1.9E+02 1.5E-04 3.8E+02 9.2E+04 2.1E+01 5.8E+02 0.21 0.42 ILW-INU

2D79 Magazines in Magnox Fuel Storage Pond



3 m³ Sellafield Enhanced box 2,832 9.2E+01 5.3E-05 2.4E+02 5.0E+04 2.5E+01 3.3E+02 0.18 0.48 ILW-INU



2D78 Decanner Settling Tank Sludge 500 litre drum 76 9.5E+01 4.0E-05 1.5E+02 3.8E+04 0 2.5E+02 0.25 0.40 ILW-INU

6C32 NDS Remote Handled ILW Enhanced 500 litre drum (pre-cast) 2 6.7E+01 3.0E-03 1.8E+02 2.6E+05 0 2.4E+02 0.03 0.07 ILW-INU

2D34 Sludge from Sand Filters and Transfers 500 litre drum 5,060 6.4E+01 4.4E-03 1.2E+02 3.1E+04 0 1.8E+02 0.20 0.37 ILW-INU

2D33 Fuel Handling Plant Sludges 500 litre drum 60 6.4E+01 5.6E-05 1.2E+02 3.1E+04 0 1.8E+02 0.20 0.37 ILW-INU 2D72 Sludge Settling Tank 500 litre drum 176 5.1E+01 2.1E-05 8.1E+01 2.0E+04 0 1.3E+02 0.25 0.40 ILW-INU



Table 3.5: Generic post-closure USLs for different categories of waste packaged in 500 litre drums and 3 m3 boxes. The enrichments indicate the values assumed in the calculations undertaken in the four generic CSAs. The USLs are equivalent for ILW-Pu and ILW-HEU waste packages.

Waste Category Waste Package USL (g)

ILW-INU (up to 1.9 wt% 235U) [42] 500 litre drum 235U = 547

3 m3 box 235U = 2,760

ILW-LEU (up to 4 wt% 235U) [41] 500 litre drum 235U = 375

3 m3 box 235U = 1,900

ILW-Pu (up to 100 wt% 239Pu) [43] or ILW-HEU (up to 100 wt% 235U) [40]

500 litre drum 235U + 1.11 x 239Pu = 240

3 m3 box 235U + 1.11 x 239Pu = 1,220

3.3 DNLEU Packaging Concepts and Inventories

3.3.1 DNLEU Packaging

The development of a packaging concept for uranium wastes is at an early stage. For the preparation of the Derived Inventory for uranium, DNLEU was assumed to be packaged as uranium oxide (UO2) and mixed with a grout encapsulant in 500 litre drums or in 3 m3 boxes, and treated as UILW [44, §3.1]. This has also been assumed in the likelihood of criticality analysis.

3.3.2 DNLEU Inventory

Table 3.6 shows the masses of DNLEU streams in the Derived Inventory Reference Case. The Derived Inventory assumes that all uranium for disposal would be in the form of U3O8. The packaged DNLEU waste volume is 94,400 m3 [44, Table 7.1] and it is anticipated this waste will be contained in 165,326 DNLEU 500 litre drums [44, Table C3].

A summary inventory of selected fissile nuclides present in the DNLEU Reference Case inventory, decayed to 2150, is presented in Table 3.7. The DNLEU includes 1.6x108 kg 238U, such that the average 235U enrichment over all DNLEU waste streams is only 0.3 wt%. However, the LEU waste stream MU006 includes 235U at much higher enrichments. LEU waste stream MU006 would be contained in 2,067 waste packages (500 litre drums), with each waste package containing on average 19.4 kg 235U and 949 kg 238U, such that the average 235U enrichment is 2 wt%. There is no plutonium in waste stream MU006 [44, Appendix E]. The MU006 LEU waste packages contain substantially more than the post-closure USL of 375 g 235U for ILW-LEU waste packages shown in Table 3.5, which indicates that the likelihood of post-closure criticality requires detailed consideration for these wastes.



Table 3.6: DNLEU streams and their masses in the Derived Inventory Reference Case [44, Table 5.2].

Stream Identifier

Stream Name Assumed Mass (tonnes Heavy Metal [tHM])

Stock at 1/04/2007 Future Arisings

MU003 DU from enrichment 31,000 87,000†

MU004 Magnox depleted uranium (from SF reprocessing)

33,000 6,000

MU005 AGR depleted uranium (from SF reprocessing) 2,000 1,300

MU006 LEU 2,000* 0

MU007 Natural uranium 2,000* 0* Balance of DNLEU mass allocated equally to LEU and natural uranium. † Balance of DNLEU allocated wholly to enrichment tails.

Table 3.7: Key fissile nuclides in the DNLEU Derived Inventory Reference Case. Radionuclide activities at 2150 are obtained from [44, Table 7.4], specific activities are from [35, Table II.1], and half-life data are from [36, Table A.1]. Calculated minimum critical masses in the form of optimally water-moderated, water-reflected spheres are from [37, Table 5].

Nuclide Half-life (yr) Specific Activity (Bq/g)

Minimum Critical Mass

(kg)

DNLEU Mass at 2150

(kg) Ratio of Minimum Critical Mass to

Mass in Inventory233U 1.59E+05 3.59E+08 5.70E-01 8.22E-03 0.01235U 7.04E+08 8.01E+04 8.20E-01 4.84E+05 590,430239Pu 2.41E+04 2.31E+09 5.31E-01 6.81E-03 0.01

3.4 HLW Packaging Concepts and Inventories

3.4.1 HLW Packaging

HLW is conditioned and packaged at Sellafield in the Waste Vitrification Plant (WVP). The vitrification process converts the highly active liquid by-product from spent fuel reprocessing facilities at Sellafield into a borosilicate glass in 150 litre stainless steel canisters (also known as WVP cans), as shown in Figure 3.4 [45, §4.2.1]). Packaging assumptions have been made based on RWMD’s three illustrative geological disposal concept examples, as discussed in the 2010 DSSC [17]. For the illustrative concept design for higher-strength rock it has been assumed that copper canisters with cast iron inserts would be used; steel disposal canisters have been assumed for the illustrative concept designs for lower-strength sedimentary rock and evaporite [45, §3.1]. Two WVP cans would be placed in each disposal canister, as shown in Figure 3.4 for the copper canister design.



Figure 3.4: Cutaway showing simulated vitrified HLW in a stainless steel container [3] (left)

and the copper canister for packaging HLW for disposal in higher-strength rock [45] (right). A carbon steel canister is assumed in the design for HLW disposal in lower-strength sedimentary rock and evaporite.

3.4.2 HLW Inventory

The HLW and SF Derived Inventory Reference Case includes four HLW streams with a total packaged volume of 7,457 m3 [45, Table 5.2]. It is anticipated that this HLW will require 3,656 disposal packages [45, Table C2].

A summary inventory of selected fissile nuclides present in the HLW Reference Case inventory for disposal in a GDF, decayed to 2150, is presented in Table 3.8. The fissile nuclides shown in Table 3.8 are present in total quantities that are greater than the relevant minimum critical mass under ideal conditions for criticality. However, these nuclides would be dispersed in small quantities in many waste packages, and the likelihood that they would contribute significantly to any accumulation of fissile material after disposal is small.



Table 3.8: Key fissile nuclides in the HLW Derived Inventory Reference Case. Radionuclide activities at 2150 are obtained from [45, Tables F2 to F6], specific activities are from [35, Table II.1], and half-life data are from [36, Table A.1]. Calculated minimum critical masses for fissile nuclides when in the form of optimally water-moderated, water-reflected spheres are from [37, Table 5].



Minimum Critical

Mass (kg)

HLW Mass at

2150 (kg)Ratio of Minimum Critical Mass to

Mass in Inventory 235U 7.04E+08 8.01E+04 8.20E-01 1.08E+01 13 239Pu 2.41E+04 2.31E+09 5.31E-01 8.72E+01 164 242mAm 1.41E+02 3.60E+11 1.90E-02 1.24E+00 65 245Cm 8.53E+03 6.37E+09 4.10E-02 2.45E+00 60

Table 3.9 presents isotope data on a per package basis for each HLW stream in the Reference Case Derived Inventory for an indicative GDF closure date of 2150. As would be expected, the HLW streams have 235U enrichments of less than 1 wt% and, whilst the 239Pu effective enrichment is as high as 9 wt% for two waste streams, the total fissile mass of the four key fissile nuclides (233U, 235U, 239Pu and 241Pu) is less than 100g per disposal package. HLW packages are thus unlikely to pose a criticality safety concern, although the potential for migration and accumulation of fissile material from multiple HLW packages requires consideration.


Not Protectively Marked NDA RWMD 17293-TR-023 Page | 32 The Likelihood of Criticality Synthesis Report

Table 3.9: HLW streams in the Derived Inventory Reference Case, with isotope data provided for an indicative GDF closure date of 2150. The colour scales for 239Pu, 241Pu, 233U, 235U, total fissile mass, 235U enrichment and 239Pu effective enrichment indicate relative low (green) to high (red) values. The waste is assumed to be packaged in a copper or steel disposal canister for vitrified HLW. There is no graphite in any of these waste streams. Waste package numbers have been rounded up.

Waste Stream

ID

Waste Stream Description


Mass per Waste Package Total Fissile

Mass (g)

Total U (g)

235U/ Total U

(%)

239Pu/ (239Pu+U)

(%) 239Pu (g) 240Pu (g) 241Pu (g) 242Pu (g) 233U (g) 235U (g) 238U (g) 242mAm (g)

Be (kg)

2D02/C Vitrified High Level Waste - Magnox

1,953 1.09E+01 3.68E+00 3.92E-04 1.50E-01 4.71E-03 2.68E+00 6.47E+02 3.27E-01 2.42E-05 1.36E+01 6.50E+02 0.4 1.6

2F01/C Vitrified High Level Waste

1,098 6.06E+01 5.62E+01 8.08E-03 3.98E+00 3.85E-02 5.08E+00 6.42E+02 4.67E-01 6.62E-06 6.57E+01 6.50E+02 0.8 8.5

2F22/C High Level Contaminated Waste

101 1.55E+00 2.02E+00 3.85E-04 1.34E-01 1.05E-03 1.24E-01 1.58E+01 1.55E-02 1.52E-07 1.68E+00 1.60E+01 0.8 8.9

2F38/C Vitrified High Level Waste from POCO

504 0 0 0 0 0 0 0 0 0 0 0 0.0 0.0



3.5 Spent Fuel Packaging Concepts and Inventories

3.5.1 Spent Fuel Packaging

AGR fuel assemblies would be dismantled prior to disposal and the fuel pins would be consolidated in bundles, with each bundle being contained within a slotted can. A total of eight consolidated fuel bundles (equivalent to the pins from 24 AGR fuel elements) would be packaged inside a disposal canister (as shown in Figure 3.5 for copper canisters). PWR fuel assemblies would not be dismantled before packaging for disposal. Four fuel assemblies would be packaged within a single disposal canister (see Figure 3.5) [45, §3.3].

The HLW and SF Derived Inventory variant scenarios [45] include other types of spent fuel that might require geological disposal (i.e., spent Magnox fuel, miscellaneous spent fuels at Sellafield, spent Prototype Fast Reactor fuel at Dounreay and submarine fuel), but for consistency with the Derived Inventory Reference Case, the Likelihood of Criticality project has focused on the disposal of AGR and PWR spent fuel.

Figure 3.5: Copper canisters for AGR (left) and PWR (right) spent fuel disposal in higher-

strength rock [45].



3.5.2 Spent Fuel Inventory

The HLW and SF Derived Inventory Reference Case comprises two SF waste streams with a packaged volume of 10,363 m3 [45, Table 5.2]. It is anticipated that this SF will require 5,996 disposal packages [45, Table C2]. Table 3.10 presents the disposal volumes and package numbers for the Reference Case Derived Inventory.

In order to calculate radionuclide activity in spent fuel, it is assumed in the Derived Inventory that the fuel has been subject to a high burn-up. That is, the emphasis in the Derived Inventory is on the radionuclide activity in spent fuel rather than on fuel of the highest reactivity; waste packages that include fresh fuel or spent fuel of low burn-up would exhibit higher reactivities. For spent AGR fuel an initial 235U enrichment of 3.5 wt% and a fuel burn-up of 39 GWd/tU was assumed [45, §4.3.3]. For spent PWR fuel an initial enrichment of 4.2 wt% and a burn-up of 60 GWd/tU have been assumed. A summary inventory of selected fissionable nuclides present in the SF Reference Case inventory for disposal in a GDF, decayed to 2150, is presented in Table 3.11.

Table 3.10: SF Derived Inventory Reference Case disposal volumes and package numbers

[45, §5.2, Table 5.2 and Table C2].

HLW and SF Reference Case

Conditioned Volume

(m3)

Packaged Volume

(m3)

Package Numbers

(-)

AGR SF 1,933.3 8,491.0 5,341

PWR SF 491.5 1,871.7 655

Total SF 2,424.8 10,362.7 5,996

Table 3.11: Key fissile nuclides in the SF Derived Inventory Reference Case. Radionuclide

activities at 2150 are obtained from [45, Tables F2 to F6], specific activities are from [35, Table II.1], and half-life data are from [36, Table A.1]. Calculated minimum critical masses for fissile nuclides when in the form of optimally water-moderated, water-reflected spheres are from [37, Table 5].




AGR SF PWR SF Mass at

2150 (kg)Ratio of

Minimum Critical Mass to Mass in

Inventory

Mass at 2150 (kg)


Inventory 235U 7.04E+08 8.01E+04 8.20E-01 2.42E+04 29,521 4.78E+03 5,828 239Pu 2.41E+04 2.31E+09 5.31E-01 1.44E+04 27,043 8.81E+03 16,586 241Pu 1.44E+01 3.82E+12 2.44E-01 5.42E+00 22 3.38E+00 14 242mAm 1.41E+02 3.60E+11 1.90E-02 7.69E+00 404 1.95E+00 103 245Cm 8.53E+03 6.37E+09 4.10E-02 5.03E+00 123 1.59E+01 387



The most important fissile nuclides in the SF inventory are 239Pu and 235U, which will be present in quantities equivalent to tens of thousands of minimum critical masses under optimum conditions for criticality (i.e., optimally water-moderated, water-reflected spheres of 239Pu or 235U).

Of the remaining fissile nuclides in Table 3.11, 241Pu, 242mAm, and 245Cm are present in total quantities that are greater than the relevant minimum critical mass under ideal conditions for criticality. It is expected that these nuclides would be dispersed in small quantities amongst wastes in many packages, and that the likelihood that they would contribute significantly to any accumulation of fissile material after disposal is negligible.

Table 3.12 presents isotope data on a per package basis for each SF waste stream in the Reference Case Derived Inventory for a GDF closure date of 2150. The SF waste streams contain about 7 kg of the four key fissile nuclides per AGR SF package and about 21 kg per PWR package. The likelihood of post-closure criticality requires detailed consideration for AGR and PWR SF packages, although the average 235U and 239Pu enrichments are less than 1 wt%.



Table 3.12: SF waste streams in the Derived Inventory Reference Case, with isotope data provided for an indicative GDF closure date of 2150. The colour scales for 239Pu, 241Pu, 233U, 235U, total fissile mass, 235U enrichment and 239Pu effective enrichment indicate relative low (green) to high (red) values. The waste is assumed to be packaged in a copper or steel disposal canister for AGR and PWR SF. There is no graphite in any of these waste streams. Waste package numbers have been rounded up.

Waste Stream

ID

Waste Stream Description


Mass per Waste Package Total Fissile

Mass (g)

Total U (g)

235U/U (%)

239Pu/ (239Pu+U)

(%) 239Pu (g) 240Pu (g) 241Pu (g) 242Pu (g) 233U (g) 235U (g) 238U (g) 242mAm (g)

Be (kg)

M2D100 AGR Spent Fuel 5,341 2.71E+03 1.88E+03 1.02E+00 8.84E+02 5.37E-02 4.54E+03 9.71E+05 1.34E+00 7.91E-06 7.25E+03 9.81E+05 0.5 0.3

M3S100 PWR Spent Fuel 655 1.36E+04 7.96E+03 5.17E+00 1.70E+03 1.08E-01 7.31E+03 1.67E+06 2.77E+00 1.58E-06 2.09E+04 1.69E+06 0.4 0.8



3.6 Pu and HEU Packaging Concepts and Inventories

3.6.1 Pu and HEU Packaging

For the U/Pu Derived Inventory, it was assumed that Pu is immobilised in a titanium-based ceramic [44, §3.1]. A nominal 10 wt% of Pu would be included in the ceramic, which would be produced in the form of pucks, with each puck containing about 50 g Pu [44, §3.2]. About 20 pucks would be loaded into each stainless steel can. The ceramic would also include the neutron poisons hafnium and gadolinium as well as depleted uranium. Up to 28 stainless steel cans would then be encapsulated in borosilicate glass within large steel canisters (the can-in-canister approach), which would be loaded into disposal canisters of the same design as assumed for HLW and SF. HEU is assumed to be immobilised in ceramic and packaged in the same way as Pu.

3.6.2 Pu and HEU Inventory

Table 3.13 gives the masses of Pu and HEU streams derived for the Reference Case. These materials represent packaged waste volumes of 6,989 m3 for Pu and 102 m3 for HEU [44, Table 7.1]. It is anticipated that this waste will require 3,478 Pu and HEU can-in-can canister packages [44, Table C3]. Table 3.14 presents the disposal volumes and package numbers for the Reference Case Derived Inventory.

Table 3.13: Plutonium and uranium streams and their masses in the U/Pu Derived Inventory Reference Case [44, Table 5.2].

Stream Identifier

Stream Name Assumed Mass (tHM)

Stock at 1/04/2007 Future Arisings

Plutonium

MPu001 Plutonium from Magnox fuel reprocessing 70 13

MPu002 Plutonium from AGR fuel reprocessing 8 7

MPu004 Other plutonium (includes materials from reprocessing, including PFR fuel)

3 0

Uranium

MU001 HEU from civil nuclear programmes 1.44 0

Table 3.14: HEU and Pu Derived Inventory Reference Case disposal volumes and package numbers [44, Table 7.1 and Table C3].

U/Pu Reference Case Conditioned Volume

(m3)

Packaged Volume

(m3)

Package Numbers

(-)

Plutonium 3,049.41 6,988.64 3,428

Uranium (HEU) 44.36 101.70 50

Total 3,093.77 7,090.34 3,478



A summary inventory of key fissile nuclides present in the U/Pu Reference Case inventory, decayed to 2150, is presented in Table 3.15. The 235U and 239Pu are present in these waste streams in quantities many times the minimum mass required for criticality under optimum conditions. However, as noted above, the ceramic wasteform would include hafnium, gadolinium and 238U, which, while they remain in the presence of the fissile material (i.e., are not depleted or degraded and removed after disposal), would increase the minimum mass required for criticality. The 241Pu is also present in the Pu waste streams at greater than the relevant minimum critical mass under ideal conditions for criticality. However, the 241Pu would be dispersed in small quantities amongst the 3,428 Pu waste packages and has a short half-life. Therefore, 241Pu would not contribute significantly to any accumulation of fissile material after disposal.

