Description
These Radioactive Waste Management Arrangements will fulfil the requirements of the Environment Agency Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant Designs (1), together with consideration of the Environment Agency REPs (2) and the applicable requirements of the Office for Nuclear Regulation (ONR) for the management of wastes and spent fuel (SF) from the Hitachi-GE UK ABWR.
Form10/00
Hitachi-GE Nuclear Energy, Ltd.
UK ABWR
UK ABWR Generic Design Assessment
Radioactive Waste Management Arrangements
Document ID : GA91-9901-0022-00001
Document Number : WE-GD-0001
Revision Number : C
Form10/00
Hitachi-GE Nuclear Energy, Ltd.
UK ABWR
DISCLAIMERS
Proprietary Information
This document contains proprietary information of Hitachi-GE Nuclear Energy, Ltd. (Hitachi-GE), its
suppliers and subcontractors. This document and the information it contains shall not, in whole or in part,
be used for any purpose other than for the Generic Design Assessment (GDA) of Hitachi-GE’s UK ABWR.
This notice shall be included on any complete or partial reproduction of this document or the information it
contains.
Copyright
No part of this document may be reproduced in any form, without the prior written permission of
Hitachi-GE Nuclear Energy Ltd. Copyright (C) 2014 Hitachi-GE Nuclear Energy, Ltd. All Rights
Reserved.
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Table of Contents
1. Acronyms ................................................................................................................................................ 1
2. References ............................................................................................................................................... 6
3. Introduction ............................................................................................................................................. 8
4. Waste Management Objectives and Principles........................................................................................ 8
4.1. Objectives ....................................................................................................................................... 8
4.2. Principles ........................................................................................................................................ 9
5. Assumptions ............................................................................................................................................ 9
6. Regulatory Context ............................................................................................................................... 10
6.1. Consideration of the REPs ............................................................................................................ 11
7. The Decision Making Process ............................................................................................................... 11
8. Management of the Radioactive Waste Streams ................................................................................... 12
8.1. Wastes Arising .............................................................................................................................. 15
8.1.1. Very Low Level Waste ............................................................................................................. 15
8.1.2. Low Level Waste ...................................................................................................................... 16
8.1.3. Intermediate Level Waste ......................................................................................................... 19
8.1.3.1. General Packaging Options .................................................................................................. 20
8.1.3.2. Sludge (Crud) ....................................................................................................................... 21
8.1.3.3. Powder Ion Exchange Resins ............................................................................................... 22
8.1.3.4. Higher Activity Metals ......................................................................................................... 22
8.1.3.5. Irradiated Metal (Decommissioning) ................................................................................... 23
8.1.3.6. Borderline Wastes ................................................................................................................. 23
8.2. Spent Fuel ..................................................................................................................................... 24
8.2.1. Fuel Characteristics .................................................................................................................. 24
8.2.2. Packaging Options .................................................................................................................... 26
8.2.2.1. Dry Cask............................................................................................................................... 26
8.2.2.2. Multi-Purpose Container ...................................................................................................... 27
8.2.2.3. Modular Vault Dry Storage System ...................................................................................... 27
8.2.2.4. KBS-3 Container .................................................................................................................. 27
8.2.2.5. On-Site Fuel Pool Option ..................................................................................................... 28
8.3. Summary of Management Options ............................................................................................... 28
9. Waste Management Infrastructure ......................................................................................................... 29
9.1. Waste Treatment and Packaging Facilities ................................................................................... 30
9.1.1. Low Level Waste .................................................................................................................. 30
9.1.2. Intermediate Level Waste ..................................................................................................... 30
9.1.3. Spent Fuel............................................................................................................................. 31
9.2. Waste Stores ................................................................................................................................. 31
9.2.1. Low Level Waste .................................................................................................................. 31
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9.2.2. Intermediate Level Waste ..................................................................................................... 31
9.2.3. Spent Fuel............................................................................................................................. 32
9.3. Waste Packages, Transport Systems and LoCs ............................................................................. 33
10. Development of an Integrated Waste Strategy ...................................................................................... 34
11. Plan for Disposability Assessments ....................................................................................................... 34
11.1. Introduction to the RWMD Disposability Assessment ............................................................. 34
11.2. Hitachi-GE Plan for the RWMD Disposability Assessment ..................................................... 35
12. Conclusions ........................................................................................................................................... 35
Appendix A – Consideration of the REPs ..................................................................................................... 37
Appendix B - Waste Hierarchy ...................................................................................................................... 39
Appendix C - Waste Stream and Spent Fuel Descriptions ............................................................................ 40
1. Introduction ........................................................................................................................................... 40
2. Nature and Quantity of the Wastes and Spent Fuel ............................................................................... 40
2.1. Operational VLLW ....................................................................................................................... 40
2.2. Operational LLW .......................................................................................................................... 41
2.3. Operational ILW ........................................................................................................................... 44
2.4. Decommissioning Non-Radioactive Waste .................................................................................. 46
2.5. Decommissioning VLLW ............................................................................................................. 46
2.6. Decommissioning LLW ................................................................................................................ 47
2.7. Decommissioning ILW ................................................................................................................. 48
2.8. Spent Fuel ..................................................................................................................................... 48
Appendix D - Descriptions of Conditioning Options .................................................................................... 50
1. Introduction ........................................................................................................................................... 50
Notes for Figure D1 – Very and Low Level Wastes ...................................................................................... 52
Notes for Figure D2 – Intermediate Level Waste; Solidification of Sludge/ Crud and IX Resins ................ 55
Notes for Figure D3 – Intermediate Level Waste; Encapsulation of Activated Metals ................................. 58
Notes for Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud and IX Resins ............................ 61
Notes for Figure D5 – Intermediate Level Waste; Drying of Activated Metals ............................................ 64
Notes for Figure D6 – Spent Fuel Management Options .............................................................................. 67
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1. Acronyms
ABWR Advanced Boiling Water Reactor
AC Atmospheric Control System
ALARA As Low As Reasonably Achievable
ALARP As Low As Reasonably Practicable
BAT Best Available Technique
BPEO Best Practicable Environmental Option
BPM Best Practicable Means
Bq Becquerel
BSS Basis Safety Standards Directive
BWR Boiling Water Reactor
C&I Control and Instrumentation
CAD Controlled Area Drain
CCI Commercially Confidential Information
CD Condensate Demineraliser
CDL Calculated Detection Limit
CF Condensate Filter
COMAH Control of Major Accident Hazards
CONW Concentrated Waste System
CP Corrosion Product
CSG Combustion Sector Guidance Note