The U/Pu Derived Inventory Reference Case comprises three Pu waste streams (MPu001, MPu002 and MPu004) and one high-enriched U waste stream (MU001), with Pu and U masses per waste package as shown in Table 3.16. The HEU has an average enrichment of 96 wt% 235U and the waste packages contain on average 27.9 kg of 235U. The separated Pu data show that the average package 239Pu content is 16 to 20 kg with effective 239Pu enrichments greater than 99 wt%. The Pu also includes uranium isotopes with 235U enrichments greater than 96 wt%, although with average package masses of less than 100 g 235U. The inclusion of 238U in the ceramic wasteform will reduce these enrichments. The likelihood of post-closure criticality requires detailed consideration for HEU and Pu waste packages.

Table 3.15: Key fissile nuclides in the Pu and HEU Derived Inventory Reference Case. Radionuclide activities at 2150 are from [44, Tables 7.2 and 7.3], specific activities are from [35, Table II.1], and half-life data are from [36, Table A.1]. Calculated minimum critical masses for fissile nuclides when in the form of optimally water-moderated, water-reflected spheres are from [37, Table 5].




Pu HEU Mass at

2150 (kg)


Inventory

Mass at 2150 (kg)


Inventory 235U 7.04E+08 8.01E+04 8.20E-01 3.01E+02 367 1.39E+03 1,689 239Pu 2.41E+04 2.31E+09 5.31E-01 6.55E+04 123,371 0.00E+00 0.00 241Pu 1.44E+01 3.82E+12 2.44E-01 3.61E+00 15 0.00E+00 0.00



Table 3.16: Pu and HEU waste streams in the Derived Inventory Reference Case, with isotope data provided for an indicative GDF closure date of 2150 on a per package basis. The colour scales for 239Pu, 241Pu, 233U, 235U, total fissile mass, 235U enrichment and 239Pu effective enrichment indicate relative low (green) to high (red) values. The Pu and HEU wastes are assumed to be packaged in a copper or steel disposal canister. There is no graphite, beryllium or 242mAm recorded in any of these waste streams. Waste package numbers have been rounded up.

Waste Stream

ID

Waste Stream

Description

Number of Waste

Packages

Mass per Waste Package Total Fissile Mass (233U+ 235U+

239Pu+ 241Pu) (g)

Total U (g)

235U/U (%)

239Pu/ (239Pu+U)

(%) 238Pu (g) 239Pu (g) 240Pu (g) 241Pu (g) 242Pu (g) 233U (g) 234U (g) 235U (g) 236U (g) 238U (g)

Pu waste streams

MPu001 Plutonium from Magnox fuel reprocessing

2,812 0 2.00E+04 7.20E+03 7.89E-01 3.51E+02 7.62E-03 1.25E-04 9.27E+01 1.24E+02 3.10E+00 2.01E+04 2.20E+02 42.2 98.9

MPu002 Plutonium from AGR fuel reprocessing

515 5.57E+01 1.57E+04 8.90E+03 2.68E+00 1.47E+03 1.09E-02 1.18E+02 6.48E+01 1.36E+02 4.93E-01 1.58E+04 3.20E+02 20.3 98.0

MPu004 Other plutonium

101 7.37E+01 1.70E+04 1.04E+04 3.35E-01 7.06E+02 3.76E-03 1.88E+02 7.78E+01 1.77E+02 3.20E+00 1.71E+04 4.46E+02 17.5 97.4

HEU waste streams

MU001 HEU from civil nuclear programmes

50 0 0 0 0 0 0 2.79E+02 2.79E+04 0 1.19E+03 2.79E+04 2.94E+04 95.0 0.0



4 Likelihood of Criticality in a GDF in Higher-strength Rock

4.1 Introduction This section summarises the analysis of the likelihood of criticality following the disposal of LLW/ILW/DNLEU and SF/HLW/HEU/Pu in higher-strength rock. First, a description of the illustrative design for geological disposal of these wastes in higher-strength rock is provided. Subsequently, LLW/ILW/DNLEU and SF/HLW/HEU/Pu are discussed in turn in terms of:

• The expected evolution of conditions and identification of criticality FEPs and scenarios.

• The evaluation of GDF post-closure criticality scenarios to determine the likelihood of criticality.

Full details of the analysis are presented in the detailed technical reports [1, 2].

4.2 Illustrative Disposal Concept for Higher-strength Rock As discussed in Section 2.1, the illustrative designs for disposal of the Baseline Inventory are based on the assumption that separate disposal areas will be developed for LLW/ILW and HLW/SF. The illustrative design layout for a GDF in higher-strength rock is shown in Figure 4.1 [17]. A closure date of 2150 has been assumed for such a facility.

It is assumed that UILW and SILW/LLW packages would be placed in separate vaults within the LLW/ILW disposal area [17, §2.5]. The DNLEU packages would also be disposed of in UILW vaults [17, §8.4] and it has been assumed that they would be emplaced in dedicated vaults. Based on the discussion in Section 3.2, post-closure criticality involving wastes in the SILW/LLW disposal vaults is not credible and, for this reason, the SILW/LLW vaults are not considered further.

The waste packages in the UILW vaults would be stacked in arrays seven packages high. The vaults would be backfilled using Nirex Reference Vault Backfill (NRVB), a cement-based material [17, §10.2]. Low-permeability seals consisting of highly compacted bentonite retained by a concrete structure would be constructed to isolate vault modules, disposal areas, shafts and the access drift [17, §10.4]. Higher permeability material (crushed rock, sand or gravel) would be used as bulk infill between the low permeability seals in the LLW/ILW disposal area. A cross-section through a UILW disposal vault is shown in Figure 4.2.

The disposal area for HLW/SF consists of disposal tunnels designed for in-tunnel vertical emplacement of individual disposal canisters in deposition holes, each surrounded by pre-compacted bentonite blocks and rings (see Figure 4.3) [17, §8.1.1]. It is assumed that the different types of waste (HLW, AGR SF and PWR SF) would be placed in separate disposal tunnels and potentially separate disposal modules. The disposal canisters for Pu and HEU would be treated as HLW/SF and disposed of in the HLW/SF disposal area of the GDF [17, §8.4], also in separate disposal tunnels.



Figure 4.1: Illustrative underground layout for a GDF accommodating the Baseline Inventory in a higher-strength host rock [17,

Figure 33].



Figure 4.2: Schematic cross-section through a UILW vault in a higher-strength host rock

[17, Figure 20].

Figure 4.3: Schematic cross-section through a HLW/SF disposal tunnel in higher-strength

rock [17, Figure 30].



When all of the deposition holes in a disposal tunnel have been filled, the tunnel would be backfilled with a mixture of crushed rock (70%) and bentonite (30%) [17, §10.3]. Low-permeability seals consisting of highly compacted bentonite retained by a concrete structure would be constructed to isolate vault modules, disposal areas, shafts and the drift [17, §10.4]. Higher permeability material (70% crushed rock and 30% bentonite) would be used as bulk infill between the low permeability seals in the HLW/SF disposal area [17, §10.5].

4.3 GDF Post-closure Criticality FEPs and Scenarios – LLW/ILW/DNLEU

Compliance with limits placed on the fissile material content of waste packages and other packaging constraints will ensure that a criticality event cannot occur during waste transport and that the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during storage and emplacement operations. These limits also ensure that criticality is prevented for such time as the waste packaging affords a high level of containment after disposal. However, as waste packages and GDF barrier materials begin to degrade, fissile and other materials may be mobilised, which could affect reactivity. Although anticipated changes in the arrangement of waste package materials could reduce disposal system reactivity through dispersion of fissile material, the likelihood of higher reactivity systems developing requires assessment. The following sub-sections discuss the expected evolution of conditions in the LLW/ILW/DNLEU disposal area of a GDF in higher-strength rock and identify post-closure criticality FEPs and scenarios. Note that, although the analysis has focused on expected evolution, uncertainties have been accounted for in definitions of parameter value distributions. However, in order to develop a reasonable understanding of the likelihood of criticality in a GDF, the deliberate selection of pessimistic bounding parameter values (as might be done in a criticality safety assessment) has been avoided where possible.

4.3.1 GDF Evolution

The expected evolution of conditions in a UILW disposal area is discussed in detail in the supporting technical report on LLW/ILW/DNLEU disposal [1, §4.3.1]. Key factors are discussed here.

After backfilling, the UILW disposal area will start to resaturate [46, §A1.5.2; 47, §4.4]. Groundwater flow into the disposal area would be predominantly through fractures in the rock; the rock matrix would be expected to have very low permeability. Full resaturation of the disposal facility is expected to take a few decades to a few centuries [46, §A1.5.2]. The incoming groundwater would rapidly equilibrate with the NRVB cement-based backfill resulting in the development of alkaline conditions (pH of about 12.5).

At the depth of the GDF, the incoming groundwater will be reducing (i.e., will contain no free oxygen). At closure, there will be some oxygen in the disposal vaults introduced while the system was open to the surface, but corrosion reactions will consume oxygen and will contribute to the re-establishment of reducing conditions [46, §A1.5.2]. Where microbial populations become established, microbial activity will also consume oxygen and influence the extent to which conditions are reducing. However, the harsh alkaline and nutrient poor environment of a cementitious vault would generally limit microbial activity.



Once the vaults have resaturated, small concentrations of uranium and plutonium could begin to be released into the backfill through vents in the containers, but the alkaline and reducing conditions will limit their solubility. Strong sorption to corrosion products [46, §3.15], encapsulant material and backfill under the prevailing geochemical conditions will slow migration of uranium and plutonium through the near-field environment.

Organic complexing agents derived from the degradation of cellulosic materials in the wastes can increase radionuclide mobility, but they would become progressively less significant as their concentrations are reduced by physical removal in flowing groundwater and/or by microbial degradation [46, §A1.5.2]. Naturally-occurring colloidal organic materials dissolved in groundwater may also affect radionuclide mobility. However, the ionic strength of cement porewater will normally be sufficiently high to suppress colloid formation and stability, and colloid-enhanced radionuclide transport is therefore not expected to be significant in cement-based barriers [48, §5.1.2].

The rate of release of uranium and plutonium from waste packages is likely to increase as the waste packages corrode and degrade. A low general corrosion rate is expected for stainless steel in the presence of a cement-based backfill (0.01 to 0.1 µm/year) [32, §8.1.1]. Thus, waste containers with a wall thickness of a few millimetres would be able to provide significant containment for 10,000 to 100,000 years. Controls on chlorides and relative humidity will ensure that localised corrosion during the operational phase is not significant and the potential for chloride-induced corrosion would become significantly lower once oxygen is consumed and anaerobic conditions are established.

Therefore, it could be expected that the waste packages would maintain their integrity for periods of tens of thousands to hundreds of thousands of years. In this time, the wasteform would become fractured, but is unlikely to dissolve significantly under expected chemical conditions. After waste package failure, flow could become established through the wasteform, with wasteform permeability dominated by the connected fracture network.

With time, the wasteforms and the NRVB would become degraded through reaction with the host-rock groundwater and eventually the pH in the LLW/ILW disposal area would fall to a level where the chemical barrier provided by the NRVB backfill loses some of its effectiveness. Dissolution of calcium silicate hydrate (CSH) gels by reaction with groundwater would give a pH that decreases gradually from 12.5 to about 10.5. However, the quantity of cementitious backfill in the LLW/ILW disposal area would probably be sufficient to maintain highly alkaline conditions for 100,000 years and the chemical barrier would remain at least partially effective on timescales of up to 1,000,000 years [46, §3.11]. The illustration in Figure 4.4 summarises how a cement-based engineered barrier system is expected to evolve over different post-closure timeframes.

Settling of solid plutonium and uranium through waste packages could occur if voids form as grout is dissolved and removed from waste packages. However, analysis of the formation of voids under GDF conditions has found that, if it occurred at all, the removal of sufficient material to cause slumping and accumulation of fissile material would take well in excess of 1,000,000 years [1, Appendix F.2].

The transport of plutonium and uranium through the near field will be affected by cracking of the backfill. Cracks would tend to channel groundwater flow, and plutonium and uranium would be transported in the cracks, with advection at lower rates in the matrix. Dissolved



uranium and possibly plutonium will eventually be released to the host rock, where migration would be retarded by sorption and diffusion into the rock matrix.

Figure 4.4: Expected evolution of a cement-based engineered barrier system [46, Figure

10].



4.3.2 Criticality FEPs and Scenarios

Based on the above understanding of the expected evolution of conditions in a GDF in a higher-strength host rock, criticality FEPs have been identified in terms of events and processes that could result in changes in reactivity after GDF closure. The criticality FEP analysis is described in detail in the supporting technical report on LLW/ILW/DNLEU disposal [1, Appendix C.2]. The analysis has been structured according to:

• FEPs that could result in water entry into a disposal package (i.e., the presence of ventilation openings or the occurrence of package failure mechanisms).

• FEPs that could result in changes in reactivity following water entry into the disposal package (i.e., degradation leading to relocation of fissile material, neutron absorbers, neutron reflectors and neutron moderators).

• FEPs that could result in the migration and accumulation of fissile material outside the package (i.e., accumulation by precipitation, sorption, filtration or gravitational settling).

• FEPs that could result in the migration and accumulation of fissile material from more than one disposal package.

Each of these criticality FEP groups is discussed in detail in the supporting technical report [1, Appendix C.2] in terms of mechanical, chemical, thermal, gas-related, hydrological, radiological, and microbiological events and processes.

For post-closure criticality to occur, package failure followed by substantial degradation and relocation of wasteform materials would be required. Relocation of fissile and other materials to form a critical configuration could be envisaged as occurring in a disposal package or at some location outside the disposal package and, for the latter, could involve mixing and accumulation of fissile material from several disposal packages. These considerations have formed the basis of criticality scenario construction based on sequences and combinations of criticality FEPs noted above. The criticality scenarios have been defined at a high level as:

• Scenarios involving fissile material accumulation inside a waste package.

• Scenarios involving fissile material accumulation outside a waste package.

• Scenarios involving mixing of fissile material from many waste packages.

The key stages in the development of these scenarios are illustrated in the event tree shown in Figure 4.5. Note that uniform chemical conditions have been assumed in the scenario modelling. Therefore, although the precipitation process is represented in the models, the conditions required for precipitation are not modelled. However, as described in the detailed report [1, §4.3.3], it is likely that the current calculations using generic sorption parameters would bound the effects of precipitation.



Figure 4.5: Event tree showing fissile material accumulation scenarios for LLW/ILW/DNLEU disposal in higher-strength rock. The

yellow boxes indicate criticality scenarios involving fissile material from a single package and the blue boxes indicate scenarios involving fissile material from multiple packages.

Multiple Package Scenarios

* Radioactive decay of 239Pu to 235U may be significant for all scenarios† Accumulation may occur as a result of sorption, precipitation, mineralisation, filtration or settling, according to the material properties and geochemical and hydrogeological conditions.


Outside Package Scenarios

Inside Package Scenarios

Water gradually enters the disposal package through a vent or defect and saturates the waste

Mechanical damage or container corrosion results in establishment of flow paths through the package

The bulk wasteform degrades (corrosion and dissolution reactions, including dissolution of Pu* and U)

Mobile material is removed from the package

When sufficiently weakened, the degraded wasteform slumps in the package

Degraded solid materials (including particulate, sorbed, precipitated or co-precipitated fissile material) remain in the package mixed with water

Dissolved fissile material removed

Colloidal fissile material removed

Particulate fissile material removed

Fissile material migrates by advection and diffusion through the host rock and may accumulate†

When sufficiently weakened, wastes in a stack of degrading packages progressively slump towards the vault floor

Fissile material from multiple packages may accumulate in the backfill†

Fissile material from multiple packages may accumulate in the system seals†

Fissile material from multiple packages may accumulate in the host rock†

Fissile material migrates by advection and diffusion through the system seals and may accumulate†

Fissile material migrates by advection and diffusion through the backfill and may accumulate†

Fissile concentration and mass calculation stage (single package). Fissile concentration and mass calculation stage (multiple packages).



4.4 GDF Post-closure Criticality Scenario Assessment – LLW/ILW/DNLEU

This section summarises the analysis of the likelihood of criticality following the disposal of LLW/ILW/DNLEU in higher-strength rock, which is presented in detail in the project report on LLW/ILW/DNLEU disposal [1, §4.4]. Consideration is given to GDF post-closure criticality scenarios that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package (Figure 4.5).

4.4.1 Single Package Model

Conceptual models for criticality scenarios involving increases in reactivity inside a waste package and surrounding a waste package have been derived. The conceptual models have been implemented in a single package-scale GoldSim model [1, §4.4.1].

The package-scale model consists of a single waste package (assumed to be a 500 litre drum, a 3 m3 box or a 3 m3 drum) surrounded by saturated NVRB. Masses of 235U, 238U, 239Pu and grout are specified for the waste package. These materials are assumed to be uniformly distributed within the waste package and contained until the container is breached by general corrosion. The time until waste package breach depends on the thickness and general corrosion rate of the container wall. At the time of container breach, uranium and plutonium are assumed to be dissolved to solubility limits and sorbed to the encapsulation grout according to distribution coefficients.

After container breach, the dissolved uranium and plutonium may be transported out of the package in flowing groundwater into the surrounding NRVB. The presence of organic complexants in the porewater, either derived from the waste or in the incoming porewater, may increase radionuclide solubility. Also, the encapsulant grout is removed from the waste package by dissolution and advection. After a fraction of the grout has been dissolved and removed from the waste package, it is assumed that the wasteform is unable to support itself and the material remaining in the waste package slumps to the base of the waste package.

The different stages in waste package evolution are depicted conceptually in Figure 4.6. The figure indicates (a) an intact container at the time of GDF closure t0 that is gradually saturating as water enters through gas vents or a package defect; (b) degradation of the container to the extent that water is able to move through the wasteform at breach time tb (as depicted by the arrows); (c) dissolution of wasteform materials and their removal in circulating groundwater, which increases wasteform porosity and leads to some settling of remaining solids; (d) eventual slumping at time tc when a limiting porosity is achieved at which the remaining wasteform is assumed to be unable to support itself, leading to (e) relocation of the wasteform towards the base of the waste package with potential collapse of the waste package.



Figure 4.6: Dissolution, settling and slumping of an LLW/ILW/DNLEU wasteform inside a

disposal package.

After release from the waste package, uranium and plutonium are advected in one dimension through the NRVB where they can sorb and precipitate. The presence of cellulosic wastes may affect distribution coefficients. The grout released from the waste package is assumed not to affect the system. The GoldSim solution grid for radionuclide transport through the saturated NRVB comprises twenty identical 0.2-m-long cells, generating a 4-m-long segment of NRVB downstream from the waste package. The grid has a cross-sectional area equal to the cross-sectional area of the waste package. The masses of radionuclides in each cell (and thus the concentrations) are evaluated based on addition of the masses entering from the upstream cell and subtraction of the masses transferring to the downstream cell. The final cell in the pathway acts as a sink for material transported through the system because there is no flux of radionuclides across the final cell boundary. The implementation of the model in GoldSim is illustrated in Figure 4.7.