CST Condensate Storage Tank
CUW Reactor Water Clean-up System
CW Circulating Water System
CWP Circulating Water Pump
D/W Dry Well
DAW Dry Active Waste
DCD Design Control Document
DECC Department of Energy and Climate Change
DEFRA Department for Environment, Food and Rural Affairs
DF Decontamination Factor
DORIS The marine dispersion model used in PC-CREAM 08
®
DPUR Dose Per Unit Release
EIA Environmental Impact Assessment
EMCLs Environmental Media Concentration Limits
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EPR/EPR10 Environmental Permitting (England and Wales) Regulations 2010
EQS Environment Quality Standards
ERICA Environmental Risk from Ionising Contaminants: Assessment and Management
ESE Environmentally Sensitive Equipment
EU European Union
f-value Fuel leakage rate
F/D Filter Demineraliser
FAP Forward Action Plan
FDP Funded Decommissioning Programme
FDW Feedwater System
FP Fission Product
FPC Fuel Pool Cooling and Clean-up System
GDA Generic Design Assessment
GDF Geological Disposal Facility
GEP Generic Environmental Permit
GNF Global Nuclear Fuel
GSD Generic Site Description
HAW Higher Activity Waste
HCEP How to Comply with your Environmental Permit
HCW High Conductivity Waste System
HEPA High Efficiency Particulate Air (Filter)
HFE Human Factors Engineering
HFF Hollow Fibre Filter
HLW High Level Waste
HNCW HVAC Normal Cooling Water System
HOP Hydrazine, Oxalic acid, Potassium permanganate
HS Heating Steam System
HSCR Heating Steam and Condensate Water Return System
HSD Hot Shower Drain
HSE Health and Safety Executive (UK)
HVAC Heating Ventilation and Air Conditioning System
HWC Hydrogen Water Chemistry
I&C Instrumentation and Control
IA Instrument Air System
IAEA International Atomic Energy Agency
ICRP International Commission on Radiological Protection
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IEX Ion-Exchange (demineraliser) system
ILW Intermediate Level Waste
IPPC Integrated Pollution Prevention and Control
IRA Initial Radiological Assessment
IWS Integrated Waste Strategy
J ABWR Japanese ABWR
KK-6 Kashiwazaki-Kariwa Nuclear Power Station Unit 6
KK-7 Kashiwazaki-Kariwa Nuclear Power Station Unit 7
LCW Low Conductivity Waste System
LD Laundry Drain System
LLW Low Level Waste
LLWR Low Level Waste Repository
LoC Letter of Compliance
LOCA Loss of Coolant Accident
LPRM Local Power Range Neutron Monitor
LS Laundry System
LWR Light Water Reactor
MCERTS Monitoring Certification Scheme
MS Main Steam System
NDA Nuclear Decommissioning Authority
NHS Non Human Species
NMCA Noble Metal Chemical Addition
NPP Nuclear Power Plant
NRW Natural Resources Wales
NUREG Nuclear Regulatory Commission Regulation (US)
OG Off gas
ONR Office for Nuclear Regulation
OSPAR Oslo and Paris Convention on Protection of the Marine Environment of the North East
Atlantic
P&D Plumbing and Drainage System
P&ID Process and Information Document for Generic Assessment of Candidate Nuclear
Power Plant Design
P/C Power Centre
PCI Pellet Cladding Interaction
PCSR Pre-Construction Safety Report
PI Personal Information
ppb parts per billion
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PWR Pressurised Water Reactor
QA Quality Assurance
QAP Quality Assurance Plan
QC Quality Control
QMP Quality Management Plan
QMS Quality Management System
R/B Reactor Building
RCLEA Radioactively Contaminated Land Exposure Assessment
RCW Reactor Building Cooling Water System
REP Radioactive Substances Regulation – Environmental Principle
RGP Relevant Good Practice
RP Requesting Party
RPDP Radiation Protection Developed Principle
RQ Risk Quotient
RSA Radioactive Substances Act
RSR Radioactive Substances Regulation
RSW Reactor Building Service Water System
RW/B Radwaste Building
RWMA Radioactive Waste Management Arrangement
RWMD Radioactive Waste Management Directorate
S/B Service Building
S/P Suppression Pool
SA Station Service Air System
SAM Sampling System
SAP Safety Assessment Principle
SF Spent Fuel
SFAIRP So Far As Is Reasonably Practicable
SFP Spent Fuel Pool
SGTS Standby Gas Treatment System
SJAE Steam Jet Air Ejector
SLC Standby Liquid Control System
SoDA Statement of Design Acceptability
SPCU Suppression Pool Clean-up System
SQEP Suitably Qualified and Experienced Person (UK)
SRNM Start-up Range Neutron Monitor
SS Spent Sludge System
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Sv Sievert
TBURD Technical Baseline and Underpinning Research and Development
T/B Turbine Building
TIP Traversing In-core Probe
TCW Turbine Building Cooling Water System
TSW Turbine Building Service Water System
TV Tank Vent Treatment System
UF Uncertainty Factor
UK United Kingdom
US United States
VLLW Very Low Level Waste
WENRA Western European Nuclear Regulators' Association
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2. References
1 Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant
Designs, Version 2, March 2013.
2 Regulatory Guidance Series, No RSR 1, Radioactive Substances Regulation – Environmental
Principles, Version 2, April 2010.
3 Demonstration of BAT, GA91-9901-0023-00001, XE-GD-0097, Rev C, March 2014.
4 Fundamentals of the management of radioactive waste: An introduction to the management of
higher-level radioactive waste on nuclear licensed sites: Guidance from the Health and Safety
Executive, the Environment Agency and the Scottish Environment Protection Agency to nuclear
licensees, December 2007.
5 An Overview of NDA Higher Activity Waste, February 2012.
6 UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear
Industry, August 2010.
7 Waste Acceptance Criteria – Very Low Level Waste Disposal, LLW Repository Ltd,
WSC-WAC-VER – Version 3.0 – April 2012.
8 Waste Acceptance Criteria – Low Level Waste Disposal, LLW Repository Ltd, WSC-WAC-LOW
– Version 3.3 –April 2012 (currently subject to revision).
9 Waste Acceptance Procedure – Overview, LLW Repository, WSC-WAP-OVR – Version 2.0 –
January 2011.
10 Justification for no active oils generated for ABWR operations; WE-GD-0005; Rev.0; February
2014.
11 Waste Package Specification for 2 metre box waste package, NDA (RWMD), WPSG No. 350/3,
January 2013.
12 Waste Package Specification for 4 metre box waste package, NDA (RWMD), WPSG No. 330/3,
January 2013.
13 http://www.nda.gov.uk/rwmd/producers/detail.cfm#specifications.
14 NDA Technical Note no. 13403461, Geological Disposal, Generic specification for robust shielded
waste packages November 2010.
15 Magnox Care and Maintenance Preparation Wastes in Ductile Cast Iron Containers (Conceptual
stage) Summary of Assessment Report, Issue date of assessment report, 19 May 2010.
16 Croft Rectangular Box Safstores® for the packaging of ILW (Conceptual stage) Summary of
Assessment Report Issue date of assessment report, 24 February 2012.
17 Guidance on decision making for management of wastes close to the LLW and ILW categorisation
boundary that could potentially cross the LLW boundary, Helen Cassidy, NWP/REP/016, Issue 2 –
February 2013.
18 A White Paper on Nuclear Power, Department for Business, Enterprise and Regulatory Reform,
CM 7296, January 2008.
19 http://www.wnfc.info/proceedings/2004/presentations/woodwardppt.pdf.
20 Packaging of Sizewell B Spent Fuel (Pre-Conceptual stage), Summary of Assessment Report,
Issue date of assessment report 23 December 2011.
21 http://www.skb.se/Templates/Standard____24109.aspx.
22 Industry Guidance, Interim Storage of Higher Activity Waste Packages – Integrated Approach,
Effective from November 2012.
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23 ENG01 - Specification and Guidance on the Content and Format of an Integrated Waste Strategy,
Revision 3, 19 October 2012.
24 Latest Liquid Waste and Solid Waste Treatment System and Development, WJ-GD-0017, to be
provided at a later date.
25 Solid waste generation arising from operation and decommissioning, WJ-GD-0018, to be provided
at a later date.
26 The Activities of Solid wastes, WJ-GD-0019, to be provided at a later date.
27 ABWR Plant General Description, December 2006.
28 Parameters for spent fuel assessment in GEP, UE-UK-0001, April 2013.
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3. Introduction
These Radioactive Waste Management Arrangements will fulfil the requirements of the Environment
Agency Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant
Designs (1), together with consideration of the Environment Agency REPs (2) and the applicable
requirements of the Office for Nuclear Regulation (ONR) for the management of wastes and spent fuel
(SF) from the Hitachi-GE UK ABWR. The Hitachi-GE Radioactive Waste Management Arrangements
have been produced with potential site operators in mind to ensure that the site specific Integrated Waste
Strategy (IWS) can follow on from these generic arrangements without major changes to the fundamental
approach.
This document describes the radioactive waste management arrangements, including how radioactive
wastes and SF will arise throughout the facility’s lifecycle (including decommissioning), and how they will
be managed and disposed of.
This document will focus on those wastes which will be treated and packaged in some way either for
immediate disposal or on-site interim storage as ‘solid’ waste, pending disposal at a later date. The systems
in place for the management and treatment of Gaseous and Aqueous Radioactive Discharges are described
in the suite of common documents, and dedicated chapters provided as part of the PCSR and will be
underpinned by a demonstration of BAT (3).
The Hitachi-GE Radioactive Waste Management Arrangements have been developed with the UK
regulatory system in mind for the UK ABWR. Where applicable, the current Japanese practice for dealing
with a specific waste stream will be outlined and discussed in the context of current UK practices. Where
UK practice differs this will be noted and management options will be discussed for the specific stream. In
such circumstances a recommendation for additional work will be included to develop appropriate
solutions, which may, at this stage comprise a number of suitable options. These arrangements will be
updated to reflect any changes or on-going developments.
4. Waste Management Objectives and Principles
The following high level objectives and principles are used within Hitachi-GE to inform decision making
and design development for systems having a bearing on the generation or treatment of radioactive wastes.
These have been developed within Hitachi-GE with future site operators in mind. The waste management
objectives and principles are used to guide the development of strategies and in due course, their
implementation on specific sites. These take due consideration of the objectives and principles published
by HSE (4) and Environment Agency (2).
4.1. Objectives
The objectives for the Hitachi-GE and UK ABWR Radwaste Management Arrangements are to:
? Safely control and account for radioactive waste;
? Protect human health and the environment both now and in the future. Where it is reasonably
practicable, deliver this objective by concentrating, containing and packaging the waste and
isolating it from the accessible environment by disposing of it in appropriate geological facilities
at the earliest opportunity;
? Balance the use of environmental, social and economic resources in an optimal way;
? Ensure undisturbed power production from the reactor, provided health, safety and environmental
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protection are not compromised.
4.2. Principles
The principles to be applied when developing and ultimately implementing waste management and
decommissioning strategies will be:
? The safety and protection of the public, the workforce and the environment, are of prime
importance and wastes should be managed to ensure an optimal level of protection. Sustainability
should be considered such that unduly restrictive burdens are not left for future generations;
? Strategies and practices will be compliant with UK Government and relevant international policies,
legislation and standards, including the policy aim of sustainable development;
? Mature and/or standardised options should be given preference;
? All radioactive wastes should be managed within strategies that are integrated, take account of
interactions and dependencies, and consider the full lifecycle of the wastes, including the
possibility of long term storage prior to disposal;
? Strategies will take account of the “Waste Hierarchy” as described in Appendix B in an attempt to
avoid waste generation and where this is not possible, to minimise the quantities and activities of
waste, so far as is reasonably practicable;
? Characterisation and segregation of wastes should be used to help ensure subsequent management
is safe, efficient and effective;
? Strategies should be informed by the best scientific knowledge. Technologies and techniques
which are planned to be deployed should be readily available, or be capable of development with
planned and realistic levels of R&D;
? The accumulation of wastes should be minimised and they should not intentionally be stored in
their raw, as produced, form for any longer than is necessary to organise their efficient packaging
in a way that is acceptable to the organisation responsible for disposal. The waste treatment and
packaging process should ensure that waste is processed into a passively safe condition as soon as
practicable;
? Appropriate documentation and records will be produced, stored and maintained covering the
selection and design of waste treatment processes, their operation and the characteristics of the
raw and packaged waste;
? Waste packages should be accessible during interim storage.
5. Assumptions
The development of these waste management arrangements is based on a number of assumptions of which
the following are the main high level ones:
? The site comprises one reactor unit;
? The design basis for the ABWR fuel is to use the GE14 type (Ref Global Nuclear Fuel BWR
design);
? Government Policy, standards, legislative and regulatory environments remain unchanged, or
changes pending have no significant impact;
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? Strategies detailed within these arrangements will reflect only currently available technologies;
? Definitions of waste categories will remain unchanged;
? The waste management arrangements take as a starting point that the wastes that will be generated
from the reference reactor design and the processes, equipment and facilities comprising that
design;
? It is assumed that SF will not be reprocessed but will be treated as a waste stream and stored,
packaged and disposed of accordingly;
? Because the GDF for HAW (ILW and HLW, including SF) is still only at a conceptual stage and
detailed packaging requirements are not available, the focus of these arrangements is to ensure that
ILW and SF can be safely stored on-site for an extended interim period of many decades;
? Materials from the abatement of liquid and gaseous discharges (such as ion exchange resins and
charcoal filter media) are included within these arrangements, but the actual discharges to the
environment are discussed in a separate paper;
? LLWR (or a successor) is available throughout the operational and decommissioning phases and
its “Conditions for Acceptance” (CFA) continue to apply unchanged. LLWR Ltd offers a
well-recognised service for the whole UK Nuclear industry. It is currently adopted by many
operational sites in the UK and also by the legacy sites. However, there are a number of other
facilities which can be used for waste disposals which are not listed in this document. The
references to LLWR are assumed for GDA and to enable the obtaining of an ‘Acceptance in
Principle’ for management/disposal of VLLW and LLW based upon the services undertaken by
LLWR Ltd. When appropriate the application of BAT will be implemented by the licensee to
determine the best route for specific wastes, when other facilities and techniques for management
of VLLW and LLW will be included;
? These arrangements take into account the liquid waste streams produced by a UK ABWR which
are delivered to the Waste Treatment Facility for treatment and solidification. It does not address
liquid and gaseous waste discharges, non-radioactive waste generated at the plant or wastes
generated from the operation of the interim storage facilities;
? These arrangements assume that the options presented can be used by utilities to demonstrate that
the design makes use of Best Available Techniques (BAT). The intention is to preserve as much
flexibility as possible whilst providing confidence that BAT solutions can be identified by utilities
in the future. This is judged to be sensible given the range of uncertainties at this stage. In
particular, issues relating to the future treatment and transport of certain wastes may need to be
resolved via consultations with RWMD.
6. Regulatory Context
The regulatory framework will be as currently applicable to UK Nuclear facilities. This framework is
largely non-prescriptive and a future operator must demonstrate to the Regulators that they fully
understand the hazards and risks associated with their operations and know how to control and reduce
them. Similarly, they are obliged to assess and minimise their impact on the environment.
To achieve high standards of nuclear safety and to restrict environmental impact, a UK operator must
adhere to the principle of continuous improvement, which is embedded in UK law, and requires nuclear
designers and operators to reduce risks as far as is reasonably practicable.
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In managing radioactive wastes in the UK, the main stakeholders for Hitachi-GE and their responsibilities
are:
? Government – To determine policy, in the light of international agreements and guidance, and
prepare statutory legislation;
? Regulators – To enforce the law and pay cognisance to Government policy and publish guidance
which interprets Government policy;
? Operators – To implement appropriate waste management strategies, in compliance with
Government policies and legislation and regulatory requirements/guidance.
The main policies, principles, guidance and regulations that relate to waste management and
decommissioning are identified below:
? International Directives and initiatives:
o The OSPAR Convention;
o The Basic Safety Standards Directive (Council Directive 96/29/EURATOM);
o Euratom Treaty Article 37.
? National Policy:
o The Review of Radioactive Waste Management Policy: Final Conclusions (Cmnd 2919)
1995, as amended;
o UK Strategy for Radioactive Discharges 2011-2020;
o Managing Radioactive Waste Safely (MRWS) – Proposals for developing a policy for
managing solid radioactive waste in the UK.;
o Policy for the Long Term Management of Solid Low Level Radioactive Waste in the
United Kingdom, By Defra, DTI and the Devolved Administrations, March 2007.
? National Legislation:
o Nuclear Installations Act 1965;
o Environmental Permitting Regulations 2010/2011;
o Ionising Radiation Regulations 1999;
o Pollution Prevention and Control 2000.
? Regulatory Guidance:
o Regulation Environmental Principles (2);
o Guidance to Inspectors on the Management of Radioactive Materials and Radioactive
Waste on Nuclear Licensed Sites.
6.1. Consideration of the REPs
As noted above, the REPs, which are listed in (2), have been considered during the development of these
arrangements. In particular, the extent to which the relevant REPs listed in the Environment Agency
Process and Information Document (1) has been addressed is shown in Appendix A.
7. The Decision Making Process
The decision making process will be as currently adopted within Hitachi-GE and as informed by the
Demonstration of BAT (3). The decision making process has been developed, with future site operators in
mind – to allow them to demonstrate full understanding of why certain techniques have been adopted, and
to allow for site specific criteria to be included.
Hitachi-GE recognises that there is a need to underpin the selection of optimal disposal routes by the
demonstration that those routes represent the BAT. These arrangements are not supported by a full BAT
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assessment as this will be conducted at a site specific stage. However, this Radioactive Waste Management
Arrangements document refers to the requirement to apply the BAT process (3), which addresses the
Claims, Arguments and the main Evidence required to select the optimal disposal route for GDA, as
described in Claim 4 of the Demonstration of BAT (3).
Where required, the appropriate technique which includes Cost/Benefit analysis will be undertaken as part
of Hitachi-GE’s decision making process.
The noted BAT assessment will also take account of application of the Waste Hierarchy and ALARP
considerations to determine optimal disposal routes and levels of discharge. Hence, such analysis will not
be repeated in this document.
The decision making process within Hitachi-GE includes making a judgement on whether a specific waste
route is considered BAT. As part of the Hitachi-GE Management arrangements information will be
gathered in the following areas to support the decision making process:
? The number of people (workers and the public) and other environmental targets that may be
exposed to radiological risk;
? The likelihood that they could be exposed to radiation, where exposure is not certain to happen;
? The magnitude and distribution in time and space of radiation doses that they will or could
receive;
? Nuclear security and safeguards requirements;
? Issues similar to those above, but relating to non-radiological hazards;
? Economic, societal and environmental factors;
? Technical viability;
? Uncertainties in any of the above.
Any decision, and the evidence used to make that decision, will be presented as Evidence in the BAT case.
8. Management of the Radioactive Waste Streams
In the UK, radioactive wastes are classified in terms of the nature and quantity of radioactivity that they
contain and their heat-generating capacity, as follows:
? High Level Wastes (HLW) which includes SF are wastes in which the temperature may rise
significantly as a result of their radioactivity, so this factor has to be taken into account in the
design of storage or disposal facilities;
? Intermediate Level Wastes (ILW) are wastes exceeding the upper boundaries for LLW, but which
do not require heat generation to be taken into account in the design of storage or disposal
facilities;
? Low Level Wastes (LLW) are wastes having a radioactive content not exceeding 4 GBq per tonne
of alpha activity, or 12 GBq per tonne of beta/gamma activity;
? Additionally, Very Low Level Waste (VLLW) is a subset of LLW and there is also a category of
exempt waste which does not require an authorisation for disposal.