Note that uniform chemical conditions have been assumed in the analysis. Therefore, although the precipitation process is represented in the models, the conditions required for precipitation are not realised. However, the accumulation of quantities of fissile material large enough to result in criticality by a process of ‘sorption’ clearly requires the deposition of multiple layers from solution. The transition between multi-layer sorption and precipitation could be defined to have occurred when the properties of the underlying substrate are no longer relevant to the behaviour of dissolved species. Given the heterogeneity and variability of natural systems, and the site-specific nature of groundwater compositions, it would be difficult to obtain data that could reliably distinguish between the two mechanisms. Indeed, given the range of dissolved species that might be present in a GDF, resulting in chemical competition for surface interactions, it is likely that the current calculations using generic sorption parameters would bound the effects of precipitation [1, §4.4.1].



Figure 4.7: Solution grid for the GoldSim package-scale model.

Note also that concentrations of natural uranium in groundwater have been assumed to be insignificant. If water entering the disposal vaults is saturated in natural uranium then the rate of dissolution of uranium in the disposal vaults would be greatly reduced. Furthermore, if uranium sorption (or precipitation) entirely fills the NRVB pore space, there would be no neutron-moderating water present. Reactivity would be greatest when the uranium to water ratio is optimum for criticality and not necessarily when the uranium concentration in NRVB is greatest. Indeed, in order to determine if criticality is possible, the concentrations and masses of fissile material in NRVB calculated using the GoldSim model are compared with the concentrations and masses required for criticality (at the relevant enrichment) for water-moderated, water-reflected spheres of fissile material in NRVB.

The GoldSim modelling approach enables parameter value uncertainty to be accounted for by randomly sampling parameter value probability distributions over many code runs (realisations). Each model application has involved undertaking 1000 such realisations in order to ensure adequate sampling of the parameter value distributions.

4.4.2 Reactivity Increase in a Waste Package

The generic CSAs for waste packages containing ILW-INU [42], ILW-LEU [41], ILW-HEU [40] and ILW-Pu [43] included calculations of the critical mass of fissile material that has been assumed to settle to the base of the waste package. These critical masses are shown in Table 4.1 for different categories of ILW packaged in 500 litre drums. The critical masses have been compared with the results of the GoldSim calculations to determine if critical systems could form. The calculations were undertaken for waste packages that include the highest average fissile masses according to the data shown in Table 3.3, Table 3.4 and Section 3.3.2, as summarised in Table 4.2.

U, Pu and grout dissolution and release

Waste package cell (U, Pu and grout masses specified)

NRVB cells (20 in total)

Advection of dissolved Pu and U across cell surfaces

Zero mass transfer boundary



Table 4.1: Minimum critical masses derived in the generic CSAs for different waste categories assuming that the fissile material has slumped to the base of the waste package (500 litre drum in this case) to form water-moderated, water-reflected configurations.

Enrichment Range Minimum Critical Mass

ILW-INU (up to 1.9 wt% 235U) 4.9 kg 235U [42, Table 5.5].

ILW-LEU (1.9 to 4.0 wt% 235U) 3.3 kg 235U [41, Table 5.5].

ILW-HEU (Greater than 4.0 wt% 235U) 2.1 kg 235U [40, Table 5.5].

ILW-Pu (up to 100 wt% 239Pu) 1.3 kg 239Pu [43, Table 5.5].

Table 4.2: Assumed contents of waste packages for key waste streams based on the derived 2007 UK Radioactive Waste Inventory (see Table 3.3, Table 3.4 and Section 3.3.2). Grout mass per package is based on a 500 litre drum volume of 0.5 m3 and a grout density of 2,300 kg/m3.

Property Value Waste materials for waste stream 239Pu 235U 238U Grout 5B23: DFR Breeder (ILW-INU) 1.9 kg 1.4 kg 220 kg 1150 kgMU006: LEU 0 kg 19 kg 950 kg 1150 kg5C54: Zenith Fuel (ILW-LEU) 0 kg 13 kg 240 kg 1150 kg5B350: Uranium Recovery Plant ILW (ILW-HEU) 0 kg 11 kg 4.5 kg 1150 kg5B31: Plutonium Nitrate Waste (ILW-Pu) 2.6 kg 0.01 kg 1 kg 1150 kg2N02: Plutonium Contaminated Material; Crated (ILW-Pu / ILW-HEU)

0.84 kg 0.18 kg 0.0 kg 1150 kg

2D03: Plutonium Contaminated Material; Drums (ILW-Pu) 0.1 kg 0.4 kg 0.01 kg 1150 kg

Most of the ILW-INU waste packages shown in Table 3.4 include on average less than the relevant minimum masses of fissile material calculated in the generic CSAs for package-scale scenarios; criticality is not credible for ILW-INU waste packages that contain less than a minimum fissile mass. Waste stream 5B23 (DFR Breeder) is an exception, but only under situations in which the uranium is dissolved and separated from the plutonium.

Calculations for a 5B23 (DFR Breeder) waste package indicate that, for all realisations, the mass of 239Pu in a waste package reduces to the critical mass of 1.3 kg (see Table 4.1) after about 14,000 years, as shown in Figure 4.8. This reduction is almost entirely due to radioactive decay, with very little plutonium dissolution and removal. In-package criticality after this time would not be possible even if the uranium and grout were removed from the waste package and slumping occurred. It may be possible for the uranium to be dissolved and removed from the waste package in this time, as indicated in Figure 4.9. However, significant grout dissolution and removal is unlikely to occur on this timescale as indicated by Figure 4.10 and therefore package-scale criticality involving ILW-INU waste packages is not credible based on the average package contents given in the Reference Case Derived Inventory.



0.0E+00

2.0E+02

4.0E+02

6.0E+02

8.0E+02

1.0E+03

1.2E+03

1.4E+03

1.6E+03

1.8E+03

2.0E+03

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06

Pu-2

39 M

ass

(g)

Time (yr)

Minimum Mean Maximum Minimum critical mass for slumped system

Figure 4.8: Mass of 239Pu remaining in a degrading waste package (5B23 DFR Breeder). After 14,000 years, the 239Pu mass is less than the minimum required for criticality following slumping.

0.0E+00

5.0E+04

1.0E+05

1.5E+05

2.0E+05

2.5E+05

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06

U-2

38 M

ass

(g)

Time (yr)

Minimum 5% Mean Maximum Time = 14,000 years

Figure 4.9: Mass of 238U remaining in a degrading waste package (5B23 DFR Breeder).



0.0E+00

2.0E+02

4.0E+02

6.0E+02

8.0E+02

1.0E+03

1.2E+03

1.4E+03

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06

Gro

ut M

ass

(kg)

Time (yr)

Minimum Mean Maximum Time = 14,000 years

Figure 4.10: Mass of grout remaining in a degrading waste package (5B23 DFR Breeder).

The ILW-HEU and ILW-LEU waste packages that include on average more than the relevant minimum critical mass of fissile material are MU006 LEU, 5C54 Zenith Fuel (ILW-LEU), 5B350 Uranium Recovery Plant ILW (ILW-HEU), 5C50 Dragon Fuel (ILW-HEU) and 5B19 Uranium Contaminated Materials (ILW-HEU).

It is not possible to demonstrate that post-closure in-package criticality is of low likelihood for these waste packages based on a simple qualitative argument given their contents and expected evolution. Alternative packaging concepts (as being developed for Dragon fuels 5C50) or analysis that includes the presence of low solubility neutron absorbers or diluents may be required to demonstrate sub-criticality. GoldSim probabilistic calculations have been performed for waste streams MU006 LEU, 5C54 Zenith Fuel and 5B350 Uranium Recovery Plant ILW (see Table 4.2 for assumed waste package contents). Not surprisingly, in all cases where low uranium solubilities were sampled, masses of uranium potentially sufficient for criticality (according to the values in Table 4.1) were calculated to be present in the waste packages on timescales in which the grout could be dissolved and removed (see Figure 4.10). For example, results for the MU006 LEU waste package are shown in Figure 4.11.

The only ILW-Pu waste packages that on average include sufficient fissile material to challenge the minimum fissile masses are those from waste stream 5B31 Plutonium Nitrate Waste. Based on the GoldSim probabilistic calculations for this waste stream, slumping of more than 2.1 kg 235U could occur on timescales of 106 years following decay of the 239Pu to 235U (Figure 4.12). The calculated mass of 235U has a broad range over all realisations because of the broad range of solubilities and flow rates sampled. Criticality as a result of slumping requires that uranium solubility is low.



0.0E+00

2.0E+03

4.0E+03

6.0E+03

8.0E+03

1.0E+04

1.2E+04

1.4E+04

1.6E+04

1.8E+04

2.0E+04

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06

U-2

35 M

ass

(g)

Time (yr)

Minimum 5% Mean Maximum Minimum critical mass for slumped system

Figure 4.11: Mass of 235U remaining in a degrading waste package (MU006 LEU).

0.0E+00

5.0E+02

1.0E+03

1.5E+03

2.0E+03

2.5E+03

3.0E+03

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06

U-2

35 M

ass

(g)

Time (yr)

Minimum Mean Maximum Minimum critical mass for slumped system

Figure 4.12: Mass of 235U remaining in a degrading waste package (5B31 Plutonium Nitrate Waste).



4.4.3 Accumulation outside a Waste Package

Any fissile material released from a waste package would migrate through the NRVB and potentially through seals and the host rock. Mechanisms such as sorption and precipitation could tend to cause accumulation of fissile material in these regions. The GoldSim model has been used to calculate fissile material transport around waste packages. Results indicate that, by comparison of the calculated mean concentrations and masses of fissile material in the backfill with minimum concentrations and masses (Table 4.3) required for criticality, conditions are generally expected to remain substantially sub-critical. For example, only minor amounts of 239Pu would be released from a degrading ILW-INU waste package and therefore criticality would not occur as a result of accumulation of 239Pu outside a waste package.

However, conditions are not demonstrably sub-critical for ILW-HEU waste streams 5B350 (Uranium Recovery Plant ILW) and ILW-Pu waste stream 5B31 (Plutonium Nitrate Waste) for all realisations. That is, for the small number of realisations in which uranium solubility and sorption are high, the 235U concentrations exceed the estimated minimum required for criticality in backfill. Critical concentrations of 235U could also occur in NRVB following release from waste stream MU006 LEU waste packages, ILW-LEU waste packages from waste stream 5C54 Zenith Fuel, ILW-HEU waste packages from waste streams 5C50 (Dragon Fuel) and 5B19 (Uranium Contaminated Materials), and ILW-Pu waste packages from waste streams 2N100 (Plutonium Contaminated Material; Drummed) and 2N02 (Plutonium Contaminated Material; Crated).

Waste stream 5B350 includes organic material and therefore radionuclide solubilities may be enhanced by the presence of organic complexants. Indeed, calculations assuming a 10% organics loading show that there is a greater potential for critical concentrations of 235U to occur in NRVB because of the effects of organic complexants. The calculated mass of 235U in the backfill cells after 200,000 years, when the 235U concentration is greatest, is shown in Figure 4.13. In the backfill cell close to the waste package, the 95% value is greater than the estimated minimum required for criticality (0.9 kg 235U as indicated in Table 4.3).

Also, when the potential effects of organic loading on radionuclide solubility and sorption are taken into account for ILW-Pu waste stream 2N02, for a small number of realisations in which there is early waste package breach and plutonium and uranium solubility are high, the sum of the 239Pu and 235U masses are greater than the estimated minimum required for criticality in backfill.

Table 4.3: Minimum critical masses and diameters of spheres of uranium and plutonium in saturated NRVB at different enrichments [49, Figure 4.2].

Enrichment Minimum Critical Mass Diameter

3 wt% 235U 4 kg 235U ~60 cm

10 wt% 235U 2 kg 235U ~40 cm

100 wt% 235U 0.9 kg 235U ~30 cm

100 wt% 239Pu 0.6 kg 239Pu ~30 cm



1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

0.0E+00 5.0E-01 1.0E+00 1.5E+00 2.0E+00

Mas

s of

U-2

35 (g

)

Distance from package (m)Mean at 200,000 years 95% at 200,000 years

Maximum at 200,000 years Minimum critical mass

Figure 4.13: Mass of 235U in backfill downstream from a degrading waste package (5B350 Uranium Recovery Plant ILW) after 200,000 years in the presence of organic complexants (10% loading).

4.4.4 Multiple Package Model

The release, migration and mixing of fissile material from many waste packages could result in the generation of systems of higher reactivity than at the time of disposal. Criticality scenarios involving slumping of fissile material through a stack of degrading waste packages and accumulation of mobilised fissile material in engineered or natural barriers have been identified. These fissile material accumulation processes have been represented in a single GoldSim model representing conditions in a cross-section of an ILW disposal vault [1, Appendix D.3.2].

The vault-scale model is a development of the single package model to include representation of stacks of waste packages surrounded by NRVB in a disposal vault. The model has been implemented in GoldSim and is used to evaluate scenarios involving accumulation of fissile material from many packages either by gravitational settling and slumping towards the vault floor or by migration of mobile material in groundwater and its accumulation at locations within waste packages, backfill or surrounding host rock.



Figure 4.14: Unit cell containing (from left to right) a 500 litre drum stillage, a 3 m3 box or a

3 m3 drum. The waste packages are surrounded by saturated NRVB.

The model is two-dimensional, comprising a vertical array of waste packages (500 litre drums, 3 m3 boxes and 3 m3 drums) surrounded by saturated backfill. This simplified modelling approach aims to balance representation of important features and processes at a suitable level of detail in a generic context, while ensuring that the probabilistic approach to addressing uncertainty is computationally tractable.

The waste packages are represented in unit cells that contain either four 500 litre drums in a stillage, a single 3 m3 box or a single 3 m3 drum, as shown in Figure 4.14. The model of the vault cross-section contains 49 such cells in a 7x7 vertical array. The type of waste package present in a unit cell is selected randomly based on the number of stillages, 3 m3 boxes and 3 m3 drums to be emplaced in the GDF.

The processes included in the multi-package model are similar to those included in the single package model. The wasteform is assumed to be uniformly distributed within each waste package and contained until the container is breached by general corrosion. At the time of container breach, uranium and plutonium are assumed to be dissolved to solubility limits and sorbed to the encapsulation grout according to distribution coefficients.

After container breach, dissolved uranium and plutonium may be transported out of the package into the surrounding NRVB. Also, the encapsulant grout is removed from the waste package by dissolution and advection. Mobile plutonium and uranium that leaves the waste packages is advected through the backfill and any breached waste packages towards the host rock. The grout released from the waste package is assumed not to affect the system.

Groundwater flow through the vault is assumed to be through a network of fractures, as indicated in Figure 4.15 to generate flow and mass transfer connections in a seven by seven array of model cells as shown in Figure 4.16.



Figure 4.15: A group of four unit cells within a seven by seven array of cells in an ILW vault cross-section. Vertical and horizontal fractures along which mobile uranium and plutonium migrates between unit cells are shown. A unit cell contains four 500 litre drums, a 3 m3 box or a 3 m3 drum and surrounding NRVB.

Figure 4.16: Groundwater flow and mass transfer connections in the seven by seven array of cells and accumulation zone in the GoldSim mode. The cell numbering system is shown.

Accumulation

zone

Fractures

Unit cells containing ILW packages and NRVB

Flow and mass transfer through fractures



The vertical and horizontal flow rates are assumed to be the same for each model cell, but the flow rate is re-sampled for each realisation. The rate of transfer of plutonium and uranium from a waste package may be increased if the radionuclides are transferred in colloid form. The presence of organic complexants in the porewater, either derived from the waste or in the incoming porewater, may increase radionuclide solubility.

The masses of radionuclides in each unit cell are evaluated based on addition of the masses entering from the upstream cells and subtraction of the masses transferring to the downstream cells. Dissolved uranium and plutonium entering a unit cell may sorb on the mass of NRVB in the cell and on the mass of grout in the waste package (or packages) in the cell if the waste package(s) have been breached.

Radionuclides that leave the array of waste packages at the downstream boundaries are assumed to collect and mix together in an accumulation zone, as shown in Figure 4.16. No specific material properties are assigned to this zone; it simply represents a void that provides a sink for all uranium and plutonium that leaves the waste package stack. The accumulation zone could be interpreted as a void space or fracture zone in the backfill, a vault seal, a GDF tunnel or the host rock. The accumulation process could be interpreted as being precipitation at a location where there is a change in geochemical conditions. This represents a cautious modelling approach in that it is difficult to envisage a situation in which all of the uranium and plutonium migrating from a vertical array of waste packages could be channelled towards a single location where geochemical conditions are such that an irreversible accumulation process, such as precipitation, occurs.

Whether the accumulated material could result in criticality can be estimated by comparison of the mass of fissile material in the accumulation zone with the minimum required for criticality in different media based on assumptions about the configuration of the accumulation (e.g., water-moderated, water-reflected spherical or slab configurations in backfill or fractured rock).

The process of accumulation of fissile material by slumping through multiple degraded waste packages is illustrated in Figure 4.17. The scenario involves progressive gravitational settling of material through a stack of waste packages as the containers corrode and wasteform and backfill materials are removed in groundwater. Slumping of a wasteform is assumed to occur when a specific proportion of the grout has been removed from a waste package. In the present analysis, it has been assumed that slumping occurs when 50% of the initial grout mass has been dissolved. The solid uranium and plutonium remaining in a waste package and surrounding backfill at the time of slumping are assumed to move to the base of the model cell and then into an underlying cell if and when slumping occurs in the underlying cell. Thus, uranium and plutonium progressively settle towards the vault floor as slumping occurs in different waste packages, as illustrated in Figure 4.17. Any uranium or plutonium that later enters the slumped cell from an upstream cell is assumed to be transferred to the base of the slump. Uranium and plutonium in a slumped cell is advected in fractures to downstream cells.



Figure 4.17: Illustration of how uranium and plutonium slump as grout is removed from degrading waste packages for an example showing five waste packages in a stack at different times: a) none of the waste packages has slumped; b) the third waste package has slumped c) the fourth waste package has slumped and the uranium and plutonium masses from the third waste package are transferred to the fourth waste package; d) the fifth waste package has slumped and the uranium and plutonium masses from the fourth waste package are transferred to the fifth waste package; e) the second waste package has slumped and uranium and plutonium masses from the second waste package are transferred to the fifth waste package.

Whether the slumped systems can become critical has been determined based on consideration of the minimum areal densities of fissile material required for criticality in water-moderated, water-reflected slabs at different enrichments. Slumped material has been assumed to accumulate at the base of a unit cell, but the presence of neutron absorbing materials, such as iron corrosion products, has been ignored. The minimum critical masses for slumped material at different enrichments above 1 wt% 235U are shown in Table 4.4.



Table 4.4: Minimum critical mass in a unit cell at different enrichments based on water-moderated, water-reflected infinite slabs.

Enrichment Range Minimum Areal Concentration for Criticality

Minimum Critical Mass in 2 m by 2 m cell

1.0 to 1.9 wt% 1.0x104 g 235U/m2 [50, Table A2.1.5] 235U + 1.6 239Pu = 40 kg

1.9 to 4.0 wt% 6.7x103 g 235U/m2 [51, §4.3.2] 235U + 1.6 239Pu = 27 kg

Greater than 4.0 wt% (up to 100 wt%)

4.3x103 g 235U/m2 [50, §4.3.2.2; 51, §6.2.2]

235U + 1.6 239Pu = 17 kg

4.4.5 Accumulation from Multiple ILW Packages

In the initial GoldSim calculations, the effects of organic degradation products on uranium and plutonium behaviour were excluded. Despite the accumulation of fissile material by slumping, the minimum critical mass at the relevant enrichment (see Table 4.4) was not exceeded at any location during the course of each realisation. Therefore, criticality by slumping is not credible based on the understanding of fissile material migration and accumulation processes represented in the waste package slumping analysis and assuming a random waste package inventory and the absence of organic degradation products.