Higher Activity Wastes (HAW) comprises all HLW and ILW, and a small fraction of LLW with a
concentration of specific radionuclides that prohibits its disposal at existing and planned future disposal
facilities for LLW. The definitions of these categories are contained within (5).
The radioactive wastes and SF that Hitachi-GE expects the UK ABWR to generate are grouped in Table
8-1 into 13 waste streams and one SF stream. The quantities of these wastes have been minimised as
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described in Claim 3 of the Demonstration of BAT (3). They are described in more detail in Appendix C.
Each waste type is described in turn below and they will be managed according to their categorisation as
VLLW, LLW, ILW or SF.
Table 8-1 Summary of Waste and Spent Fuel Streams
No. Title Description Category Form Arising during
1 Dry active waste Miscellaneous dry, low
activity wastes in various
forms; including metals,
concrete cloths, paper etc.
VLLW
Solid
Operations &
Decommissioning
2 HVAC Filters Arising from filter changing
in air treatment facilities from
exhausts from Reactor,
Turbine (including high
radiation), Radwaste and
Service buildings
LLW Solid Operations
3 Bead resin Arising from the CD, LCW
and HCW demineralisers;
Styrene divinylbenzene
copolymer matrix
LLW Wet Operations
4 Concentrates Arising from the HCW
evaporators comprises
particulate and dissolved
species.
LLW Wet Operations
5 Miscellaneous
combustible
Includes spent hollow fibre
filter membrane, plastic
sheets, paper, wood cloth,
spent bead resin
LLW Solid Operations
6 Miscellaneous
non-combustible
Metal, pipes, cables, lagging,
gas filters, concrete, glass
LLW Solid Operations
7 Sludge (crud) Arising from backwashing of
various filters from the CF
and the LCW systems
ILW Wet Operations
8 Powder resin Arising from the CUW and
FPC filter demineralisers;
cross linked polystyrene
matrix. Contains particulate
corrosion product
ILW Wet Operations
9 Higher activity
metals – control
rods
Cruciform shape metallic
construction containing
stainless steel tubes in each
wing of the cruciform filled
with boron carbide powder.
Hafnium may also be
employed to perform the
same function of reactivity
control
ILW Solid Operations
10 Higher activity Zircaloy box which ILW Solid Operations
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No. Title Description Category Form Arising during
metals – channel
boxes
1
surrounds the fuel bundle.
Approx. 4.3 m long and 15 x
15 cm square
11 Higher activity
metals - others
Various reactor core
components from SRNM and
LPRM systems.
ILW Solid Operations
12 Contaminated
and irradiated
metal and
concrete
Non-combustible, largely
metal and concrete items.
Some very large and
requiring size reduction
LLW Solid Decommissioning
13 Irradiated metal Reactor core components and
from areas subject to
activation
ILW
Solid
Decommissioning
14 Spent Fuel Used fuel elements HLW Solid Operations
Following its generation, radioactive waste will undergo a number of management stages, before it is
finally disposed of. Joint guidance from the HSE, Environment Agency and SEPA identifies six stages (4)
comprising pre-treatment, treatment, packaging, storage, retrieval and disposal. The Regulators note that
implementation depends on the type of waste and the strategy selected for its management.
For managing the waste streams from UK ABWR, the following stages will be used:
? Short term temporary buffer storage – segregated waste may be temporarily stored, for
example, in a buffer tank in order to accumulate an adequate batch size for treatment. At this stage
or during subsequent treatment stages the waste stream will be characterised to determine its
physical, chemical and radionuclide properties. Characterisation methods will include the use of
sampling and analysis (typically for wet wastes) and radiological assay (typically for solid
wastes);
? Treatment – involves changing the characteristics of the waste by processes such as volume
reduction (using drying, cutting, compaction or incineration), cleaning/decontamination by
filtration or ion exchange, or precipitation;
? Packaging for Interim Storage - involves transforming the waste into a form that is suitable for
handling, transfer, storage and, potentially, disposal. This might involve immobilisation of the
waste and placing it into steel drums or other engineered containers to create a waste package. In
the case of VLLW, LLW and ILW, such a package should be suitable for disposal but for SF it is
possible that additional packaging may be needed at the end of interim storage and before
disposal;
? Transfer to Interim Store – movement of the package, within a shielded overpack if necessary,
from the packaging facility to the on-site interim store;
? Interim Storage – on-site interim storage will be necessary for ILW and SF until the planned
Geological Disposal Facility (GDF) is available to receive the waste. Waste may need to be stored
in a passively safe condition for many decades. In the case of LLW, temporary storage of
packaged waste may be required in order to accommodate the requirements of the transport
system and the receipt arrangements at a disposal site (e.g. the Low Level Waste Repository);
1
Classified here as ILW, although for disposal it will be assumed that they remain with the spent fuel
elements
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? Packaging for Disposal – it is possible that SF may need to be re-packaged in order to make it
suitable for disposal. It is likely that such packaging could not be defined confidently until an
actual GDF is available;
? Transport for Disposal – movement of the packaged waste and SF, within suitably shielded and
protected flasks or containers, from the reactor site to the GDF (for SF and ILW), and to a suitably
permitted disposal site for VLLW and LLW (e.g. LLWR);
? Disposal - packages of radioactive waste are deposited in a disposal facility with no intention of
retrieval.
Adequate records will need to be kept during the whole management cycle, as outlined above. In particular
NDA/Radioactive Waste Management Directorate (RWMD) require that records are kept to ensure that the
history of the wastes and specific package properties are preserved to inform their disposability and that
their long term evolving properties will remain within acceptable limits.
8.1. Wastes Arising
The following Sections provide additional information of each category of waste arising together with the
options for their management.
For all VLLW and LLW streams, “Agreements in Principle” will be sought to demonstrate that the wastes
can be managed and, where appropriate, disposed of in the UK. For the ILW and spent fuel streams a
disposability assessment will be sought from RWMD to confirm acceptability of these, when suitably
packaged and subject to detailed LoC assessments, into a future GDF.
The following Sections describe each category of waste envisaged to arise from a UK ABWR and the
options for packaging, interim storage and disposal. These options are currently based upon good practice
as applied to similar wastes on some UK legacy sites which would be considered as BAT at this time.
However, the selection of the appropriate option by the operator, when developing their plans for
packaging, interim storage and disposal, will be subject to the application of BAT within their decision
making process.
8.1.1. Very Low Level Waste
VLLW is a subset of LLW and falls into two distinct categories:
? Low Volume VLLW (‘dustbin disposal’) - wastes that can be safely disposed of to an unspecified
destination with municipal, commercial or industrial waste, each 0.1 cubic metre of material
containing less than 400kBq of total activity, or single items containing less than 40 kBq of total
activity. There are additional limits for carbon (C-14) and tritium (H-3) in wastes containing these
radionuclides.
The radioactive risk from such material is low enough that controls on disposal, after it has been
removed from the premises where it arose, are not necessary;
? High Volume VLLW (bulk disposals) – wastes with maximum concentrations of 4 MBq per tonne
of total activity or 40 MBq per tonne for tritium total activity that can be disposed to specified
landfill sites. There is an additional limit for tritium (40 MBq/t) in wastes containing this
radionuclide (7).
After the waste is removed from its site of origin, it will be subject to controls on its disposal, which will
be specified by the environmental Regulators.
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For a nuclear power station the “dustbin disposal” sub-category is not relevant and only “bulk disposals”
are considered in this document. This is because the Environment Agency does not allow disposal of low
volume VLLW originating from a nuclear site via this route.
VLLW will be dealt with according to current UK practices (6) and subject to, for example, the VLLW
acceptance criteria for the LLWR (7). The main criteria are as noted above with specific restrictions being
determined through for example, the Waste Enquiry Process, as determined by the LLWR.
Only one VLLW stream has been identified for UK ABWR, Dry Active Waste, which is mixed waste that
will arise during reactor operations and decommissioning. The waste consists of contaminated personal
protection equipment, monitoring swabs, plastic, equipment, structures and contaminated plant.
These wastes will require specific removal, handling, sorting and size reduction techniques depending on
their physical form and characteristics prior to treatment. The preferred strategy for this mixed dry active
VLLW, produced during operations and during decommissioning, is to recycle the metallic portion where
practicable and to dispose of the remainder, following pre-compaction if possible, to permitted disposal
sites within the UK.
No credible alternative options have been identified for this waste stream at this time.
8.1.2. Low Level Waste
Most LLW in the UK today arises from the operation of nuclear power stations and nuclear fuel
reprocessing facilities, as well as the decommissioning and clean-up of nuclear sites. Operational LLW is
principally lightly contaminated miscellaneous waste, arising from maintenance and monitoring, such as
plastic, paper and metal. LLW from decommissioning comprises building materials and metal plant and
equipment. Most LLW from nuclear licensed sites is currently managed via the LLWR Ltd organisation,
which provides a service for the range of wastes produced, as described below, and where applicable
specific wastes are disposed of at the LLWR site near Drigg in Cumbria.
LLW is waste having a radioactive content not exceeding 4 GBq per tonne of alpha, or 12 GBq per tonne
of beta/gamma activity.
As an example, based upon current practices LLW can be dealt with in accordance with the strategy to
manage waste through the LLW Joint Waste Management Plan, which is agreed between the waste
producing site and LLWR Ltd. This organisation encourages an approach based upon the Waste Hierarchy
and a number of recycling options are offered for waste producers to use. Where solid wastes are produced
these can be disposed of to the LLWR, in accordance with their published acceptance criteria (8). LLWR
Ltd offers a ‘complete’ service to waste producers which is assumed to be the case for the GDA submission.
However, any future site operator may choose their own arrangements based on the site specific
requirements and the demonstration that they have selected the optimal disposal route at the time. The
service from LLWR Ltd comprises four main options, as detailed in (9) and briefly described below:
a. Metallic Waste Treatment Service
This service is a recycling option. Waste can be treated by decontamination, blasting or melting to
remove the radiological content. The vast majority of the metal can then be recycled as Exempt
Waste thus reducing the volume for disposal. The remaining Secondary Waste is either consigned
to the LLWR for disposal as LLW or VLLW, depending upon its activity, or disposed of by the
Service Supplier, depending upon contract arrangements between the waste producer, LLWR Ltd
and the Service Supplier.
b. Combustible Waste Treatment Service
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This service is a volume reduction option. Waste can be incinerated to reduce its volume. The
remaining Secondary Waste is either disposed of by the Service Supplier or consigned to the Low
Level Waste Repository for disposal as either LLW or VLLW, depending upon waste
characteristics and the contractual arrangements as noted above.
c. Supercompactable Waste Treatment Service
This service is a volume reduction option. Waste can be treated by shredding and / or high force
compaction to reduce its volume. The Secondary Waste product is consigned to the LLWR for
disposal.
d. Low Level Waste Disposal Service
This service is a disposal option. Waste that is not suitable or selected for treatment, has already
been treated or is Secondary Waste from a treatment process is consigned to the LLWR for
disposal. LLWR Ltd offers a range of standard disposal containers which are mostly centred on the
use of the half height ISO container. The containers can be filled with waste in various forms,
compliant with the applicable waste acceptance criteria, either at the waste producer’s site or at the
repository. For the containers consigned from the waste producer, after transport to the repository
the interspaces within the ISO container are filled with a cement grout prior to emplacement in the
repository trench.
A site operator will need to make business decisions, which will also include commercial and logistical
requirements which may indicate that there are other ways of dealing with specific wastes. For example,
the site operator may decide to contract directly with a waste Service Supplier rather than utilising the
LLWR contract arrangements. It is not envisaged that such alternative arrangements would change the
outcome as far as waste management was concerned.
One of the sources of decommissioning LLW in the UK ABWR will be from materials within the
controlled area, such as concrete.
For IX resins, incineration has been used for some sites in Japan however this approach is not being
developed for the UK ABWR GDA submission.
Six distinct LLW streams have been identified as arising during operations only and some additionally
during decommissioning. They are described briefly below together with the options available for
treatment, packaging and disposal. The UK practice generally is to consign packaged LLW to the LLWR
via the Joint Waste Management Plan, as noted above. Significant on site storage facilities are not
necessary but a facility for buffer storage, likely to be for approximately 2 years’ worth of LLW arisings,
will be provided to allow flexibility to manage consignments without adversely affecting generation rates,
which could otherwise cause bottlenecks. For each LLW stream, after accumulation of a number of drums
of packaged waste in the buffer store they will be consigned to the LLWR in accordance with the Joint
Waste Management Plan.
The following Sections briefly describe each stream and propose a preferred method for dealing with the
waste. This could include some recycling or other processes aimed at a reduction in the wastes which will
need to be disposed of to the LLWR, in accordance with the Waste Hierarchy. Where a preferred method is
identified this will be based upon current UK practice for the specific waste, where this is well established.
However, for site specific application, which will occur in several years’ time, when the plans for dealing
with actual wastes are developed, there may be other means available for dealing with specific streams.
Where this occurs, such alternative means will be subject to the applicable BAT assessments and other
optioneering requirements needed to develop the overall business case for implementation by the operator.
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1) HVAC Filters
These filters are installed in the exhaust outlets of various buildings, as listed in Table 8-1 and will
included a number of medium efficiency, bag type filters and several High Efficiency Particulate
Air (HEPA) filters. The detailed type and construction of these filters will depend upon the
operational and safety case requirements.
The construction of the medium efficiency filters usually comprise a metal frame with a bag type
glass fibre filter medium attached.
The construction of the HEPA filters proposed for the UK ABWR, as currently used in Japanese
reactors, comprises a wooden frame, glass fibre element and aluminium separators between
adjacent elements.
The design and application of each type of filter for the UK ABWR is yet to be decided and will
follow UK practice for these types of filters for both performance/safety requirements and
design/construction. The filters can be supplied in a range of shapes, sizes and materials, including
wood and metal, mostly with glass fibre media. It is envisaged that whatever materials of
construction are used the currently available techniques of compaction and cement grouting are
likely to be adequate for packaging of filters for disposal.
2) Bead Resin
These are used within demineraliser beds to remove soluble radioactive species from various
water circuits from the Condensate Demineraliser (CD), Low Conductivity Waste (LCW) and
High Conductivity Waste (HCW) systems. The precise resins used by any UK site operator will be
determined by the operator themselves; however the following are assumed for the GDA
submission. The resins used in the Japanese ABWR systems comprise a styrene divinylbenzene
copolymer matrix which is chosen for optimal removal of the specific radionuclides present. When
the resins are spent the beds are periodically discharged to buffer storage tanks to allow
accumulation in order to implement efficient periodic packaging campaigns.
Common practice in the UK is to store the spent resins under water for a period and then apply a
solidification process; hence it will be assumed that bead resins are solidified with polymer or a
suitable cement formulation in disposal drums, typically 200 litre.
3) Concentrates
The Concentrated Waste System (CONW) treats wastes arise from the HCW evaporators, which
comprises wastes with such a high conductivity because of particulate and dissolved species that
they are not suitable for demineralisation. The HCW streams arise from sources such as RW/B
sumps, Service Building (S/B) sumps, CD drains and other similar sources.
It has been determined that as cementation is common practice in the UK it is assumed to be
applied to the UK ABWR. Japanese practice for these wastes is to solidify with a suitable cement
formulation in disposal drums, typically 200 litre.
4) Miscellaneous Combustible
These wastes are generated through routine operations, maintenance and decommissioning in
radioactive areas. They consist mainly of contaminated personal protective equipment, CF and
LCW spent hollow fibre membrane, polyethylene (sheet, bag), paper, wood, cloth, rubber gloves,
turbine oil waste, spent bead resin, spent active carbon filter media.
However, it is expected that there will be no radioactive oils in this category because of the
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location of lubricated items and methods of operation of the turbines and other equipment where
lubricating oils are used (10).
The treatment options include incineration, followed by encapsulation in cement formulation and
disposal of ash, or compaction within a drum and direct disposal within a container suitable for the
disposal facility. This will be assumed to be undertaken via the LLWR Ltd services and recycling/
management options
5) Miscellaneous Non-combustible
These wastes are generated as above for the combustible streams, through routine operations,
maintenance and decommissioning in radioactive areas comprising such materials as metals,
concrete, lagging, glass etc. The metals category, during decommissioning, with consist mainly of
removed plant items and is included as a separate category below.
These wastes may well include some items which could be dealt with in ways other than direct
disposal and the requirements of the Waste Hierarchy will be applied. For example contaminated
metal items could be decontaminated and the metal recycled via a suitable route. LLWR Ltd
currently offers this as a service to waste producers and would be a viable route for this recycling.
Treatment options will be influenced by the quantities arising and during normal operations it is
likely that size reduction followed by grouting or compaction within a drum and direct disposal to
a suitable facility will be the best option for wastes requiring disposal.
6) Contaminated and Irradiated Metal and Concrete (Decommissioning)
These wastes will be generated during the decommissioning phase and will comprise large
volumes of metal and concrete items. Many will be very large and requiring size reduction. The
strategy for management of these wastes is described in Chapter 27 of the PCSR.
However, the means of dealing with these wastes arising in the LLW category will be consistent
with the Waste Hierarchy and (for example) the Joint Waste Management Plans developed with
LLWR Ltd. For some contaminated metals the option of decontamination and recycling will be
adopted whilst some irradiated metals may need to be packaged and disposed of to the repository.