The slumping process can be seen for a stack of waste packages (defined as Cells 7, 14, 21, 28, 35, 42 and 49 from top to bottom in the stack) in the evolution of the mean 235U mass in each cell shown in Figure 4.18. Uranium and plutonium masses are only recorded once a waste package has corroded and failed such that the wasteform is mobilised in groundwater. Thus a gradual increase in 235U mass in each cell as waste packages fail and 239Pu decays over around 105 years is apparent in Figure 4.18. Thereafter, progressive slumping causes variations in the mean 235U mass in each cell, eventually leading to reductions in the mean 235U mass in the upper cells and an increase in the mean 235U mass in the lowest cell (Cell 49). The masses of 235U in each cell in the stack are shown for a single realisation in Figure 4.19 in order to illustrate the slumping process. Increments in 235U content occur in each cell at early times as packages fail and 235U is dissolved. Step changes in 235U contents occur as different slumping events occur, with 235U ultimately being transferred to the lowest waste package. The slumping events for this realisation are described in Table 4.5.



0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0.40

0.45

1.0E+03 1.0E+04 1.0E+05 1.0E+06

235 U

mass (kg)

Time (years)

Cell 7 Cell 14 Cell 21 Cell 28 Cell 35 Cell 42 Cell 49

Figure 4.18: Mean mass (over 1,000 realisations) of 235U in each cell in a stack, from Cell 7 at the top of the stack to Cell 49 at the base of the stack.

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.70

0.80

0.90

1.00

1.0E+03 1.0E+04 1.0E+05 1.0E+06

235 U

mass (kg)

Time (years)


Figure 4.19: Mass of 235U in each cell in a stack, from Cell 7 at the top of the stack to Cell 49 at the base of the stack for a single realisation.



Table 4.5: Times of waste package slumping for one GoldSim model realisation (illustrated in Figure 4.19).

Time Slumping Cell Outcome

2.4x105 years Cell 21 None, because the material is unable to move until the lower cell slumps.

3.3x105 years Cell 28 Material in Cell 21 slumps into Cell 28.

4.0x105 years Cell 14 Material in Cell 14 slumps through Cell 21 into Cell 28.

4.5x105 years Cell 49 None, because material in Cell 42 above Cell 49 has not yet slumped.

5.6x105 years Cell 42 Material in Cell 42 slumps into Cell 49.

1.4x106 years Cell 7 Material in Cell 7 slumps through Cells 14 and 21 into Cell 28.

1.7x106 years Cell 35 Material in Cell 35 slumps through Cell 42 into Cell 49 and material in Cell 28 slumps through Cells 35 and 42 into Cell 49.

For realisations in which the flow rate and uranium solubility are high, uranium can be removed from failed waste packages before slumping occurs. For example, Figure 4.20 shows the transfer of dissolved 235U through Cells 7, 14, 21, 28, 35, 42 and 49. The pulses in 235U in Cells 42 and 49 peaking at around 104 years indicate the advection of 235U from neighbouring upstream cells through Cells 42 and 49. Similar behaviour can be seen to a lesser extent in other cells and later peaks indicate the transfer of pulses of 235U from cells further upstream. In this realisation, almost the entire uranium content in the vault cross-section (about 1,000 kg uranium) is transferred to the accumulation zone in one million years and the enrichment of the uranium in this zone is never more than about 1 wt% 235U.

At the other extreme, there are also realisations for which the uranium solubility and flow rate are so low that there is little mass transfer from failed waste packages. For example, Figure 4.21 shows the results of a realisation in which the 235U content of Cells 7, 14, 21, 28, 35, 42 and 49 is little changed during the simulation period after waste package failure, apart from ingrowth from 239Pu, where present. In this case, only a few grams of uranium reach the downstream accumulation zone on a timescale of one million years.

Critical systems could only develop if the fissile material is assumed to accumulate with water in optimum spherical configurations in porous material such as backfill or grout. For example, the slumped system could be assumed to include NRVB. In about 2% of the realisations (excluding the effects of organic material degradation products), at particular locations and times the 235U mass exceeds that required for criticality in a water-moderated sphere of 235U in NRVB with a porosity of 50%. Even if all of the fissile material leaving the disposal vault cross-section accumulates in a single location (assumed to be NRVB), the minimum critical mass under optimum conditions is exceeded in only about 2% of realisations. The conditions of concern can generally be traced back to specific high-fissile-material-content waste packages identified in Section 3.2.2 and, in particular, those highlighted red in Table 3.3, which represent just under 2% of UILW packages.



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Figure 4.20: Mass of 235U in each cell in a stack, from Cell 7 at the top of the stack to Cell 49 at the base of the stack for a single realisation where the flow rate and uranium solubility are high.

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Figure 4.21: Mass of 235U in each cell in a stack, from Cell 7 at the top of the stack to Cell 49 at the base of the stack for a single realisation where the flow rate and uranium solubility are low.



There is less potential for fissile material accumulation to result in criticality in the disposal vault if organic material degradation products are assumed to affect the migration behaviour of uranium and plutonium everywhere. The increased mobility of the uranium and plutonium means that there is less available to contribute to fissile material accumulation by slumping. However, the more rapid transport of fissile material to the hypothetical accumulation zone means that there is greater potential for critical systems to develop in that zone. For example, in numerous realisations, masses of the order 1 kg 239Pu migrate to the accumulation zone. Such masses of 239Pu would not be sub-critical in a water-moderated spherical configuration in NRVB. Similarly, in a number of realisations, the 235U is present in the accumulation zone in masses that would not be sub-critical in a water-moderated spherical configuration in NRVB at the relevant enrichment.

If the hypothetical accumulation is assumed to occur in a low porosity (~1%) crystalline host rock, then at least 80 kg 235U (uranium as pure 235U) would be required for criticality under idealised conditions [53, Figure 4.3]. Such a mass of 235U does not accumulate in the hypothetical accumulation zone in any realisation.

4.5 Summary of Assumptions – LLW/ILW/DNLEU The likelihood of criticality following the disposal of LLW/ILW and DNLEU in higher-strength rock has been analysed based on consideration of the expected evolution of conditions in the GDF. Post-closure criticality scenarios have been analysed using GoldSim probabilistic models. Details of the results and conclusions of the scenario analyses are presented in the main report on LLW/ILW/DNLEU disposal [1]. The analysis has been based on many assumptions about the waste inventory, waste packaging concepts, disposal facility design, barrier system properties and GDF evolution. The main assumptions and alternative assumptions and their potential impacts on the likelihood of criticality analysis and results are discussed in Table 4.6.

A key finding of work on GDF evolution undertaken as part of this project, and discussed in detail in [1, Appendix F], is that there may be little potential for void formation under expected geochemical and hydrological conditions in a GDF. Therefore, the formation of voids such as to cause slumping may not be possible. This is a significant finding because it means that assumptions made in this project about grout dissolution rates may be cautious rather than realistic and post-closure criticality scenarios involving fissile material slumping on the package or vault scale may not be credible.



Table 4.6: Key assumptions made in the likelihood of criticality analysis for LLW/ILW/DNLEU disposal in a higher-strength host rock. Alternative assumptions and their potential impacts on the analysis are noted.

Assumptions Comment on assumptions Impact of making alternative assumptions

Inventory and waste packaging

The Derived Inventory has been used, which is based on the Baseline Inventory estimate and detailed characterisation data from the 2007 UK RWI.

The 2013 UK RWI has been published and some waste packaging concepts have been revised since the production of the 2007 UK RWI.

Changes to waste packaging concepts could have a significant impact on the criticality scenario analysis, especially where they involve reductions in the fissile material content of the waste packages deemed to be most challenging to post-closure criticality safety, or large increases in the fissile material content of waste packages that the criticality analysis indicates are currently not a concern.

The Reference Case of the Derived Inventory has been used.

Two additional inventory scenarios are available in the Derived Inventory that address uncertainties in forecasts of future waste arisings: the Lower and Upper Inventories.

The impact on the criticality scenario analysis of using the Lower and Upper Inventories is uncertain. Again, the main impacts would be associated with any significant changes in the fissile material contents of waste packages.

Average waste package contents have been assumed based on the expected radionuclide content and number of waste packages in each waste stream.

The fissile material content of waste packages in each waste stream may have a broad range. If the uncertainties in the distribution of waste package contents could be quantified then the impacts on the criticality analysis could be evaluated.

Large variations about the average fissile material content of waste packages for any one waste stream could impact the criticality analysis significantly given that the criticality scenarios analysis has confirmed that only a small number of waste packages with a high fissile material content are of concern.

ILW and DNLEU are conditioned in waste packages using a cementitious grout.

Alternative immobilisation and encapsulation matrices may be used, such as glass, ceramics and polymers.

The rates of degradation and release of fissile nuclides from these alternative immobilisation and encapsulation matrices may be slower than from grout, which would reduce the potential for fissile material post-closure accumulation.

Host rock and disposal concept

The host rock properties, disposal concept design and underpinning parameter values are consistent, as far as possible, with the 2010 DSSC.

Alternative assumptions about host rock properties (e.g., groundwater flow rates), disposal facility designs and barrier properties could be made.

The current assumptions about groundwater flow rates are considered to encompass conditions likely for a UK GDF. However, the impacts of assuming an alternative barrier system are uncertain.




For the multiple package analysis, the array of waste packages in the disposal vault has been selected based on random sampling from the population of all waste packages in the inventory.

Waste packages could be emplaced in campaigns, such that a random distribution is not representative.

If a number of similar waste packages with high fissile material content are assumed to be emplaced together, then the accumulated masses of fissile material (e.g., from slumping) may be significantly greater than those currently calculated.

Criticality scenario analysis

Package failure is assumed to occur as a result of general corrosion on a timescale of hundreds of years to hundreds of millions of years.

Other processes could lead to package failure. Seismic events have been identified as possible events that could disrupt waste packages, but such events are site specific and are difficult to evaluate in a generic analysis.

The range of corrosion rates is broad and would most likely encompass the effects of other disruptive events. Therefore, alternative assumptions would not affect the analysis significantly.

Slumping of fissile material occurs in waste packages on a timescale of several hundred thousand years to hundreds of millions of years as a result of grout dissolution (when 50% has been removed).

Analysis of the potential for void formation under expected geochemical and hydrological conditions in a GDF indicates that the formation of voids such as to cause slumping may not be possible.

If slumping is assumed not to occur then the potential for in-package criticality or criticality as a result of slumping of fissile material onto the vault floor is eliminated.

Judgments on the possibility of critical masses or concentrations of fissile nuclides occurring are based on comparisons with the conditions required for criticality in idealised systems (e.g., optimally water-moderated spheres or slabs of fissile material).

Alternative assumptions about the geometry of fissile material accumulations and the presence of other materials such as iron or graphite could be made. Analysis of the behaviour of iron corrosion products under disposal conditions has indicated that iron would remain in solid form for long periods and graphite is relatively insoluble under disposal conditions.

Alternative assumptions about materials involved in fissile material accumulations and their configurations could affect the criticality analysis significantly. The inclusion of iron and graphite in non-idealised configurations would reduce the likelihood of criticality. However, more pessimistic assumptions could be made about the location of graphite with respect to accumulated fissile material (i.e., the graphite could be assumed to form a neutron reflecting medium around the fissile material), which would mean that less fissile material would be required to accumulate for criticality to occur.




Correlation between uranium solubility and uranium sorption has been assumed based on a correlation coefficient of -0.5.

The selected correlation coefficient ensures that calculated concentrations of uranium are physically realistic. However, the correlation could be modified to further reduce the possibility of high solubility and high sorption conditions being sampled for uranium in the probabilistic analysis.

Reducing the correlation coefficient below -0.5 would reduce the likelihood of critical concentrations of uranium being calculated.

Concentrations of natural uranium in groundwater are insignificant.

If water entering the disposal vaults is saturated in natural uranium then the rate of dissolution of uranium in the disposal vaults would be greatly reduced. The issue could be addressed when site-specific data on groundwater compositions are available.

Assuming groundwater to include concentrations of natural uranium would reduce the rate of release of uranium from waste packages and the enrichment of dissolved uranium in the GDF would be lower than calculated in this analysis.

Sorption sites do not become saturated.

Uranium and plutonium sorption site saturation could be included in the model.

If sorption sites become saturated it would reduce the likelihood of critical concentrations of plutonium and uranium being calculated.

Calculations that considered the effects of organic degradation products assumed a high organic loading fraction in all waste packages and that the organic materials persist.

The approach could be modified such that the effects of organic matter (at expected loadings) are only included for waste packages that include such material and that concentration of mobile organic degradation products reduces with time.

The current approach is considered to be cautious. A more realistic approach to modelling the effects of organic matter would reduce the likelihood of critical concentrations of plutonium and uranium being calculated.

A two-dimensional flow distribution has been assumed through the disposal vault based on an assumed uniform fracture system. Flow has been assumed to occur uniformly through corroded waste packages and all flow leaving the vault has been assumed to be channelled through an accumulation zone.

Alternative flow distributions could be adopted.

The current approach is considered to be cautious in that it ensures material from all modelled waste packages could accumulate at a single location. However, the impacts of variations in flow rates and distributions are uncertain.




Uniform chemical conditions have been assumed in the vault model. Therefore, although precipitation is represented, the conditions required for precipitation are not modelled.

A distribution in chemical conditions could be imposed across the vault if such a distribution could be defined.

Spatially variable chemical conditions could result in regions in which higher concentrations of fissile material occur than have been calculated in this analysis. However, the current calculations using generic sorption parameters may bound the effects of precipitation.



4.6 GDF Post-closure Criticality FEPs and Scenarios – SF/HLW/HEU/Pu

The following sub-sections discuss the expected evolution of conditions in the SF/HLW/HEU/Pu disposal area of the GDF in higher-strength rock and identify post-closure criticality FEPs and scenarios.

4.6.1 GDF Evolution

The expected evolution of conditions in a SF/HLW disposal area is discussed in detail in the supporting technical report on SF/HLW/HEU/Pu disposal [2, §4.3.1]. Key factors are discussed here.

After emplacement, the bentonite buffer and tunnel backfill will begin to resaturate, causing the bentonite to swell [46, §3.8.1 and §A1.4.1]. A swelling pressure would develop as the buffer expands to seal the disposal canister in the deposition hole. On saturation, the buffer would provide a low-permeability barrier around the canister. Full saturation of the buffer will take years to decades depending on the hydrological characteristics of the fractures intersecting the deposition hole. The larger volumes of bentonite in the disposal tunnels would take decades to centuries to saturate. Hydraulic conditions in the near-field would slowly equilibrate with those in the surrounding host rock after saturation of the buffer and backfill.

If fluid flow through fractures intersecting deposition holes is sufficiently large, then the bentonite may be susceptible to piping and erosion [46, §3.10.4]. These processes have the potential to remove significant amounts of bentonite from a deposition hole and, hence, reduce the ability of the buffer to protect the canister. As far as possible, the intersection of features of such high hydraulic conductivity would be avoided, or if a deposition hole does intersect such a feature the deposition hole may not be used.

Chemical reactions between groundwater and bentonite minerals may occur, including mineral dissolution and precipitation and ion-exchange reactions. These reactions are expected to result in chemical buffering of the bentonite pore water to a near-neutral or slightly alkaline pH [46, §3.10.3 and §A1.4.1]. In the long term, the capacity for chemical buffering of the bentonite by reaction with groundwater would diminish as the reactants in the bentonite are exhausted and, depending on the chemical composition of the groundwaters, the composition of the buffer might be altered (e.g., Na-bentonite could be replaced by Ca-bentonite). Such alterations could affect the swelling pressure and hydraulic conductivity of the buffer.

The initial redox potential of the bentonite pore water might be oxidising, but the oxygen would be exhausted quickly (in tens to hundreds of years) as a result of redox reactions with dissolved reducing species and mineral impurities in the bentonite, and as a result of aerobic corrosion of the copper disposal canisters [46, §3.10.2]. Reducing conditions would be maintained by the ingress of reducing groundwater.

Development of a reducing redox potential in the vicinity of the disposal canisters would substantially reduce the corrosion rate of copper, leading to long canister lifetimes [52]. Aggressive species, such as sulphides and microbes, could accelerate corrosion, but the transport of these species to disposal canister surfaces will be limited by the low permeability



of the bentonite. Generally, the copper canisters are expected to maintain their integrity and provide full radionuclide containment for at least 100,000 years. Seismic events caused by earthquakes or glacial loading and unloading could contribute to canister failure as a result of shear along fractures that intersect canister deposition holes. However, the buffer fills the space between the copper canister and the host rock, and is designed inter alia to protect the canister and the wastes from the effects of deformation. Also, it would be expected that large faults and potentially significant fractures would, as far as possible, be avoided when locating deposition holes. In this way the potential for earthquakes to affect the disposal canisters would be managed. However, movement on unidentified and ‘secondary’ faults and fractures still has the potential to lead to canister failure under certain circumstances.

If a copper canister was to fail, pore waters would come into contact with the cast iron insert present inside the canister and would initiate dissolution of the wasteform. In general, the release rates of radionuclides such as uranium and plutonium would be restricted because of their low solubility under expected reducing conditions.

The transport of any uranium and plutonium released from the wasteform would be retarded by a combination of sorption on, and co-precipitation with, secondary phases formed by the degradation of waste package materials, as well as by sorption on surfaces provided by the bentonite buffer. Uranium and plutonium transport through the bentonite would occur primarily by diffusion [46, §A1.4.3] before eventual release into the host rock.

Degradation of bentonite (e.g., chemical destabilisation by contact with infiltrating waters or mechanical disruption under high groundwater flow rates) could cause the formation of bentonite colloids [46, §3.10.4]. Colloids could also be generated by the corrosion of copper and steel in the waste packages. Such colloids might have a high sorptive affinity for uranium and plutonium. In addition, degradation of the HLW and SF matrices has the potential to produce uranium and plutonium colloids [48, §4.3.1]. Such colloids could contribute to enhanced uranium and plutonium transport (via advective colloid transport) or uranium and plutonium retardation (if colloid filtration processes are significant). Provided the bentonite swelling pressure is maintained, the buffer would filter out and immobilise any uranium and plutonium colloids generated within waste containers.

After hundreds of thousands to millions of years, the engineered barriers would degrade and their capacity for radionuclide containment and retardation would diminish. Retardation of long-lived radionuclides would be provided by sorption and diffusion into the host rock matrix [46, §A1.4.3].


Criticality FEPs have been identified in terms of events and processes that could result in changes in reactivity after GDF closure (see Section 2.2). The criticality FEP analysis is described in detail in the technical report on SF/HLW/HEU/Pu disposal [2, Appendix D]. Similar to the approach taken for LLW/ILW/DNLEU disposal, the FEP analysis has been structured as follows:

• FEPs that could result in water entry into a disposal package (i.e., the occurrence of package failure mechanisms).

• FEPs that could result in changes in reactivity following water entry into the disposal package (i.e., degradation leading to relocation of fissile material, neutron absorbers, neutron reflectors and neutron moderators).