Concrete is likely to be subject to decontamination and possible reuse as infill or aggregate for
other constructions whilst the removed contaminated portion or any activated portion (e.g.
reinforcing bar) would be disposed of to the repository.
Processes for dealing with these portions will largely comprise size reduction and packaging in a
suitable disposal container. As noted above the LLWR is largely based on the use of the half height
ISO container but the RWMD standard 2m or 4m box with appropriate shield thickness (11) &
(12) may also be applicable for some items which would also be grouted in place with a suitable
cement formulation. The use of this route and how much of the waste is consigned in this way will
be dependent upon the application of the Waste Hierarchy and whether some items may be
suitable for recycling or reuse via the (for example) LLWR waste management services.
8.1.3. Intermediate Level Waste
ILW has radioactivity levels that are higher than LLW but which do not generate enough heat to require
special storage or disposal facilities. However, like other radioactive waste, it needs to be contained to
protect workers and the public from the radiation and contamination. ILW includes metal items such as
fuel cladding and redundant reactor components, and sludges and resins from the treatment of radioactive
liquid effluents. The main waste types are described below and comprise:
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? Sludge, also referred to as crud, arising from filtration of water streams
? Powder ion exchange resins, arising from water stream filter/demineralisers. This stream will also
contain some sludge
? Activated metals arising from locations within the reactor which are subjected to irradiation
At this point in time (i.e. GDA) several viable options exist to manage radioactive waste over the longer
term in the UK. A site operator will make the final decision as to how they intend to manage their wastes;
however for GDA all the viable options have been described. In order to demonstrate that the waste
produced by the UK ABWR can be disposed of and managed in the UK regulatory system a single option
has been selected to enable more detailed studies and analysis. At this stage the option of cement
encapsulation (for solid items) and solidification (for wet/slurry wastes) into unshielded stainless steel
containers will be selected as the packaging method to be adopted for a disposability assessment by
RWMD.
The ILW will be dealt with according to the requirements of RWMD and will be subject to disposability
assessments during the GDA process. For later, site specific waste streams and packaging methods a Letter
of Compliance (LoC) process is used to obtain endorsement. The adoption of the noted solution for GDA
will not preclude a future operator from selecting another method, subject to completion of relevant
optioneering and BAT assessments, business case analysis and LoC assessments.
The physical, chemical and radionuclide properties of the wastes will be assessed by RWMD to determine
compatibility with the current GDF concept. This process has been developed for the arisings of legacy
wastes from the currently operating and decommissioning sites and it is envisaged will be applicable to
wastes arising from the UK ABWR, when site specific information is available.
The management of ILW streams generated from operating a single UK ABWR are likely to fall within a
number of currently applied processes, as discussed in the following Sections.
8.1.3.1. General Packaging Options
There are currently a range of different containers and associated packaging processes which could be used
for packaging these wastes in the UK. They fall into two broad categories:
1) Thin walled containers
These containers are usually manufactured from stainless steel within which a wasteform is
produced, either by addition of an immobilisation matrix or by intimate mixing of the waste,
where it is water based as a slurry, with a suitable cement based formulation to form a wasteform.
For these containers there would be a need for a shielded onsite store to shield against direct dose
arising from the packages. In this case the combination of container and wasteform produce a
package which complies with the specification requirements of RWMD (13) to ensure durability
and predictable evolution over the periods of on-site interim storage and eventual transport to and
disposal in a GDF. Alternative immobilisation matrices are also available for specific wastes; for
example certain polymers have been used for specific, often ‘niche’ or unusual wastes. For
transport, these packages will be overpacked in a shielded transport container to protect the public
from radiation dose, in compliance with the relevant transport regulations and the package from
the requirements of the hypothetical accident requirements which are defined within those
regulations.
2) Robust shielded waste containers
In these containers, the waste contents have been subject to simpler processes like size reduction
and drying without the introduction of an immobilising matrix. In this case the requirements for
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the package for interim storage, transport and disposal to a GDF are largely met by the container
alone with little credit being claimed for the wasteform inside. These containers would also
provide sufficient external shielding and integrity to comply with the relevant transport regulations
depending upon the properties of the radioactive contents. It is also envisaged that interim stores
for these packages would be much simpler, and hence cheaper, than for the other option, above, as
they would only need to be weather proof and secure, although for long periods. (14) provides a
description of how such packages could meet the requirements of the RWMD specification
requirements noted above.
For the purposes of GDA, either option would be considered as viable, subject to a satisfactory
disposability assessment by RWMD. A number of LoCs for packages comprising stainless steel containers
and cemented wasteforms have been issued by RWMD for a range of wastes which are directly analogous
to the ILW streams arising from the UK ABWR; hence adoption of this approach is based upon well
recognised and informed approaches. RWMD have also assessed two submissions for conceptual LoCs
using robust shielded containers (15) and (16). The assessments show that this type of package can be
made to be disposable provided certain specific underpinning information is submitted as summarised
below:
a. Substantiate the performance of the packages under the accident scenarios for GDF operations,
either demonstration of complete containment or provide refined Release Fraction values;
b. Understand the constraints on the water content of the waste and any associated risks due to
pressurisation of the containers;
c. Provide confidence that the functionality of the lid bolts can be maintained during storage prior to
transport, allowing the seals to be changed as necessary; and
d. Demonstrate that the requirements for Data Recording and Management Systems are understood.
Where wastes are packaged during the operational period they would then be stored on site in an ILW store,
pending the availability of a GDF. The store type (shielded or unshielded) would be determined in
conjunction with the choice of waste packaging option; whether the waste packages are unshielded or
self-shielded as noted in 1 & 2 above. Until a waste container is selected for the UK ABWR – the on-site
interim storage requirements are unknown. However, the storage requirements for the alternative package
types are briefly described in later Sections of this document.
8.1.3.2. Sludge (Crud)
Sludges are forms of wet ILW that result from the filtration of liquids, from fuel cooling pools, active
drains and from the settlement of process residues and corrosion products. They are treated in the UK
ABWR in the Spent Resin and Sludge system (SS) arising from backwashing of various filters from the
Condensate Filter facility (CF) and the Low Conductivity Waste (LCW) systems. The waste from the CF
system will be within the ILW category as it arises from the condensate system after the steam has passed
through the turbines. The particulate activity will comprise small quantities of corrosion product which
may have been carried over with the steam.
The LCW system collects wastes from various sources including the Reactor Building (R/B), Turbine
Building (T/B) and Radwaste Building (RW/B) drains, and processes them via filters, demineraliser and
sampling tanks. The sludge will arise from backwashing the filters.
The main options for packaging sludges are either to dry using heat and vacuum, and store within a robust
shielded container or solidification in a suitable cement formulation intimately mixed within a thin walled
stainless steel container to form a homogeneous monolithic wasteform. Typically for cemented wasteforms
3m
3
or 500 litre drums are applicable, using a ‘lost paddle’ mixing system. The latter method is common
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practice within the UK, although the drying method is being adopted by some waste producers (e.g.
Magnox Limited) as their analysis shows that significant savings in short term capital expenditure will
result because shielded stores will not need to be built. Hence, a site operator may wish to consider the
option of using robust shielded containers when evaluating their business case for ILW packaging and
storage.
8.1.3.3. Powder Ion Exchange Resins
The Reactor Water Clean-up System (CUW) and Fuel Pool Cooling Clean-up System (FPC) filter
demineralisers process the main reactor circuit feed water and the water used in the SF cooling pool. This
stream comprises the powder resins which are used for removal of dissolved radioactivity and will include
any particulate sludge which is filtered from the CUW and FPC systems comprising particles of metal
oxide (iron oxide and others) in the reactor water/fuel pool water. These resins are discharged, together
with any entrained particulate sludge, when the particulate capture capacity of the resin is exhausted.
The powder resins as used in Japan have a cross linked polystyrene matrix and hence it will be assumed
that a similar material will be used for the UK ABWR as this resin material type is similar to those
currently used for radioactive water clean-up in other UK reactor systems; for example, for clean-up of
cooling pool water in the Magnox reactors.
The two main options for packaging these streams is the same as for the sludges above, comprising drying
and storage within robust shielded containers or solidification in a suitable cement formulation intimately
mixed within a thin walled stainless steel container to form a homogeneous monolithic wasteform.
8.1.3.4. Higher Activity Metals
These wastes arise during operations primarily, from the Zircaloy channel boxes which surround each of
the fuel element bundles in order to contain the boiling water region. These items are approximately 4.3 m
long and 15 x 15 cm square and used to be removed from the fuel element assembly after a period of
cooling in the fuel pool. Currently, these items are stored with the spent fuel assemblies in the spent fuel
pool in Japan. The fuel element and channel box are later dispatched together for further storage prior to
reprocessing.