• FEPs that could result in the migration and accumulation of fissile material outside the package (i.e., accumulation by precipitation, sorption, filtration or gravitational settling).

• FEPs that could result in the migration and accumulation of fissile material from more than one disposal package.

Each of these criticality FEP groups is discussed in detail in the supporting technical report [2, Appendix B].

As for the LLW/ILW/DNLEU disposal analysis, criticality scenarios have been defined at a high level as:




The key stages in the development of these scenarios are illustrated in the event tree shown in Figure 4.22.



Figure 4.22: Event tree showing fissile material accumulation scenarios for SF/HLW/HEU/Pu disposal in higher-strength rock. The

yellow boxes indicate criticality scenarios involving fissile material from a single package and the blue boxes indicate scenarios involving fissile material from multiple packages.


* Radioactive decay of 239Pu to 235U may be significant for all scenarios. † Accumulation may occur as a result of sorption, precipitation, mineralisation, filtration or settling, according to the material properties and geochemical and hydrogeological conditions.



Water gradually enters the disposal package through a defect if one exists


The wasteform degrades (corrosion and dissolution reactions, including dissolution of Pu* and U)


When sufficiently weakened, the degraded wasteform slumps in the package




Particulate fissile material removed


Fissile material from multiple packages may accumulate in the buffer†




Fissile material migrates by advection and diffusion through the buffer and may accumulate†






4.7 GDF Post-closure Criticality Scenario Assessment – SF/HLW/HEU/Pu

This section summarises the analysis of the likelihood of criticality following the disposal of SF/HLW/HEU/Pu in higher-strength rocks, which is presented in detail in the project report on SF/HLW/HEU/Pu disposal [2, §4.4]. Consideration is given to GDF post-closure criticality scenarios that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package (Figure 4.22).

4.7.1 GoldSim Model

On the basis of the understanding of the expected evolution of conditions in a GDF for SF/HLW/HEU/Pu in higher-strength rock and the FEP analysis, conceptual models for post-closure criticality scenarios have been derived. The conceptual models have been implemented in GoldSim models for each waste type.

The model consists of a segment of a disposal tunnel (assumed to be of rectangular cross-section) above eight deposition holes (cylindrical), as shown in Figure 4.23. This number of waste packages has been chosen to allow consideration of the behaviour of multiple packages while ensuring that the modelling is computationally tractable. The waste packages are surrounded by a bentonite buffer in the deposition holes and the tunnel is assumed to be backfilled with crushed rock and bentonite. The eight waste packages in the tunnel segment contain the same wasteform (SF, HLW, HEU or Pu). The GoldSim model cells are indicated in Figure 4.23. There are three buffer components in each deposition hole: cylindrical components above and below the waste package and an annulus around the waste package. The tunnel comprises eight segments, with one segment above each deposition hole. Broadly, the gradual degradation of the waste packages is modelled and when the wasteform is calculated to be exposed to groundwater, the release of uranium and plutonium and their transfer through the barrier system is modelled.

The time until a wasteform is exposed to groundwater and begins to dissolve depends on the rate of degradation of the different barriers in the waste package. The copper canister represents the only barrier in AGR and PWR SF waste packages; the cast iron insert is assumed not to prevent groundwater access to the spent fuel once the copper canister has been breached. The HLW package has two barriers: the copper canister and the inner stainless steel WVP (Waste Vitrification Plant) canister. The HEU/Pu packages have four barriers: the copper canister, the stainless steel canister that contains the glass-encapsulated stainless steel waste cans, the encapsulation glass and the waste cans themselves.

Based on the expected rate of copper corrosion under disposal conditions, complete general corrosion of a 5-cm-thick copper canister would take well in excess of 106 years. However, in addition to package breach by general corrosion, it is assumed that the copper canisters could be breached as a result of external disruption, such as a seismic event, or a degradation mechanism that is more rapid than general corrosion. In such cases, an early breach time is assumed for the copper canister.



Figure 4.23: Components of the SF/HLW/HEU/Pu model for the higher-strength rock

disposal concept. The naming convention for GoldSim model cells is shown and these names are used in the descriptions of modelling results. Transport pathways for advection of uranium and plutonium from the failed “Upstream Package” are indicated by arrows.

It is assumed that as the cast iron and stainless steel degrade a fraction of the corrosion product (iron oxide) remains in the package. Other corrosion products and glass degradation products are assumed to be removed and replaced by water.

On exposure to groundwater at the time of failure, the various wasteforms are assumed to degrade according to a fractional degradation rate. The uranium, plutonium and other wasteform materials are released as the wasteform degrades.

At the time of waste package failure, it is assumed that the buffer has become sufficiently permeable that groundwater can flow through the buffer, waste packages and deposition tunnel at a specific discharge equal to that of the host rock. This is a cautious assumption because, for such a flow system to develop, the deposition hole would need to be close to a highly conductive fracture where the flow would be capable of eroding a substantial amount of bentonite; deposition hole siting would aim to avoid such fractures.

The direction of groundwater flow is assumed to be such that mobile uranium and plutonium are advected into the buffer annulus around the canister and upwards into the buffer above the canister, then into the deposition tunnel and along the tunnel to a downstream accumulation zone. The surfaces across which uranium and plutonium are advected from the “Upstream Package” are shown in Figure 4.23. Constraining advection to the deposition holes and tunnel is cautious in that it forces mixing and accumulation of fissile material from multiple waste packages. Diffusion can occur across all surfaces, including from the buffer and backfill cells into the host rock. The uranium and plutonium may sorb and precipitate along the transport pathway. The masses of radionuclides in each cell (and thus the



concentrations) are evaluated based on addition of the masses entering from upstream cells and subtraction of the masses transferring to downstream cells.

Whether calculated accumulations of fissile material in the buffer and backfill could result in criticality can be estimated by comparison of the accumulated mass of uranium or plutonium with the minimum mass required for criticality in different media based on assumptions about the configuration of the accumulation. The reactivity of backfill and tunnel materials containing fissile material that has migrated from degrading waste packages has been determined from previously documented criticality calculations [53]. Also, the reactivity of a waste package will evolve as the package and wasteform materials degrade, and plutonium and uranium decay and relocate. MCNP has been used to evaluate how the reactivity of a waste package changes based on assumptions about the location of the different materials remaining in an evolving waste package.

4.7.2 MCNP Model

To determine whether the evolving SF/HLW/HEU/Pu waste packages remain sub-critical requires evaluation of the neutron multiplication factor keff for the systems. Provided keff is less than 1.0, the system is sub-critical. Neutron transport calculations have been undertaken using the MCNP code to support assessments of whether the conditions within degrading waste packages remain sub-critical.

The MCNP application requires the configuration and properties of the waste package materials to be specified. The time-dependent volumes of the different materials in each type of waste package are calculated using the GoldSim model. These volumes are assumed to be distributed in different configurations depending on assumptions about the re-location of degrading waste package materials. Two material re-location configurations have been modelled:

• In the first configuration (the ‘segregated’ case shown in Figure 4.24), the waste package components have been assumed to be so degraded that the solid components (including 10% of the iron) have slumped to form a homogenous layer at the base of the package, with a layer of water overlying the slumped material. Uranium and plutonium are present in the slumped material (solid) and in the water layer (dissolved). The height of the slumped layer is derived from the waste volume calculated by GoldSim and the package geometry. The degraded system is surrounded by bentonite.

• In the second configuration (the ‘water mixed’ case shown in Figure 4.24), the material remaining in the waste package is assumed to be suspended uniformly in the water.

The two configurations represent the two extremes of possible material distributions in the waste packages, although it is acknowledged that the highest reactivity conditions may occur for a configuration between the segregated and fully mixed cases. In the early breach model, the copper shell of the disposal canister has been included. It is possible that the bentonite buffer would collapse and mix with the wasteform but, cautiously, the presence of bentonite in the mixture has been ignored. The neutron multiplication factor is then calculated at selected times for each configuration based on the GoldSim results at those times.



Figure 4.24: Geometry of degraded waste packages modelled using MCNP. In the

‘segregated’ model, layers of solid waste (grey) and water containing fissile material (blue) are modelled. In the ‘water mixed’ model a uniform mixture (blue) of waste and water is modelled. The copper canister (brown) is included in the early failure model. In each case, the canister is surrounded by bentonite (yellow).

4.7.3 Modelling Results

Calculations to evaluate the evolution and changing reactivity of HLW, SF, HEU and Pu waste packages are presented in the following sub-sections. The accumulation of fissile material in the buffer, accumulation zone and surrounding host rock is also discussed.

4.7.3.1 HLW HLW contains small amounts of fissile material. Calculations have confirmed that the reactivity of HLW packages remains small as the waste packages evolve. Most of the HLW uranium inventory is eventually transferred into the buffer material surrounding the degrading waste package, but such accumulations of 235U are too small to present a criticality concern and insignificant amounts of 235U are calculated to reach the tunnel or downstream accumulation zone in 108 years. Accumulation of a critical mass of 235U in the tunnel or host rock from multiple packages is not credible.

4.7.3.2 AGR SF Figure 4.25 shows calculated material volumes in an AGR spent fuel waste package for a typical single realisation in which container failure occurs by general corrosion. Note that the grey shaded arrows in these figures indicate the period beyond 106 years where there are large uncertainties associated with the effects of natural processes on disposal and the results are simply indicative of performance.

The volume of copper decreases as the container corrodes until failure occurs. By this time, the 239Pu in the spent fuel has decayed to 235U. After container failure, the cast iron corrodes relatively quickly, with a fraction of the iron assumed to remain in the container in the form of



iron corrosion products. Most of the uranium remains in solid form for the duration of the realisation, with dissolution being solubility limited. A small amount of dissolved uranium is advected out of the container over 108 years.

The neutron multiplication factor (keff) of the waste package has been calculated using MCNP as a function of material volume after container failure for the “segregated” and “water mixed” configurations. The water mixed configuration results in the highest value of keff of about 0.35 at the time of container failure. These results are not surprising given that the fuel is assumed to been subject to a high burn-up (see Section 3.5.2) such that the 235U content of the AGR SF is only about 0.8 wt% and such uranium enrichments will always be sub-critical under natural-water-moderated conditions.

Early canister failure has been assumed to occur between 200,000 and 300,000 years. Even at the earliest failure time, insignificant amounts of 239Pu will remain. The results of one realisation for which early package failure occurred are shown in Figure 4.26. The increase in keff as the iron corrodes and is removed from the waste package is apparent.

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Copper Cast Iron Spent Fuel Iron Degradation Products Segregated keff Water Mixed keff Figure 4.26: Volumes of package materials (left axis) and keff (right axis) for an AGR SF

package undergoing early failure.

Accumulation of fissile material outside the waste package could occur in the bentonite buffer. The maximum calculated 235U mass occurs in the buffer component surrounding the AGR SF waste package (Mid Buffer component as indicated in Figure 4.23), but is no more than about 1 kg 235U after 108 years (Figure 4.27). Such accumulations of uranium would not result in criticality even if the uranium derived from fresh AGR fuel at enrichments of a few weight percent 235U. A mass of at least 16 kg of 235U is required for criticality in bentonite for an optimum water-moderated spherical mass of UO2 at an enrichment of 3 wt% 235U [53, Figure 4.2]. Note that the releases prior to 1.25x107 years relate to the small number of realisations in which early canister failure has occurred.

The mass of 235U in the tunnel components and downstream accumulation zone is less than 0.1 kg in the simulation period. Such accumulations of uranium would not result in criticality.




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Figure 4.27: Mass of 235U in the buffer component surrounding the AGR SF waste package (Mid Buffer) in the higher-strength rock concept.

4.7.3.3 PWR SF Figure 4.28 shows the GoldSim calculated volumes for a single realisation in which PWR spent fuel container failure occurs by general corrosion. The MCNP calculations for the water mixed configuration results in the highest value of keff of about 0.5 at the time of container failure. Note that the PWR spent fuel considered in this analysis has been assumed to be subject to a high burn-up (see Section 3.5.2) such that it has an effective enrichment of about 1.2 wt% 235U and is unsurprisingly sub-critical.

The results of a realisation in which early failure occurs are shown in Figure 4.29. In this case, the maximum keff is about 0.55 for the water mixed configuration, reducing as the solid uranium slowly dissolves and is removed from the waste package.

The maximum calculated 235U mass in the buffer occurs in the component surrounding the PWR SF waste package (Mid Buffer component), but is at most little more than 1 kg 235U after 108 years (Figure 4.30). Such accumulations of uranium would not result in criticality even if the uranium derived from fresh PWR fuel at enrichments of a few weight percent 235U in UO2, because a mass of at least 16 kg of 235U is required for criticality in bentonite under optimum conditions for criticality [53, Figure 4.2]. The calculated mass of 235U in the tunnel components and downstream accumulation zone is less than 0.1 kg in the simulation period. Again, such accumulations of uranium would not result in criticality.




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Copper Cast Iron Spent Fuel Iron Degradation Products Segregated keff Water Mixed keff Figure 4.28: Volumes of package materials (left axis) and keff (right axis) for a PWR SF

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Copper Cast Iron Spent Fuel Iron Degradation Products Segregated keff Water Mixed keff Figure 4.29: Volumes of package materials (left axis) and keff (right axis) for a PWR SF






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Figure 4.30: Mass of 235U in the buffer component surrounding the PWR SF waste package (Mid Buffer) in the higher-strength rock concept.

4.7.3.4 HEU/Pu In the case of separated Pu waste packages, the 239Pu in the package will have almost entirely decayed to 235U by the time of earliest breach of the copper canister (200,000 years). Therefore, the results for Pu and HEU waste packages are virtually the same.

Figure 4.31 shows the calculated volumes for a single realisation in which container failure occurs by general corrosion. Once the ceramic wasteform is exposed to water, it begins to degrade, but most of the uranium remains in solid form for the duration of the realisation, with dissolution being solubility limited. A small amount of dissolved uranium is advected out of the container based on the assumed flow rates.

Cautiously, as the ceramic degrades, the poisons (hafnium and gadolinium) are assumed to be dissolved and removed in the flowing groundwater. Therefore, the calculated values of keff increase as the ceramic degrades, as shown in Figure 4.31. The water mixed configuration results in the highest value of keff of about 0.75 when the ceramic has fully degraded and all of the poisons have been removed. Thereafter, keff decreases as the uranium is gradually dissolved and removed from the waste package. The results of a realisation in which early failure occurs are shown in Figure 4.32. In this case, the maximum keff is about 0.9 when the ceramic has fully degraded for the water mixed configuration, reducing as the solid uranium slowly dissolves and is removed from the waste package.




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Stainless Steel Glass Ceramic Copper Cast Iron Iron Degradation Products Segregated keff Water Mixed keff

Figure 4.31: Volumes of package materials (left axis) and keff (right axis) for an HEU/Pu package undergoing general corrosion.

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Stainless Steel Glass Ceramic Copper Cast Iron Iron Degradation Products Segregated keff Water Mixed keff Figure 4.32: Volumes of package materials (left axis) and keff (right axis) for an HEU/Pu






The maximum calculated uranium mass in the buffer includes about 25 kg 235U (i.e., almost the entire 235U content of a waste package) and occurs after 108 years in the buffer component surrounding the HEU/Pu waste package (Mid Buffer), as shown in Figure 4.33. The uranium has an enrichment of about 30 wt% 235U (reduced from almost 100 wt% 235U by the 238U added to the wasteform to act as a diluent and neutron absorber). A mass of about 4 kg 235U (UO2 at 10 wt% 235U) would be required for criticality in bentonite under optimum conditions for criticality [53, Figure 4.2]. Therefore, it is possible that criticality could occur in the buffer on timescales in excess of 107 years.

The mass of 235U in the tunnel components and downstream accumulation zone is calculated to be less than about 3 kg in the simulation period (i.e., slightly less than the above-noted critical mass of 4 kg 235U), but the accumulation of a critical mass of 235U in the tunnel or accumulation zone may be possible on timescales in excess of 108 years.

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Figure 4.33: Mass of 235U in the buffer component surrounding the HEU/Pu waste package (Mid Buffer) in the higher-strength rock concept.

4.8 Summary of Assumptions – SF/HLW/HEU/Pu The likelihood of criticality following the disposal of SF/HLW/HEU/Pu in higher-strength rock has been analysed based on consideration of the expected evolution of conditions in the GDF. Post-closure criticality scenarios have been analysed using GoldSim probabilistic models. Details of the results and conclusions of the scenario analyses are presented in the main report on SF/HLW/HEU/Pu disposal [2]. A key factor in the analysis is that the copper disposal canisters are expected to maintain their integrity and provide full radionuclide




containment for at least 105 years. It can be expected that 239Pu will have largely decayed to 235U before canister failure. Therefore, the analysis of SF/HLW/HEU/Pu disposal in higher-strength rock has focused on the likelihood of criticality involving 235U.

The analysis has been based on many assumptions about the waste inventory, waste packaging concepts, disposal facility design, barrier system properties and GDF evolution. The main assumptions and alternative assumptions and their potential impacts on the likelihood of criticality analysis and results are discussed in Table 4.7.

Table 4.7: Key assumptions made in the likelihood of criticality analysis for SF/HLW/HEU/Pu disposal in a higher-strength host rock. Alternative assumptions and their potential impacts on the analysis are noted.



The Derived Inventory has been used, which is based on the Baseline Inventory estimate and detailed characterisation data from the 2007 UK RWI.

The 2013 UK RWI has been published and some waste packaging concepts have been revised since the production of the 2007 UK RWI.

The analysis is dependent on the longevity of the copper canister. Changes to waste packaging concepts could have a significant impact on the criticality scenario analysis. For example, a different container material could be selected, or large multi-purpose canisters could be used that contain more fissile material than the waste packages considered here.

The Reference Case of the Derived Inventory has been used.

Two additional inventory scenarios are available in the Derived Inventory that address uncertainties in forecasts of future waste arisings: the Lower and Upper Inventories.

The impact on the criticality scenario analysis of using the Lower and Upper Inventories is likely to be small if there are no changes to the waste packaging concepts. However, the impacts of spent fuels from new nuclear reactors (such as MOX fuel or fuels of increased initial enrichment or higher burn-up) would require consideration.

Relatively high fuel enrichments and burn-ups have been assumed but the effects of fission products have not been included.

The range of fuels to be disposed of will have been subject to different burn-ups. The impacts of uncertainties in the nuclide content of fuels could be evaluated.

Low burn-up fuels are likely to have a higher reactivity under conditions represented by the post-closure scenarios considered in this report. However, the inclusion of fission products may reduce calculated reactivity.

A ceramic wasteform and can-in-canister approach has been assumed for HEU and Pu disposal.

Alternative immobilisation matrices and packaging concepts may be used.

The rates of degradation and release of fissile nuclides from alternative immobilisation matrices and different containers may differ from those considered in this project, which would affect the potential for post-closure fissile material accumulation.





The host rock properties, disposal concept design and underpinning parameter values are consistent, as far as possible, with the 2010 DSSC.


The current assumptions about groundwater flow rates are considered to encompass conditions likely for a UK GDF. However, the impacts of assuming alternative barrier systems are uncertain.

Criticality scenario analysis

Canisters have been assumed to contain the wastes for on the order of 105 years.

Canisters could be assumed to fail at earlier times than considered here if judged necessary, based on new understandings of processes such as copper corrosion and copper creep.

Canister failure prior to significant 239Pu decay would result in higher calculated reactivities.

The scenario analysis has assumed that after container failure and cast iron corrosion, only 10% of iron corrosion products remain in the container.

Analysis of the behaviour of iron corrosion products under disposal conditions has indicated that iron would remain in solid form for long periods.

Alternative assumptions about materials involved in fissile material accumulations and their configurations could affect the criticality analysis significantly. The inclusion of iron would reduce the likelihood of criticality.

In some cases, judgments on the possibility of critical masses or concentrations of fissile nuclides occurring are based on comparisons with the conditions required for criticality in idealised systems (e.g., optimally water-moderated spheres of fissile material).

Alternative assumptions about the geometry of fissile material accumulations could be made.

Alternative assumptions about fissile material accumulations could affect the criticality analysis significantly in that greater masses of fissile material would be required for criticality.




Calculations of the neutron multiplication factor for in-package scenarios have involved specific assumptions about material behaviour (e.g., dissolution and removal of neutron poisons in groundwater).

Alternative assumptions about the geometry of fissile material accumulations and the presence of materials such as iron could be made. Analysis of the behaviour of iron corrosion products under disposal conditions has indicated that iron would remain in solid form for long periods under disposal conditions. Also, the presence of neutron poisons hafnium and gadolinium in the degraded/accumulated material could be considered for HEU and Pu wastes.

Alternative assumptions about materials involved in fissile material accumulations and their configurations could affect the criticality analysis significantly. The inclusion of greater amounts of iron and neutron poisons would reduce reactivity.

Concentrations of natural uranium in groundwater are insignificant.

If water entering the waste packages is saturated in natural uranium then the rate of dissolution of uranium from the wastes would be greatly reduced. The issue could be addressed when site-specific data on groundwater compositions are available.

Assuming groundwater to include concentrations of natural uranium would reduce the rate of release of uranium from waste packages and the enrichment of dissolved uranium in the GDF would be lower than calculated in this analysis.

The buffer and backfill have been assumed to be sufficiently degraded that advective mass transfer can occur through a failed disposal container and buffer and backfill.

Alternative flow distributions or a diffusion dominated system could be assumed.

The current approach is considered to be cautious in that there is greater potential for material from several waste packages to accumulate at a single location. However, the impacts of variations in flow rates and distributions are uncertain.

It has been assumed that the buffer would filter out and immobilise any uranium and plutonium colloids generated within waste containers.

Colloid-facilitated uranium and plutonium transport through the buffer could be considered.

Colloids may have the effect of increasing calculated uranium and plutonium masses in the buffer. However, uranium and plutonium are anyway strongly sorbed on the buffer and the presence of colloids may not increase calculated masses significantly.

Sorption sites do not become saturated.

Uranium and plutonium sorption site saturation could be included in the model.

If sorption sites become saturated it would reduce the likelihood of critical concentrations of plutonium and uranium being calculated.




Uniform chemical conditions have been assumed in the model. Therefore, although precipitation is represented, the conditions required for precipitation are not modelled.

A distribution in chemical conditions could be imposed across the model if such a distribution could be defined.

Spatially variable chemical conditions could result in regions in which higher concentrations of fissile material occur than have been calculated in this analysis. However, the current calculations using generic sorption parameters may bound the effects of precipitation.



5 Likelihood of Criticality in a GDF in Lower-strength Sedimentary Rock

5.1 Introduction This section summarises the analysis of the likelihood of criticality following the disposal of LLW/ILW/DNLEU and SF/HLW/HEU/Pu in lower-strength sedimentary rock. First, a description of the illustrative design for geological disposal of these wastes in lower-strength sedimentary rock is provided. Subsequently, LLW/ILW/DNLEU and SF/HLW/HEU/Pu are discussed in turn in terms of:



Full details of the analysis are presented in the detailed technical reports [1, 2].

5.2 Illustrative Disposal Concept for Lower-strength Sedimentary Rock

The illustrative design layout for a GDF in lower-strength sedimentary rock is shown in Figure 5.1 [17], which indicates the separate disposal areas for LLW/ILW and HLW/SF. A closure date of 2150 has been assumed for such a facility.

The UILW containers would be emplaced in vaults, with the containers stacked five high in arrays, as shown in Figure 5.2 [17, §8.2.1]. The vaults would be backfilled and sealed using a cementitious grout [17, §10.2]. Seals at each vault entrance would be provided by shield doors, and seals would be placed at the exit end of the vaults, and in the tunnels that provide access to each disposal module [17, §10.4]. It is assumed that the disposal concept for DNLEU would be the same as that presented for UILW [17, §3.1].

The HLW/SF disposal area consists of disposal tunnels designed for the horizontal deposition of individual carbon steel disposal canisters on top of a layer of bentonite blocks (see Figure 5.3). Following emplacement of each disposal canister (3 m apart), a mobile bentonite hopper would be used to place pre-compacted bentonite pellets in the disposal tunnel, thereby providing progressive backfilling of the disposal tunnel [17, §8.3]. Each HLW/SF disposal area tunnel would be sealed at one end with highly compacted bentonite and a concrete bulkhead; a shield door would provide a seal at the tunnel entrance [17, §10.4].



Figure 5.1: Illustrative underground layout for a GDF accommodating the Baseline Inventory in a lower-strength sedimentary host

rock [17, Figure 34].



Figure 5.2: Schematic cross-section through a UILW vault in a lower-strength sedimentary

host rock [17, Figure 22]. The stacking arrangement for 3 m3 boxes is shown.

Figure 5.3: Schematic of a HLW/SF disposal tunnel in lower-strength sedimentary rock

[17].




As waste packages and GDF barrier materials begin to degrade after emplacement, fissile and other materials may be mobilised, which could affect reactivity and the likelihood of criticality. Although anticipated changes in the arrangement of waste package materials are expected to reduce disposal system reactivity through dispersion of fissile material, the likelihood of higher reactivity systems developing requires assessment. The following sub-sections discuss the expected evolution of conditions in the LLW/ILW/DNLEU disposal area of a GDF in lower-strength sedimentary rock and identify post-closure criticality FEPs and scenarios.

5.3.1 GDF Evolution


After sealing, the ILW disposal vaults would start to resaturate, but the rate of resaturation would be low because of the low permeability of the host rock. Oxygen in the vaults would be consumed by corrosion and other processes, such as microbial activity. Alkaline conditions would develop as the incoming groundwater reacts with the cementitious backfill [46, §A2.5.1]. Complete resaturation of the disposal vaults would be expected to take about 500 years, although this is highly dependent on the hydraulic conductivity of the rock [46, §A2.5.2].

High pH conditions that limit solubility and promote sorption and precipitation of uranium and plutonium would develop in the disposal vaults as the backfill reacts with the incoming groundwater. The wasteform is an important barrier to the release of radionuclides during the resaturation period while the chemical barrier provided by the cementitious backfill is becoming established.

As the engineered barriers slowly degrade, mobile radionuclides would be released as dissolved species from the ILW containers and would start to migrate through the engineered barrier system. Groundwater movement is expected to be sufficiently slow that mass transfer will be dominated by diffusion.

Many tens to hundreds of thousands of years after closure, the engineered barriers in the ILW disposal area would begin to lose their effectiveness and would no longer be able to perform their intended safety functions fully. With time, the wasteforms and the cementitious backfill would alter through reaction with the groundwater and may be disrupted as a result of rock creep and associated compaction. Eventually the pH in the ILW disposal area would fall to a level where the chemical barrier provided by the cementitious backfill loses some of its effectiveness. Uranium and plutonium may be released to the host rock in this period.

In the very long term (many hundreds of thousands to millions of years), the engineered barrier system is not expected to provide complete containment. However, the degraded engineered barriers are likely to continue to retard the release of uranium to some degree. By this time the plutonium inventory would have decayed substantially.




Based on the above understanding of the expected evolution of conditions in a GDF in a lower-strength sedimentary host rock, criticality FEPs have been identified in terms of events and processes that could result in increases in reactivity after GDF closure. The criticality FEP analysis is described in detail in the supporting technical report on LLW/ILW/DNLEU disposal [1, Appendix C] and has been structured as for the higher-strength host rock analysis described in Section 4.3.2. Again, the criticality FEP groups are discussed in detail in the supporting technical report [1, Appendix C] in terms of mechanical, chemical, thermal, gas-related, hydrological, radiological, and microbiological events and processes.

For post-closure criticality to occur, package failure followed by substantial degradation and relocation of wasteform materials would be required. Relocation of fissile and other materials to form a critical configuration could be envisaged as occurring in a disposal package or at some location outside the disposal package and, for the latter, could involve mixing and accumulation of fissile material from several disposal packages. These considerations have formed the basis of criticality scenario construction based on sequences and combinations of criticality FEPs. The criticality scenarios have been defined at a high level as:




The key stages in the development of these scenarios are illustrated in the event tree shown in Figure 5.4.



Figure 5.4: Event tree showing fissile material accumulation scenarios for LLW/ILW/DNLEU disposal in lower-strength sedimentary

rock. The yellow boxes indicate criticality scenarios involving fissile material from a single package and the blue boxes indicate scenarios involving fissile material from multiple packages.





Water gradually enters the disposal package through a vent or defect and saturates the waste


The bulk wasteform degrades (corrosion and dissolution reactions, including dissolution of Pu* and U)


When sufficiently weakened, the degraded wasteform slumps in the package and creep closure may result in compaction




Particulate fissile material removed (if flow sufficient)

Fissile material migrates by advection and diffusion through the host rock and may accumulate† (diffusion may dominate)

When sufficiently weakened, wastes in a stack of degrading packages progressively slump towards the vault floor and may be compacted




Fissile material migrates by advection and diffusion through the system seals and may accumulate† (diffusion may dominate)

Fissile material migrates by advection and diffusion through the backfill and may accumulate† (diffusion may dominate)


* Radioactive decay of 239Pu to 235U may be significant for all scenarios † Accumulation may occur as a result of sorption, precipitation, filtration or settling, according to the material properties and geochemical and hydrogeological conditions.




This section considers the likelihood of nuclear criticality following the disposal of ILW and DNLEU in lower-strength sedimentary rocks, which is presented in detail in the project report on LLW/ILW/DNLEU disposal [1, §5.4]. Consideration is given to GDF post-closure criticality scenarios that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package (Figure 5.4).

5.4.1 Reactivity Increase in a Waste Package

Based on the assumption that there is no significant component of groundwater flow, the potential for material to be removed from waste packages in sufficient quantities for voids to be created and slumping to occur is small. It is likely that diffusive processes will result in the generation of a uniform mineralogy in the NRVB and encapsulation grouts and material transfer will be limited.

Host rock creep could cause the gradual compaction of the saturated backfill and degrading waste packages. Compaction will cause reduction in pore space and expulsion of water from waste packages. However, given that dissolution and removal of materials from waste packages is expected to occur slowly, the rate of compaction of the wasteform will be slow.

The concentration of fissile material within the package will be increased by compaction. However, waste packages that meet the package-scale screening levels, as described in Section 3.2.2, would remain sub-critical following compaction. This is because the analysis to derive the post-closure screening levels assumed optimum concentrations of fissile material in water, with no credit taken for the diluting or neutron-absorbing effects of other wasteform materials. Compaction could not result in higher reactivity systems than assumed in the screening level calculations.

The maximum uranium concentration in a package has been estimated by considering an instantaneous point source of uranium and assuming isotropic diffusion in three dimensions in the surrounding medium. Based on this analysis, the concentration of uranium in a failed waste package reduces by an order of magnitude in about 500 years and two orders of magnitude in about 2000 years. Criticality inside the waste package will only be of concern in a diffusing system if reconfiguration of material in the waste package occurs before significant fissile material is removed from the waste package. However, based on the point source model and isotropic diffusion, the uranium concentration in a waste package would reduce by several orders of magnitude before slumping could occur (if it occurs at all), which implies that the uranium concentration would be much smaller than the minimum required for criticality following slumping.

5.4.2 Accumulation outside a Waste Package

The scenario for accumulation of fissile material outside a single waste package for a disposal facility in lower-strength sedimentary rock is similar to that considered for a disposal facility in higher-strength rock (see Section 4.4.3), except that mass transfer is dominated by



diffusion. No specific preferential accumulation site has been included in the scenario analysis.

The minimum concentration of 235U required for criticality in NRVB or a clay host rock is of the order 10 kg/m3 at any enrichment [49, Figures 4.1 and 4.6]. Based on the point source diffusion model results, the uranium concentration in a waste package, and therefore the uranium concentration in material surrounding the waste package, would reduce to less than 10 kg/m3 in a few hundred years. For the range of diffusion coefficients expected for uranium transfer in grout, NRVB and lower-strength sedimentary rock (10-11 to 10-10 m2/s), the diffusion length after 500 years is no more than about 3 m. Therefore, significant concentrations could not develop far from the uranium source by diffusion. Also, the point source model is cautious and, in reality, the uranium would be released at lower concentrations over a longer period than implied by instantaneous release. Therefore, criticality following diffusion of uranium from a waste package into the surrounding backfill and host rock would not occur, even for the highest loaded waste packages.

5.4.3 Accumulation from Multiple Waste Packages

In diffusion dominated systems, uranium concentrations in excess of those occurring in the waste packages cannot be achieved, as discussed above. As the system evolves and multiple waste packages fail and release uranium, the uranium concentration will tend to reduce to an average for the inventory as a whole, limiting the possibility of criticality to those areas where the uranium concentration is high initially (i.e., in high loaded waste packages).

5.5 Summary of Assumptions – LLW/ILW/DNLEU The likelihood of criticality following the disposal of LLW/ILW and DNLEU in lower-strength sedimentary rock has been analysed based on consideration of the expected evolution of conditions in the GDF. Post-closure criticality scenarios have been analysed using simple diffusion models. Details of the results and conclusions of the scenario analyses are presented in the main report on LLW/ILW/DNLEU disposal [1]. The key assumption made in the analysis is that mass transfer is diffusion dominated (see Table 5.1). Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (see Table 4.6).



Table 5.1: Key assumptions made in the likelihood of criticality analysis for LLW/ILW/DNLEU disposal in a lower-strength sedimentary host rock. Alternative assumptions and their potential impacts on the analysis are noted. Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.6).



Groundwater flow is negligible such that mass transfer is dominated by diffusion.


The analysis undertaken for the higher-strength rock disposal concept is considered to be bounding of any situation in which there could be a significant advective component of mass transfer in a lower-strength sedimentary rock.


The following sub-sections discuss the expected evolution of conditions in the SF/HLW/HEU/Pu disposal area of a GDF in lower-strength sedimentary rock and identify post-closure criticality FEPs and scenarios.

5.6.1 GDF Evolution

The expected evolution of conditions in a SF/HLW disposal area is discussed in detail in the supporting technical report on SF/HLW/HEU/Pu disposal [2, §5.3.1]. Key factors are discussed here.

As water enters the HLW/SF tunnels and deposition holes, the bentonite buffer around the disposal canisters would swell. The disposal tunnels would be sealed such that swelling of the bentonite is resisted as voids between the disposal canisters and the host rock are filled and the required mechanical and hydraulic properties of the bentonite are achieved. The swelling would increase buffer density and mechanical strength, allowing the buffer to resist the pressure caused by creep of the surrounding host rock, and the permeability of the saturated bentonite would be low, such that diffusion is the dominant solute transport mechanism [46, §A2.4.1]. Complete resaturation of the disposal system would take on the order hundreds of years because of the expected low hydraulic conductivity of the host rock [46, §A2.4.2].

Infiltrating pore water would interact with the bentonite buffer resulting in minor changes to the physical and chemical properties of the buffer. The pore waters would generally have a neutral or slightly alkaline pH provided by the buffering capacity of the bentonite and minerals in the host rock [46, §A2.4.2].

The carbon steel waste packages would begin to corrode after emplacement. Initially, the corrosion rate would be relatively high under aerobic conditions, but the corrosion reaction



would consume oxygen. The buffer pore waters would quickly become reducing as a result of such corrosion reactions and as a result of reactions with redox-active minor mineral constituents of the bentonite and host rock and the infiltration of reducing groundwater. The corrosion of carbon steel occurs at a much slower rate under anaerobic conditions and the disposal packages are expected to provide complete radionuclide containment for at least 10,000 years, or at least 1000 years under worse-case conditions [46, §A2.3.1 and §A2.4.3].

Once the carbon steel waste packages have been breached, pore waters would begin to leach radionuclides from the wasteforms. The release rates of uranium and plutonium will be restricted because of their low solubility under expected reducing conditions. The transport of uranium and plutonium after release from the wasteform would be retarded by sorption on and co-precipitation with the products of steel corrosion and sorption on the clay minerals in the bentonite buffer. Migration of uranium and plutonium through the low-permeability bentonite and host rock would be diffusion-dominated.

Colloids could be generated by corrosion and degradation of the waste packages and wasteform, which could facilitate uranium and plutonium transport [48, §4.3.1]. However, the bentonite buffer would be expected to filter colloids and prevent them from moving away from the site of their generation. Similarly, the host rock would act to filter colloids.


The criticality FEP analysis is described in detail in the supporting technical report on SF/HLW/HEU/Pu disposal [2, Appendix D] and has been structured as for the higher-strength host rock analysis described in Section 4.6.2. Again, the criticality FEP groups are discussed in detail in the supporting technical report [2, Appendix D] in terms of mechanical, chemical, thermal, gas-related, hydrological, radiological, and microbiological events and processes.

As for the LLW/ILW/DNLEU disposal analysis, criticality scenarios have been defined at a high level as:




The key stages in the development of these scenarios are illustrated in the event tree shown in Figure 5.5. Note that the figure includes representation of advection, but a diffusion-dominated system has been assumed in the scenario analysis. Note also that, as in the analysis of disposal in a higher-strength host rock, uniform chemical conditions have been assumed in the scenario modelling. Therefore, although the precipitation process is represented in the models, the conditions required for precipitation are not modelled. However, it is expected that the calculations using generic sorption parameters would bound the effects of precipitation.



Figure 5.5: Event tree showing fissile material accumulation scenarios for SF/HLW/HEU/Pu disposal in lower-strength sedimentary

rock. The yellow boxes indicate criticality scenarios involving fissile material from a single package and the blue boxes indicate scenarios involving fissile material from multiple packages.


* Radioactive decay of 239Pu to 235U may be significant for all scenarios † Accumulation may occur as a result of sorption, precipitation, filtration or settling, according to the material properties and geochemical and hydrogeological conditions.



Water gradually enters the disposal package through a weld defect

Mechanical damage (e.g. creep) or container corrosion creates flow paths through the package

The wasteform degrades (corrosion and dissolution reactions, including dissolution of Pu* and U)


When sufficiently weakened, the degraded wasteform slumps in the package and creep may result in compaction




Particulate fissile material removed (if flow sufficient)


(diffusion may dominate)












This section considers the likelihood of criticality following the disposal of SF/HLW/HEU/Pu in lower-strength sedimentary rocks, which is presented in detail in the project report on SF/HLW/HEU/Pu disposal [2, §5.4]. Consideration is given to GDF post-closure criticality scenarios that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package (Figure 5.5).