For the UK ABWR, for GDA, it will be assumed that the channel boxes will be consigned with the SF and
not removed from the fuel assembly on discharge from the reactor. As it will be assumed that the UK
practice is not to reprocess the SF, this option could be advantageous. It will eliminate any need to store
and process or package as ILW, with attendant benefits in reducing operator dose and saving cost and
environmental impact of any facilities needed for processing/packaging to be undertaken. Adoption of this
approach will be subject to a satisfactory disposability assessment by RWMD and future discussion with
the site operator.
The other main contribution to activated metals during the operational period are the control rods, which
are of cruciform shape metallic construction containing stainless steel tubes in each wing of the cruciform
filled with boron carbide powder. The use of Hafnium may also be employed, depending upon the control
rod design. This material has a larger neutron capture cross section and longer lifetime which may offer
advantages and reduce the quantity of waste generated from replacement and management of control rods.
They are approximately the same length as the channel boxes and the activity is currently assumed to be
the same as for the channel boxes. There is, however a smaller volume of these items arising than the
channel boxes, at 5 units/year, as the control rods remain in the reactor for several fuel cycles. Their
management will need to be discussed further with the site operator to determine the optimum approach
which satisfies an acceptable business case.
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Japanese practice for these items is also to store in a water filled bunker. For the UK ABWR the control
rods may be managed in a number of ways. The method of interim storage of these items prior to
packaging is to be decided as UK practice with gas reactors is to store activated metals in a dry, usually
shielded ‘vault’ facility. This will allow a period of decay to minimise dose to operatives and to simplify
facility requirements when packaging is eventually undertaken. The logistics of this will need further
detailed analysis by the site operator as it will also involve development of an acceptable business case for
interim raw waste storage and a packaging facility. For this option it will be assumed that the control rods
will be size reduced to cut to length and reduce voidage possibly by shearing, to minimise secondary waste
and compaction prior to loading into a suitable disposal container. The type of control rod to be used for
the UK ABWR has not yet been determined. However, if they incorporate sealed tubes containing Boron
Carbide as the neutron absorber the size reduction will need to be undertaken to avoid breaching them. If
Hafnium is used as the neutron absorber instead, then the implications of the presence of this material will
also need to be considered. The envisaged quantity of these items over the operational life is approximately
30ton; hence further consideration by the site operator will be necessary to develop an acceptable business
case for the optimum solution.
The main container options for packaging these items are stainless steel boxes of similar size (3 m
3
) to
those used for sludge/resin aqueous slurry solidification, 500 litre stainless steel drums or robust containers
which are self-shielded and are also qualified as transport containers. For the latter category there are a
number of ductile cast iron container (DCIC) products available (e.g. by Crofts
2
or GNS
3
) of different
sizes.
The items would be grouted in to the stainless steel containers but not in the robust DCIC type. The
standard RWMD 2 m and 4 m box may also be suitable for these activated items after a decay storage
period. These boxes are primarily designed for packaging of decommissioning wastes of Low Specific
Activity and hence it may be logical to store the control rods until end of operations and deal with them
during decommissioning. Within these containers the wastes may be grouted or un-grouted, subject to
disposability assessment advice from RWMD. The interim storage on site of the ‘raw’ items could be in a
dry shielded environment to allow some decay (probably in the region of one order of magnitude) prior to
packaging and disposal or under water storage could also be implemented.
A number of other items have been identified. These include various reactor core components from the
SRNM and LPRM systems. These items could be dealt with on a campaign basis during the operational
period either by grouting into a stainless steel container, as described above or loading into a robust
shielded container.
8.1.3.5. Irradiated Metal (Decommissioning)
Where irradiated metal arises during decommissioning, which is determined to be ILW, this can be dealt
with using any of the containers noted in Section 8.1.3.4, depending upon their physical (shape, size and
weight), chemical and radiological characteristics and how much processing (e.g. size reduction) is
necessary. As for other ILW streams the acceptability of the packages to the GDF will be determined via
the LoC process, as noted in Section 8.1.3.1 above.
8.1.3.6. Borderline Wastes
Where a specific waste stream is likely to be considered as ‘borderline’ in that it may be close to the LLW
and ILW categorisation boundary, these will be assessed using an agreed methodology. LLWR Ltd have
2
http://www.croftltd.com/about/news_items/Developement-of-shielded-packages.php
3
http://www.gns.de/language=en/4877
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published a guidance document (17) which contains a number of decision making factors for consideration.
This document has been developed in conjunction with RWMD and its application will be linked with the
LoC process and the waste acceptance criteria adopted by LLWR.
RWMD are currently considering several aspects related to borderline wastes which include:
? Use of decay storage for up to 300 years for some short lived radionuclides;
? Further development of the LLWR boundary wastes report (17) to better optimise the balance
between LLW and ILW;
? Specific consideration of Tritiated wastes and whether to decay store or remove the Tritium;
? Consideration, in conjunction with LLWR Ltd to manage the wastes via the Safety case rather than
purely the Waste Acceptance Criteria.
Developments in this area by LLWR Ltd and RWMD will be monitored, possibly by the site operator via
the relevant User Groups for the LLWR and the GDF, when waste management practices are being planned
for implementation, to ensure that the optimal route for specific waste streams are chosen.
8.2. Spent Fuel
The current assumption, based upon the Government White Paper ‘Meeting the Energy Challenge’ (18) for
the management of Spent Nuclear Fuel from new build is to treat it as a ‘HAW’ and not reprocess, as is
undertaken in other countries This document specifically states that ‘unless otherwise stated, references to
the Government position on waste refer to “higher activity waste”, which includes intermediate level waste
waste and spent fuel from any new nuclear power stations’. The current convention is to define HAW as
comprising all HLW and ILW, and a small fraction of LLW. Hence, the terms ‘HAW’ and ‘HAW and
spent fuel’ are likely to be used interchangeably. Therefore, following a period of cooling in the SF Pool
(SFP) the SF will be considered as a waste and will be stored and packaged appropriately and disposed of
once a GDF is available.
SF is produced only during the operational phase of the reactor. However, all of the SF stored in the fuel
cooling pool, after final shutdown, is assigned to the decommissioning phase. From a technical perspective
there is no difference between these two phases and all SF will be managed using the same strategy.
A small number of fuel assemblies may experience defects. The characteristics of these will be discussed
with RWMD as part of the disposability assessment and any specific measures to manage them agreed.
The main defects that are likely to occur during a 60 year operational lifetime are fuel rod defects caused
by debris fretting or pellet-cladding interaction, for which the mitigation technologies have been
introduced.
At the highest level there are several options available for the long term on-site storage of SF in the UK.
The following Sections describe the spent fuel characteristics from a UK ABWR and the options for
packaging, interim storage and disposal. These options are based upon good practice as currently applied
in other countries (including one option planned for implementation at Sizewell B) and which would be
considered as BAT at this time. However, the selection of the appropriate option by the operator, when
developing their plans for packaging, interim storage and disposal will be subject to the application of
BAT within their decision making process.
8.2.1. Fuel Characteristics
The GE14 fuel assembly consists of a fuel bundle (composed of fuel rods, water rods, spacers, and upper
and lower tieplates and channel fasteners), and a channel that surrounds the fuel bundle (Figure 8.2-1).
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This fuel type has 92 fuel rods (of which fourteen are partial length rods) in a 10x10 rod array and two
large central water rods. The fuel rods and water rods are spaced and supported by the upper and lower
tieplates with intermediate spacing provided by eight spacers. The upper and lower tieplates are fixed by
eight tie rods, which hold the fuel bundle together. The upper tieplate has a handle for transferring the fuel
assembly.
Figure 8.2-1 – GE14 Fuel Assembly
Based upon preliminary advice from RWMD the assumed cooling period before geological disposal is
likely to be of the order of 100 years. General information for fuel and plant operation are shown in Table
8.2-1 below. Potential fuel failure mechanisms are debris fretting or pellet-cladding interaction, for which
the mitigation technologies have been introduced.
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Table 8.2-1 General information for fuel and plant operation
Item Value
Power (MWt) 3926
Operational cycle (month) 18
Average discharged fuel burn-up (GWd/t) 50
Weight of UO2 per assembly (kg) 200*
Total number of fuel assemblies discharged over
60 years operation
9600*
*:approximate
8.2.2. Packaging Options
The SF assemblies will be stored in the SF pool in order that the heat output and activity levels can reduce.
The SF will then be transferred to another on site storage facility prior to it being re-packaged, if required,
for disposal in the GDF.
The option to store the channel boxes with the SF assemblies has been considered for minimising waste
volumes and avoiding the potential operator dose associated with removal of them and subsequent size
reduction. It is assumed for GDA that the channel boxes will remain with the SF and hence the
treatment/packaging of them will be the same as for the SF.
Management of any defective/failed fuel rods/assemblies is to be determined.