5.7.1 GoldSim Model

On the basis of the understanding of the expected evolution of conditions in a GDF for SF/HLW/HEU/Pu in lower-strength sedimentary rock and the FEP analysis, conceptual models for post-closure criticality scenarios have been derived. The conceptual models have been implemented in GoldSim models for each waste type.

The model consists of a segment of a horizontal disposal tunnel (cylindrical in cross-section) containing six waste packages surrounded by bentonite. This number of waste packages has been chosen to allow consideration of the behaviour of multiple packages while ensuring that the modelling is computationally tractable. The six waste packages in the tunnel segment contain the same wasteform (SF, HLW, HEU or Pu). The model layout showing three waste packages is indicated in Figure 5.6. The GoldSim model cells are also indicated in Figure 5.6. There are three buffer components around each waste package: cylindrical components to the left and right of the waste package and an annulus around the waste package. Broadly, the gradual degradation of the waste packages is modelled and when the wasteform is calculated to be exposed to groundwater, the release of uranium and plutonium and their diffusive transfer through the bentonite and into the host rock is modelled.

The modelling approach is similar to that described for a GDF in higher-strength rock. Again, the time until a wasteform is exposed to groundwater and begins to dissolve (defined as the package failure time) depends on the rate of degradation of the different barriers in the waste package. The carbon steel canister represents the only barrier in AGR and PWR SF waste packages. The HLW package has two barriers: the carbon steel canister and the inner stainless steel WVP canister. The HEU/Pu packages have four barriers: the carbon steel canister, the stainless steel canister that contains the glass-encapsulated stainless steel waste cans, the encapsulation glass and the waste cans.

On exposure to groundwater at the time of time of failure, the various wasteforms are assumed to degrade according to a fractional degradation rate. After release from the waste package, uranium and plutonium diffusion can occur across all surfaces represented in the model (the surfaces of the Goldsim model cells shown in Figure 5.6). A coarse model grid has been assumed for diffusion, which tends to result in overestimates of the rate of diffusion between cells.



Figure 5.6: Components of the SF/HLW/HEU/PU model for the lower-strength sedimentary

rock disposal concept. The naming convention for GoldSim model cells is shown and these names are used in the descriptions of modelling results. Diffusion can occur across each model cell boundary.

The uranium and plutonium may sorb and precipitate in the buffer and host rock. The masses of radionuclides in each cell (and thus the concentrations) are evaluated by aggregating the masses entering from, or leaving to, surrounding cells. If the radionuclide concentration exceeds the solubility limit, then a precipitated mass is calculated. The precipitated mass is dissolved when the radionuclide concentration falls below the solubility limit. Similar to the analysis for a GDF in higher-strength rock, uniform chemical conditions have been assumed. Therefore, the conditions required for precipitation are not realised, but the current calculations using generic sorption parameters are considered to bound the effects of precipitation. Also, concentrations of natural uranium in groundwater have again been assumed to be insignificant. If the groundwater is saturated in natural uranium then the rate of dissolution of uranium in the waste packages would be greatly reduced.

All of the uranium and plutonium that enters the host rock is assumed to accumulate in a single model cell (an accumulation zone). The accumulation zone could be interpreted as a region hydraulically connected to a fracture network in the host rock disturbed zone. This represents a cautious modelling approach in that it is difficult to envisage a situation in which all of the uranium and plutonium migrating from a row of waste packages could be channelled towards a single location where geochemical conditions are such that an irreversible accumulation process, such as precipitation, occurs.

Whether calculated accumulations of fissile material in the backfill or host rock could result in criticality can be estimated by comparison of the accumulated mass of uranium or plutonium with the minimum mass required for criticality in different media based on assumptions about the configuration of the accumulation (e.g., idealised water-moderated, water-reflected spherical configurations in bentonite [53]). Also, the reactivity of a waste package will evolve as the package and wasteform materials degrade, and plutonium and uranium decay and relocate. MCNP has been used to evaluate how the reactivity of a waste package changes based on assumptions about the location of the different materials remaining in an evolving waste package.



5.7.2 MCNP Model

To determine whether the evolving SF/HLW/HEU/Pu waste packages remain sub-critical requires evaluation of the neutron multiplication factor keff for the systems. The neutron transport calculations have been undertaken using the MCNP code.

The neutron multiplication factor has been evaluated for two degrading SF/HLW/HEU/Pu package configurations (the segregated case and the water mixed case as shown in Figure 5.7). The time-dependent volumes of the different materials in each type of waste package are calculated using the GoldSim model. These two cases represent the two extremes of possible material configurations in the waste packages, although it is acknowledged that the highest reactivity conditions may occur for a configuration between the segregated and fully mixed cases. It is possible that the bentonite buffer would collapse and mix with the wasteform but, cautiously, the presence of bentonite in the mixture has been ignored. The neutron multiplication factor is then calculated at selected times for each configuration based on the GoldSim results at those times.

Figure 5.7: Geometry of degraded waste packages modelled using MCNP. In (a) the

‘segregated’ model, layers of solid waste (grey) and water containing fissile material (blue) are modelled. In (b) the ‘water mixed’ model, a uniform mixture (blue) of waste and water is modelled. In each case, the canister is surrounded by bentonite (yellow).

5.7.3 Modelling Results

Calculations to evaluate the evolution and changing reactivity of HLW, SF, HEU and Pu waste packages are presented in the following sub-sections. The accumulation of fissile material in the tunnel buffer and in the surrounding host rock is also discussed.

(a)

(b)



5.7.3.1 HLW HLW contains small amounts of fissile material. Calculations have confirmed that the reactivity of HLW packages remains small as the waste packages evolve. Most of the HLW uranium inventory is eventually transferred into the buffer material surrounding the degrading waste package, but such accumulations of 235U are too small to present a criticality concern and insignificant amounts of 235U are calculated to reach the tunnel or downstream accumulation zone in 108 years. Accumulation of a critical mass of 235U in the tunnel or host rock from multiple packages is not possible.

5.7.3.2 AGR SF Figure 5.8 shows the calculated material volumes for a typical single realisation in which container failure occurs by general corrosion. The volume of carbon steel decreases as the container corrodes until failure occurs; 10% of the iron in the steel is assumed to remain in the container in the form of iron corrosion products. Most of the spent fuel remains in solid form for the duration of the realisation, with dissolution being solubility limited. A small amount of dissolved uranium diffuses out of the container. The water mixed configuration results in the highest value of keff of almost 0.3 at the time of container failure. As the wasteform degrades and the uranium is transported into the buffer, the reactivity of the waste package decreases.

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Carbon Steel Spent Fuel Iron Degradation Products Segregated keff Water mixed keff Figure 5.8: Volumes of package materials and keff for an AGR SF package undergoing

general corrosion in the lower-strength sedimentary rock concept.




Accumulation of fissile material outside the waste package could occur in the bentonite. The highest calculated 235U mass occurs in the buffer component surrounding the AGR SF waste package (Mid Buffer component), but is no more than about 7 kg 235U after 108 years (Figure 5.9). Such accumulations of uranium would not result in criticality even if the uranium derived from fresh AGR fuel at enrichments of a few weight percent 235U. A mass of at least 16 kg of 235U is required for criticality in bentonite for an optimum water-moderated spherical mass of UO2 at an enrichment of 3 wt% 235U [53, Figure 4.2].

The maximum accumulated mass in the host rock is 3 kg 235U at the end of the simulation period. Such accumulations of uranium would not result in criticality. About 600 kg 235U in UO2 at 3 wt% 235U would be required for criticality in a typical sedimentary rock [53, Figure 4.4].

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Figure 5.9: Mass of 235U in the buffer around the AGR SF waste package (Mid Buffer) in the lower-strength sedimentary rock concept.

5.7.3.3 PWR SF Figure 5.10 shows the calculated volumes for a typical single realisation in which container failure occurs by general corrosion. The volume of carbon steel decreases as the container corrodes until failure occurs. Most of the spent fuel remains in solid form for the duration of the realisation; that is, although the spent fuel degrades at a prescribed rate, it remains as a solid according to the solubility limit. A small amount of dissolved uranium is diffused out of the container. The water mixed configuration results in the highest value of keff of about 0.4 after container failure. As the wasteform degrades and the uranium is transported into the buffer, the reactivity of the waste package decreases.




The maximum calculated mass of 20 kg 235U occurs in the buffer component surrounding the PWR SF waste package (Mid Buffer), as shown in Figure 5.11. Such an accumulation of uranium would not result in criticality for PWR SF at typical fuel burn-ups. However, if the uranium derived from fresh PWR fuel at enrichments of a few weight percent 235U, the calculated mass in the buffer is of the same order of magnitude as the minimum required for criticality (i.e., 16 kg of 235U [53, Figure 4.2] for UO2 in bentonite at an enrichment of 3 wt% 235U), although this would require the uranium to accumulate in an optimum water-moderated spherical configuration. The maximum accumulated mass in the host rock is 8 kg 235U, which occurs at the end of the simulation period. About 600 kg 235U in uranium at 3 wt% 235U would be required for criticality in a typical sedimentary rock [53, Figure 4.4] and therefore criticality in the host rock is not credible.

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Carbon Steel Spent Fuel Iron Degradation Products Segregated keff Water mixed keff Figure 5.10: Volumes of package materials (left axis) and keff (right axis) for a PWR SF

package undergoing general corrosion in the lower-strength sedimentary rock concept.




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Figure 5.11: Mass of 235U in the buffer around the PWR SF waste package (Mid Buffer) in the lower-strength sedimentary rock concept.

5.7.3.4 HEU/Pu Figure 5.12 shows the calculated volumes for a single realisation of an HEU waste package in which container failure occurs by general corrosion. The volume of carbon steel decreases as the container corrodes until failure occurs and 10% of the iron is assumed to remain in the container in the form of iron corrosion products. Once the ceramic wasteform is exposed to water, it begins to degrade, but most of the uranium remains in solid form for the duration of the realisation, with dissolution being solubility limited. A small amount of dissolved uranium diffuses out of the container.

Cautiously, as the ceramic degrades, the poisons (hafnium and gadolinium) are assumed to be dissolved and removed in the groundwater. Therefore, the calculated values of keff increase as the ceramic degrades, as shown in Figure 5.12. The water mixed configuration results in the highest value of keff of about 0.7 when the ceramic has fully degraded and all of the poisons have been removed. Thereafter, keff decreases as the uranium is gradually dissolved and removed from the waste package.

The highest calculated uranium mass includes about 25 kg 235U and occurs after about 5x106 years in the buffer component surrounding the HEU/Pu waste package (Mid Buffer), as shown in Figure 5.13. The uranium has an enrichment of about 30 wt% 235U (reduced from almost 100 wt% 235U by the 238U added to the wasteform to act as a diluent and neutron absorber). A mass of about 4 kg 235U (UO2 at 10 wt% 235U) would be required for criticality in bentonite under optimum conditions for criticality [53, Figure 4.2]. Therefore, it is possible that criticality could occur in the buffer on timescales in excess of 106 years.




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Carbon Steel Stainless Steel Glass Ceramic Iron Degradation Products Segregated keff Water Mixed keff Figure 5.12: Volumes of package materials (left axis) and keff (right axis) for an HEU package

undergoing general corrosion in the lower-strength sedimentary rock concept.

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Figure 5.13: Mass of 235U in the buffer around the HEU/PU waste package (Mid Buffer) in the lower-strength sedimentary rock concept.





The maximum accumulated mass in the host rock is about 30 kg 235U at the end of the simulation period. About 5 kg 235U (100 wt% 235U) would be required for criticality in a typical sedimentary rock [53, Figure 4.4]. Therefore a critical mass could be generated in the host rock if the uranium accumulated in an optimum configuration for criticality.

5.8 Summary of Assumptions – SF/HLW/HEU/Pu The likelihood of criticality following the disposal of SF/HLW/HEU/Pu in lower-strength sedimentary rock has been analysed based on consideration of the expected evolution of conditions in the GDF. Wasteform degradation rates are expected to be slow and mass transfer is expected to be diffusion-dominated and, as such, it can be expected that 239Pu will have largely decayed to 235U before any significant fissile material accumulation can occur. Post-closure criticality scenarios have been analysed using GoldSim probabilistic models. Details of the results and conclusions of the scenario analyses are presented in the main report on SF/HLW/HEU/Pu disposal [2].

The key assumption made in the analysis is that mass transfer is diffusion dominated (see Table 5.2). Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (see Table 4.7).

Table 5.2: Key assumptions made in the likelihood of criticality analysis for SF/HLW/HEU/Pu disposal in a lower-strength sedimentary host rock. Alternative assumptions and their potential impacts on the analysis are noted. Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.7).



A carbon steel disposal canister has been assumed based on the packaging concepts consistent with the 2007 UK RWI.

The 2013 UK RWI has been published and some waste packaging concepts have been revised since the production of the 2007 UK RWI. Alternative disposal canister concepts have been considered.

The analysis is dependent on the corrosion allowance of the carbon steel canister. Changes to waste packaging concepts could have a significant impact on the criticality scenario analysis. For example, the use of a thinner-walled canister that includes a large void space could affect the analysis, potentially resulting in higher calculated reactivity.


Groundwater flow is negligible such that mass transfer is dominated by diffusion.


The analysis undertaken for the higher-strength rock disposal concept is considered to be bounding of any situation in which there could be a significant advective component of mass transfer in a lower-strength sedimentary rock.



6 Likelihood of Criticality in a GDF in Evaporite

6.1 Introduction This section presents the analysis of the likelihood of criticality following the disposal of LLW/ILW/DNLEU and SF/HLW/HEU/Pu in evaporite. First, a description of the illustrative design for geological disposal of these wastes in evaporite is provided. Subsequently, LLW/ILW/DNLEU and SF/HLW/HEU/Pu are discussed in turn in terms of:



Details of the analysis are presented in the supporting technical reports [1, 2].

6.2 Illustrative Disposal Concept for Evaporite The illustrative design layout for a GDF in evaporite is shown in Figure 6.1 [17], which indicates the separate disposal areas for LLW/ILW and HLW/SF. A closure date of 2150 has been assumed for such a facility.

In the illustrative concept for LLW/ILW disposal in salt, the waste packages are stacked three high in arrays in disposal rooms (see Figure 6.2), with separate disposal modules for SILW and UILW packages. The DNLEU packages would be disposed of in separate UILW vaults [17, §8.4]. It is assumed that no backfilling would be required within the disposal vaults, as natural creep would close the excavations over time. In the case of disposal in thinner salt formations, crushed salt may be used to minimise the local perturbations resulting from such creep. However, chemical conditioning (buffering) is envisaged to limit actinide solubilities and container corrosion rates and to promote sorption of selected radionuclides. This buffering would be achieved by placing bags of magnesium oxide (MgO) on top of each column of waste packages [17, §10.2]. The MgO bags are designed to burst as a result of creep closure and swelling of the contents, and to fill the void space surrounding the waste packages. UILW vault access tunnels would be backfilled with crushed rock salt and sealed by construction of a rigid concrete wall [17, §10.4].

The HLW/SF disposal area consists of tunnels designed for horizontal emplacement of individual disposal canisters (3 m apart) on the tunnel floor (see Figure 6.3). The area around the disposal canister would be filled with crushed rock salt immediately after emplacement of each disposal canister using a mobile hopper [17, §8.3]. Once all of the disposal canisters have been placed within a disposal tunnel, a tunnel seal would be constructed [17, §8.3].



Figure 6.1: Illustrative underground layout for a GDF accommodating the Baseline Inventory in an evaporite host rock [17, Figure 35].



Figure 6.2: Schematic cross-section through a UILW vault in an evaporite [17, Figure 24].

The stacking arrangement for 500 litre drums in stillages is shown.

Figure 6.3: Schematic of a HLW/SF disposal tunnel in evaporite rock [17].




The following sub-sections discuss the expected evolution of conditions in the LLW/ILW/DNLEU disposal area of a GDF in evaporite and identify post-closure criticality FEPs and scenarios.

6.3.1 GDF Evolution


In an undisturbed evaporite, groundwater transport occurs only through the rock matrix (since fractures are largely absent, unless significant bedding is present) and flow rates are typically extremely low or even non-existent [46, §A3.5 and §3.8.3]. Groundwater may be present, typically as brines, and the system may be saturated. However, the groundwater or brine is likely to be held within isolated pores, so there would be little mobile water in any one location. In the absence of connected porosity in an undisturbed evaporite, there would be no pathways for radionuclide migration. In the early post-closure phase, dry near-field evolution would be ensured by shaft and access-tunnel seals that prevent water entry from overlying formations and waste package corrosion would be limited.

A key characteristic of many evaporites, particularly rock salt and anhydrite, is the ability of the rock to creep, i.e. to flow plastically at relatively fast rates [46, §3.9]. Following an excavation, an evaporite would deform to fill the void created and re-establish isostatic conditions. As a result of this characteristic, an evaporite host rock has the potential to creep and to seal the openings around and within a GDF and, over time, to provide a complete, uninterrupted barrier around a GDF. This self-sealing behaviour means that any pathways that might enable more rapid migration of fluid and hence, radionuclide transport away from the engineered barrier system, are typically short-lived. Depending on the prevailing conditions (temperature, stress, rheology etc.) creep would be expected to render the permeability of disturbed evaporite rocks similar to that of the undisturbed rock within approximately 200 years [46, §A3.5.2]. Consequently, evaporite host rocks are able to provide a high degree of containment over the very long term.

After several thousands of years, the engineered barriers are likely to be significantly degraded [46, §A3.5.3]. Multiple breached waste packages are expected due to compaction and as a result of various chemical degradation processes. However, by the time significant waste package degradation has occurred and other engineered barriers have also failed, complete containment of the waste in a GDF would be provided by the host rock.


Based on the above understanding of the expected evolution of conditions in a GDF in evaporite, criticality FEPs have been identified in terms of events and processes that could result in increases in reactivity after GDF closure. The criticality FEP analysis is described in detail in the supporting technical report on LLW/ILW/DNLEU disposal [1, Appendix C.4] and has been structured as for the higher-strength host rock analysis described in Section 4.3.2. Again, the criticality FEP groups are discussed in detail in the supporting technical report [1,



Appendix C.4] in terms of mechanical, chemical, thermal, gas-related, hydrological, radiological, and microbiological events and processes.

For post-closure criticality to occur, package failure followed by substantial degradation and relocation of wasteform materials would be required. However, relocation of fissile and other materials to form a critical configuration cannot be envisaged in the assumed dry GDF. Therefore, the only criticality scenario that has been defined is one involving fissile material accumulation inside a waste package.


This section considers the likelihood of criticality following disposal of ILW and DNLEU in evaporite. However, in this analysis, the evaporite host rock is assumed to be dry. Therefore, there is no significant waste form dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. There is no potential for fissile material accumulation outside waste packages or as a result of mixing of fissile material from multiple waste packages.

The conceptual model for the scenario involving increases in reactivity inside a waste package is similar to that described for a GDF in a lower-strength sedimentary host rock (Section 5.4.1). However, based on the assumption that there is no water present in the host rock, the waste packages will not become saturated and there is no potential for material to be removed from waste packages.

Creep could cause relatively rapid waste package compaction, which will increase the concentration of fissile material in the waste packages. However, with limited quantities of neutron moderators present (e.g., in grout and/or polythene) and neutron absorbing materials being retained in the waste packages (e.g., iron), the waste packages will remain substantially sub-critical. Note also that waste packages will be stacked only three high so that the effects of any increased neutron interaction between a compacted stack of waste packages will be limited.

6.5 Summary of Assumptions – LLW/ILW/DNLEU The likelihood of criticality following the disposal of LLW/ILW and DNLEU in an evaporite rock has been analysed based on consideration of the expected evolution of conditions in the GDF. The analysis has been based on many assumptions about the waste inventory, waste packaging concepts, disposal facility design, barrier system properties and GDF evolution. The key assumption made in the analysis is that host rock is dry (see Table 6.1). Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.6).



Table 6.1: Key assumptions made in the likelihood of criticality analysis for LLW/ILW/DNLEU disposal in an evaporite host rock. Alternative assumptions and their potential impacts on the analysis are noted. Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.6).



The host rock is dry Alternative assumptions about host rock properties (e.g., groundwater presence) could be made. The introduction of groundwater as a result of human intrusion and subsequent mass transfer could be considered.

The analysis undertaken for the higher-strength rock disposal concept is considered to be bounding of any situation in which there could be groundwater present. Consistent with the GRA, the effects of human intrusion would be considered in consequence analysis.


As waste packages and GDF barrier materials begin to degrade after emplacement, fissile and other materials may be mobilised, which could affect reactivity and the likelihood of criticality. Although anticipated changes in the arrangement of waste package materials might reduce disposal system reactivity through dispersion of fissile material, the likelihood of higher reactivity systems developing requires assessment. The following sub-sections discuss the expected evolution of conditions in a GDF in evaporite and identify post-closure criticality FEPs and scenarios.

6.6.1 GDF Evolution

Salt creep would begin immediately after excavation due to the presence of differential stress caused by the creation of void spaces [46, §A3.3.1 and §A3.4.2]. It is expected that creep and compaction would reduce the permeability of tunnel and salt seals to values similar to that of the host rock after about 200 years [46, A3.9.1].

The SF and HLW would generate significant amounts of heat, especially during the initial stages as a result of radioactive decay of the short-lived component of the inventory. Elevated temperatures generated by the waste would accelerate the creep rate and sealing of the GDF [46, §A3.4.2]. Complete containment through rock creep is expected after several hundred years for heat-generating wastes, whereas complete containment through creep closure would be expected to take several thousand years for non-heat generating wastes. As the heat output in the near field decreases, thermal contraction may reduce compressive stresses, or even introduce tensile stresses in the host rock causing it to crack [46, §A3.4.2].

As a consequence of their low permeability and porosity, the total amount of liquid in evaporite environments is very low and is dominated by brine trapped in the



rock [46, §3.8.3]. Any resaturation of the salt backfill would be slow (taking hundreds of years or more) [46, §A3.4.1].

SF/HLW/HEU/Pu canisters would be expected to resist the mechanical effects of creep closure and maintain their integrity until processes such as corrosion have reduced their structural strength [46, §3.9]. The disposal canisters are not expected to be significantly degraded until several thousands of years after disposal.


The criticality FEP analysis is described in detail in the supporting technical report on SF/HLW/HEU/Pu disposal [2, Appendix D] and has been structured as for the higher-strength host rock analysis described in Section 4.6.2. Again, the criticality FEP groups are discussed in detail in the supporting technical report [2, Appendix D] in terms of mechanical, chemical, thermal, gas-related, hydrological, radiological, and microbiological events and processes.

For post-closure criticality to occur, package failure followed by substantial degradation and relocation of wasteform materials would be required. However, relocation of fissile and other materials to form a critical configuration cannot be envisaged in the assumed dry GDF. Therefore, the only criticality scenario that has been defined is one involving fissile material accumulation inside a waste package.


This section considers the likelihood of criticality following disposal of SF/HLW/HEU/Pu in evaporite. However, in this analysis, the evaporite host rock is assumed to be dry. Therefore, there is no significant wasteform dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. There is no potential for fissile material accumulation outside waste packages or as a result of mixing of fissile material from multiple waste packages. Based on the assumption that there is no water present in the host rock, the waste packages will not become saturated and there is no potential for material to be removed from waste packages.

Creep will cause waste package compaction, although the waste packages would be designed to resist such compaction. Compaction will increase the concentration of fissile material in the waste packages as any void space is reduced (see Figure 6.4). However, with limited quantities of neutron moderators present and neutron absorbing materials being retained in the waste packages, the waste packages will remain substantially sub-critical.

Indeed, the enrichment of fuel used in thermal reactors is too low for criticality in any configuration without neutron moderation. The HEU and separated Pu waste packages contain 28 kg 235U and at most 20 kg 239Pu, respectively. That is, the waste packages contain more than the minimum critical masses of pure 239Pu metal and pure 235U metal spheres (21.8 kg 235U and 5.4 kg 239Pu, respectively [1, Appendix A.3]). However, in the presence of the neutron absorbing materials included in the HEU and separated Pu waste forms, the critical masses of un-moderated spheres are much greater. For example, in excess of 18 tonnes of plutonium would be required for criticality for the type of wasteform considered in this analysis [54, 2.1.1]. Therefore, the HEU and Pu waste packages would remain substantially sub-critical even if compacted by creep.



Figure 6.4: Illustrative scenarios for the compaction of HLW/SF/Pu/HEU waste packages as

a result of rock creep.

6.8 Summary of Assumptions – SF/HLW/HEU/Pu The likelihood of criticality following the disposal of HLW, SF, HEU and Pu in an evaporite rock has been analysed based on consideration of the expected evolution of conditions in the GDF. Evaporite has been assumed to be dry, such that there is no significant waste form dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. Creep of the host rock may reduce void space in waste packages, increasing fissile material concentrations, but neutron absorbing material will be retained, limiting the potential for criticality. The key assumption made in the analysis is that the host rock is dry (see Table 6.2). Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.7).

Table 6.2: Key assumptions made in the likelihood of criticality analysis for SF/HLW/HEU/Pu disposal in an evaporite host rock. Alternative assumptions and their potential impacts on the analysis are noted. Assumptions about inventory and waste packaging are as discussed for the higher-strength rock GDF analysis (Table 4.7).



The host rock is dry. Alternative assumptions about host rock properties (e.g., groundwater presence) could be made. The introduction of groundwater as a result of human intrusion and subsequent mass transfer could be considered.

The analysis undertaken for the higher-strength rock disposal concept is considered to be bounding of any situation in which there could be groundwater present. Consistent with the GRA, the effects of human intrusion would be considered in consequence analysis.



7 Summary and Conclusions RWMD is responsible for implementing geological disposal of the UK’s higher-activity radioactive wastes. RWMD’s research into geological disposal considers safety during waste transport to a disposal facility, during waste disposal operations, and once the facility has been closed. The wastes for disposal comprise a wide range of materials and include some fissile radionuclides. The presence of fissile materials requires an assessment of the potential for criticality.

After disposal, the engineered barrier system of a GDF will ensure that criticality is prevented for such time as the waste packaging affords a high level of containment. However, as waste packages begin to degrade, fissile and other materials may be mobilised and this could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to lead to criticality requires consideration.

RWMD established the Likelihood of Criticality research project to develop, document and communicate the qualitative and quantitative arguments required to evaluate the probability of nuclear criticality after closure of a GDF. The evaluation of the likelihood of criticality has been underpinned by consideration of the FEPs that could affect reactivity after GDF closure. The FEP analysis led to the construction of post-closure criticality scenarios and these scenarios have been evaluated using a range of approaches, from high-level judgments about scenario credibility, to modelling of evolving conditions in a GDF to determine if critical systems could develop. Understanding the radioactive waste inventories, the GDF concepts for the different waste types and geological settings, and the expected evolution of conditions in the different GDF concepts, as well as associated uncertainties, has been of fundamental importance to the criticality FEP and scenario analysis. Uncertainties have been accounted for in definitions of parameter value distributions.

The project has focused on the analysis of a GDF containing the Baseline Inventory of radioactive wastes and materials. The Baseline Inventory defined in the 2007 UK Radioactive Waste Inventory has been assumed, and disposal concepts for ILW/LLW, DNLEU, SF, HLW, Pu and HEU have been considered. The project has considered disposal concepts for these wastes in three generic host-rock settings (higher-strength rocks, lower-strength sedimentary rocks and evaporites). This report provides a summary of the approach taken to evaluate the likelihood of criticality for each disposal concept and the results of the analysis.

Criticality scenarios have been identified and addressed in terms of those that involve rearrangement of materials in a waste package, accumulation of fissile material in the barriers outside a waste package and accumulation of fissile material from more than one waste package. The results for LLW/ILW/DNLEU and SF/HLW/HEU/Pu disposal are described below.

7.1 LLW/ILW/DNLEU Disposal GoldSim probabilistic models have been developed to evaluate the behaviour of degrading LLW/ILW/DNLEU packages for the higher-strength rock disposal concept, whereas a more qualitative approach has been undertaken for the lower-strength sedimentary rock and evaporite disposal concepts. The main conclusions from the analysis of LLW/ILW/DNLEU



disposal in higher-strength rock, lower-strength sedimentary rock and evaporite are summarised in Table 7.1, Table 7.2 and Table 7.3, respectively.

Table 7.1: Main results of the analysis of post-closure criticality scenarios for LLW/ILW/DNLEU disposal in higher-strength rock.

Criticality Scenario Likelihood of Criticality


Criticality is not possible for ILW-INU packages.

Criticality is not possible for most ILW-LEU waste packages, but slumping of fissile material in 5C54 and MU006 waste packages could generate conditions that are critical.

Criticality is not possible for most ILW-HEU waste packages, but slumping of fissile material in 5B350, 5C50 and 5B19 waste packages could generate conditions that are critical.

Criticality is not possible for most ILW-Pu waste packages, but slumping of fissile material in 5B31 waste packages could generate conditions that are critical.


Criticality is not possible for ILW-INU packages.

Criticality is not possible for ILW-LEU waste packages.

Criticality is not possible for most ILW-HEU waste packages, but accumulations in backfill of fissile material from 5B350, 5C50 and 5B19 waste packages could generate conditions that challenge criticality safety criteria.

Criticality is not possible for most ILW-Pu waste packages, but accumulations in backfill of fissile material from 5B31, 2N100 and 2N02 waste packages could generate conditions that are critical.


Criticality is not possible, except under conditions in which uranium and plutonium solubilities and sorption are high and accumulations in backfill could generate conditions that are critical. Such conditions were achieved in about 2% of probabilistic model realisations. The fissile material in such accumulations can be traced back to the above-noted waste packages that have a high fissile material content (representing just under 2% of unshielded ILW packages).



Table 7.2: Main results of the analysis of post-closure criticality scenarios for LLW/ILW/DNLEU disposal in lower-strength sedimentary rock.



There is no significant component of groundwater flow so that the potential for material to be removed from waste packages in sufficient quantities for voids to be created and slumping to occur is small. Uranium concentrations in waste packages would gradually reduce by diffusion following loss of package integrity and dissolution in groundwater and the waste packages would remain sub-critical.


Diffusion from degraded waste packages would result in gradual migration of uranium into the backfill. Uranium concentrations would be lower than the concentrations in grout at the time of disposal.


As the system evolves and multiple waste packages fail and release uranium, the uranium concentration will tend to reduce to an average for the inventory as a whole.

Table 7.3: Main results of the analysis of post-closure criticality scenarios for LLW/ILW/DNLEU disposal in evaporite.



There is no water present so that there is no significant waste form dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. With limited quantities of neutron moderators present and neutron absorbing materials being retained in the waste packages the waste packages will remain sub-critical.


There is no water present in the host rock so that the waste packages will not become saturated and there is no potential for fissile material to be removed from waste packages.


There is no water present in the host rock so that the waste packages will not become saturated and there is no potential for fissile material migration and accumulation.

7.2 HLW/SF/HEU/Pu Disposal GoldSim probabilistic models have also been developed to evaluate the behaviour of degrading HLW/SF/HEU/Pu disposal packages for the higher-strength rock and lower-strength sedimentary rock disposal concepts, whereas a more qualitative approach has been undertaken for the evaporite disposal concept. The main conclusions from the analysis of LLW/ILW/DNLEU disposal in higher-strength rock, lower-strength sedimentary rock and evaporite are summarised in Table 7.4.



Table 7.4: Main results of the analysis of post-closure criticality scenarios for SF/HLW/HEU/Pu disposal.



The GoldSim model developed for a GDF in higher-strength rock consists of a disposal tunnel above a number of vertical deposition holes, with each deposition hole containing a waste package. The waste packages are surrounded by bentonite and the tunnel is backfilled with a mixture of crushed rock and bentonite. The waste packages are copper canisters that contain HLW, SF, HEU or Pu. The containers may be breached by, for example, corrosion, after which time dissolution and migration of uranium and plutonium from the waste packages and into the buffer, tunnel backfill and host rock may occur. Mass transfer is assumed to be advection dominated. The uranium and plutonium released from the waste package may accumulate by sorption in the engineered barriers and host rock. However, based on the expected lifetime of a copper canister under disposal conditions, 239Pu is calculated to have almost entirely decayed to 235U before canister failure.

HLW HLW packages include relatively small amounts of fissile material and accumulation of such material from many waste packages would be required for criticality. GoldSim calculations have indicated that, after loss of canister integrity, insignificant amounts of 235U are likely to migrate into any accumulation zone.

SF The Advanced Gas-cooled Reactor (AGR) SF and Pressurised Water Reactor (PWR) SF considered in this analysis are of sufficient burn-up that the fuel will always be sub-critical under natural-water-moderated conditions. The GoldSim calculations confirm that insufficient uranium for criticality would accumulate in the buffer surrounding a failed container. It is unlikely that large amounts of unirradiated or low burn-up fuel would require disposal. However, waste packages that include such fuel would exhibit higher reactivities than calculated in this analysis for post-closure criticality scenarios. In particular, waste packages that contain fresh or low burn-up PWR fuel may not be sub-critical when in-package post-closure scenarios are considered. However, the GoldSim calculations indicate that accumulations of uranium in the buffer surrounding the waste packages would not result in criticality even if the uranium derived from fresh fuel.

HEU/Pu Combined GoldSim and MCNP calculations indicate that failed HEU/Pu waste packages are sub-critical because of the neutron absorbing components of the wasteform. However, based on the calculated migration of 235U from HEU/Pu waste packages and its accumulation in the buffer over long periods (in excess of 107 years), criticality may be possible even based on average parameter values.


The GoldSim model developed for a GDF in lower-strength sedimentary rock consists of a horizontal disposal tunnel containing a number of waste packages surrounded by bentonite. The waste packages are carbon steel canisters that contain HLW, SF, HEU or Pu. The containers may be breached by corrosion, after which time dissolution and migration of uranium and plutonium from the waste packages and into the tunnel backfill and host rock may occur. Mass transfer is assumed to be diffusion dominated. The uranium and plutonium released from the waste package may accumulate by sorption in the backfill and host rock.

HLW Results of calculations for HLW are similar to those obtained for disposal in higher-strength rock. GoldSim calculations have found that, after loss of canister integrity, insignificant amounts of 235U are likely to migrate into any accumulation zone.




SF The GoldSim calculations confirm that insufficient uranium for criticality would accumulate in the backfill surrounding a failed container if credit is taken for burn-up. However, enough 235U for criticality could accumulate in the buffer if the uranium derives from fresh PWR fuel.

HEU/Pu Results of calculations for HEU/Pu are similar to those obtained for disposal in higher-strength rock. Criticality may be possible on a timescale of 106 years as a result of accumulation of fissile material in the backfill, even based on average parameter values.

Evaporite

No modelling has been undertaken for disposal of SF/HLW/HEU/Pu in an evaporite host environment because the rock has been assumed to be dry. Under such conditions, there would be no significant wasteform dissolution and no mechanism for mass transfer other than as a result of the effects of rock creep. Creep of the host rock may reduce void space in waste packages, increasing fissile material concentrations, but neutron absorbing material will be retained, limiting the potential for criticality for all of the wasteforms.

7.3 Key Assumptions The analysis of post-closure criticality scenarios has, of course, been based on many assumptions about the waste inventory, waste packaging concepts, disposal facility design, barrier system properties and GDF evolution. Key assumptions are as follows:

• Based on the assumed 2007 UK RWI, a number of waste streams will be packaged such that the packages have a high fissile material content. However, few of the waste packaging concepts where the fissile material content is high have been assessed by RWMD through application of the Letter of Compliance disposability assessment process. Also, alternative approaches to waste packaging (e.g., involving vitrification, non-encapsulation or immobilisation in polymers) have been and may be proposed by waste producers for some wastes. Other packaging concepts are still under development, such as for HEU and Pu disposal. The methodologies and tools developed in the project could be used or developed to support future assessments of waste packaging proposals based on understanding of the expected behaviour of the waste packages under disposal conditions.

• The Likelihood of Criticality analysis has been undertaken based on average waste package inventories. Large variations about the average fissile material content of waste packages for any one waste stream could affect the conclusions of the criticality scenario analysis.

• Uniform chemical conditions have been assumed in the GoldSim calculations. Spatially variable chemical conditions could result in regions in which higher concentrations of fissile material occur than have been calculated in this analysis. However, the current calculations using generic sorption parameters are considered to bound the effects of precipitation. Also, potential concentrations of natural uranium in groundwater have not been considered. Assuming groundwater to include concentrations of natural uranium would reduce the rate of release of uranium from waste packages and the enrichment of dissolved uranium in the GDF would be lower than calculated in this analysis.



• Slumping of fissile material through voids in LLW/ILW/DNLEU disposal vaults has been assumed to occur on timescales of several hundred thousand years to hundreds of millions of years following dissolution of waste encapsulation grout. However, project work to improve understanding of GDF evolution has found that there may be little potential for void formation under expected geochemical and hydrological conditions. This is a significant finding because it means that assumptions made in this project about grout behaviour may be cautious rather than realistic and post-closure criticality scenarios involving fissile material slumping on the package scale or disposal vault scale may not be credible.

• Many waste packages contain significant quantities of iron, and iron corrosion products have largely been assumed to be dissolved and removed from corroded waste packages, which may be cautious. However, analysis of the behaviour of iron corrosion products under disposal conditions has indicated that iron would remain in solid form for long periods. The persistence of iron may significantly reduce the likelihood of criticality within a waste package.

• The waste packages in LLW/ILW/DNLEU disposal vaults have been assumed to be randomly distributed. If similar waste packages with high fissile material content are emplaced together in a vault, then this may also affect the conclusions of the criticality scenario analysis.

• SF/HLW/HEU/Pu canisters have been assumed to contain wastes for on the order 105 years or more. Canister failure prior to significant 239Pu decay would result in higher calculated reactivities. In particular, if early failure of copper canisters is deemed to be possible, this could have an important impact on the results of the Likelihood of Criticality project for the higher-strength rock disposal concept.

7.4 Conclusions In conclusion, the Likelihood of Criticality research project has provided qualitative and quantitative arguments about the probability of criticality after closure of a GDF, based on assumptions about GDF evolution and, in particular, uranium and plutonium behaviour. The project has provided results and tools that can be used to support a demonstration, as part of an environmental safety case, that the possibility of a local accumulation of fissile material in a GDF such as to produce a neutron chain reaction is not a significant concern, thereby meeting regulatory requirements on GDF post-closure criticality safety.



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Documents

The Likelihood of Criticality: Synthesis report