Options for the on-going management of the SF, will be subject to a satisfactory disposability assessment
by RWMD, and will include the following candidate packaging/storage options:
8.2.2.1. Dry Cask
This is current Japanese practice of removal of SF elements from the fuel cooling pool after an appropriate
period and loading a number of them into a large shielded cask, which is capable of storing the SF. This
cask could then be used for transport, after fitting of suitable high performance shock absorbers, to a
central interim storage site, if appropriate. For GDA it is assumed that the SF remains on site in a suitable
facility.
This method stores the spent nuclear fuel which has already been cooled in a fuel storage pool for a period
of time. The SF assemblies are sealed within a large leak-tight container and surrounded by an inert gas.
There are two types of cask system that can be used:
? Concrete – a number of SF elements are placed in a stainless steel canister which is welded shut
before being placed into a concrete outer casing;
? Metal – the SF elements are placed into a metal cask which is sealed and bolted shut.
Both systems are designed to provide protection against external hazards and provide adequate radiation
shielding to both the site workforce and members of the public. There are a number of dry storage
container designs in use around the world and different designs can be used for storage, transportation or
both. In some designs, the steel cylinders containing the fuel are placed vertically in a concrete cask; in
other designs the cylinders are placed horizontally. The concrete casks provide radiation shielding. Other
cask designs place the steel cylinder vertically on a concrete pad at a dry cask storage site and use either
metal or a combination of metal and concrete outer cylinders for radiation shielding.
SF assemblies would be loaded into the chosen cask system in the Reactor Building. The cask would then
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be sealed and decontaminated before being taken to the dry cask storage facility which would provide
weather protection.
The SF elements would need to be prepared for transport to and disposal in a GDF by removal from the
cask storage into another container which would be suitable for transport and acceptable for disposal. Such
a container may be as described in 8.2.2.4 below.
8.2.2.2. Multi-Purpose Container
To avoid the need for repackaging the SF into a disposal container, after a period of storage for cooling (for
example in a cask storage system, as described above) the concept of using a Multi-Purpose Container
(MPC) has been proposed. This would involve the packaging of a smaller number of SF elements into a
container which can be used for extended on site storage and subsequent disposal to a GDF.
There are systems available for this which are currently used in the USA. For example, the Holtec system,
similar to that shown in (19) has been proposed for Sizewell B PWR SF, which accommodates 24 elements
in a single MPC. Use of this container has been assessed by RWMD (20) which concluded that use of this
MPC was not optimum when taking into account all of the management stages to the GDF and that re
packaging for disposal may still be necessary. The current situation is that RWMD plan to propose an
MPC
4
which would be suitable for on-site interim storage and subsequent disposal when a GDF is
available. This MPC will then be considered for use with the UK ABWR SF and hence, use of this system
is considered to be a viable option, subject to the RWMD disposability assessment.
8.2.2.3. Modular Vault Dry Storage System
Dry vault storage is a method, already used in many other countries, for storing spent nuclear fuel that has
already been cooled in the SF pool. In this concept individual fuel assemblies are stored within sealed
channels within the shielded vault. Current examples of these vaults have been built in such a way that
additional storage vault capacity can be constructed when needed as new modules.
After removal from the reactor the SF assemblies would be loaded into a flask in the Reactor Building. The
flask would then be sealed and decontaminated before being taken to the vault store. Here, the fuel
assemblies would be lifted from the flask and dried. A shielded fuel handling machine would then transfer
the fuel assemblies to the allocated storage channel in the vault, which would then be filled with inert gas
and sealed.
The SF assemblies are stored vertically within the channels and cooled by natural circulation. This acts as a
self-regulating system in which higher SF temperatures cause increased airflow through the vault, thereby
increasing heat removal.
The SF elements would need to be prepared for transport to and disposal in a GDF by removal from the
store into another container which would be suitable for transport and acceptable for disposal. Such a
container may be as described in 8.2.2.4 below.
8.2.2.4. KBS-3 Container
The reference conceptual design assumption for spent UK PWR fuel is that it is packaged in a KBS-3V
type canister, as used by SKB in Sweden (21) and this is considered to be a viable option for UK ABWR
SF with suitable modifications to take account of the different SF element dimensions and other
characteristics.
For PWR SF use of this concept is currently subject to the assumed temperature constraint applied to the
4
RWMD plan to issue a report in due course which will be referenced in this document.
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inner bentonite buffer surface within the GDF and that a canister contains four SF assemblies. If all the SF
were to be subject to the maximum burn-up considered, which is 65 GWd/tU, then the SF might need of
the order of 100 years of cooling in interim storage before it could be disposed of in a GDF. This aspect
will be discussed with RWMD during the disposability assessment for the UK ABWR where it is likely
that a more realistic case for cooling time would be developed which may include (for example) a mixture
of SF elements within a single container whose burn-ups would not all be at the stated maximum.
This container may be used for disposal of SF elements after a period of interim storage in any one of the
previously described storage systems, unless the MPC system is one where the MPC itself would be
acceptable to RWMD for disposal to the GDF.
Hence, use of the KBS-3 container is considered to be a viable option, subject to the RWMD disposability
assessment.
8.2.2.5. On-Site Fuel Pool Option
On-Site Fuel Pool Storage within an independent building is a method, already used in many countries, for
storing spent fuel that has been stored in the SFP in the Reactor Building. On-Site Fuel Pool Storage is the
method that is adopted in UK-EPR and is a viable option to consider for UK ABWR. The characteristics
are that cooling performance is high and, the condition of the spent fuel can be confirmed by visual
inspection from above the water, which will provide the necessary shielding.
After the period of storage for cooling in the SFP in the Reactor Building, the spent fuel would be transfer
to a separate On-Site Fuel Pool by a transport cask. The spent fuel would be stored in racks in the SFP. The
spent fuel elements would need to be prepared for transport to and disposal in a GDF by removal from the
pool into another container for transport and the package would be made acceptable for disposal. Such a
container may be as described in para. 8.2.2.4 above.
The use of this option will require cooling and clean-up equipment for managing the SF pool water quality.
8.3. Summary of Management Options
The available management options for each waste category and the SF are summarised in Table 8.3-1
below. These are based upon existing practices within the UK for legacy sites (e.g. Magnox, Sellafield,
Dounreay) and some operational sites (e.g. Sizewell B) and from international experience, especially for
the management of SF. The options have not been selected by a rigorous BAT process but are assumed to
be so at this stage of GDA because of their current applications, as noted above.
Further details of each option are included in Appendix D.
Table 8.3-1 Summary of Waste and Spent Fuel Streams
No. Title Category Form Management options
1 Dry active
waste
VLLW Solid Recycle metals where practicable, compaction,
where possible and direct disposal of remainder
to permitted disposal site
2 HVAC Filters LLW Solid To be determined, depending upon filter details
utilising a combination of compaction and
cement grouting
3 Bead resin LLW Wet Solidify with polymer or cement formulation in
disposal drum, typically 200 litre
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No. Title Category Form Management options
4 Concentrates LLW Wet Solidify with cement formulation in disposal
drum, typically 200 litre
5 Miscellaneous
combustible
LLW Solid ? Incineration (off site)
? Compaction within a drum
6 Miscellaneous
non-combustibl
e
LLW Solid ? Size reduce and grout within a disposal
drum
? Compaction within a drum
7 Sludge (crud) ILW Wet ? Dry and store within robust shielded
container
? Encapsulate in cement formulation
? Encapsulate in polymer formulation
8 Powder resin ILW Wet ? Dry and store within robust shielded
container
? Encapsulate in cement formulation
? Encapsulate in polymer formulation
9 Higher activity
metals –
control rods
ILW Solid
Store on site in dry shielded facility OR
Size reduce and:
? Dry and store within robust shielded
container
? Place in disposal container (typically 3m
3
box) and grout with cement formulation
? Channel boxes stay with SF
10 Higher activity
metals –
channel boxes
ILW Solid
11 Higher activity
metals – others
ILW Solid
12 Contaminated
and irradiated
metal and
concrete
LLW Solid
? Size reduce and place in disposal
container (typically 2m or 4m box) and
grout with cement formulation
? Decommissioning ILW only – may also use
robust shielded containers; DCIC or 4 m/2
m box
13 Irradiated metal
(from
decommissioni
ng)
ILW
Solid
14 Spent Fuel HLW Solid ? Dry cask storage with future packaging for
disposal
? Multi-Purpose Container storage for future
disposal without repackaging
? Dry vault storage ( Modular Vault Dry
Store)
? Package and store in SKB type container
? Extended storage in on-site fuel pool
9. Waste Management Infrastructure
A number of facilities will be required to conduct the various activities of waste management, conditioning,
packaging and storage of each of the waste and SF categories described above. The detailed descriptions of
each facility will depend, to some extent upon the conditioning and/or packaging method chosen for each
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specific waste stream.
The infrastructure requirements will be developed by Hitachi-GE, including consultation with Horizon,
during the initial phase of GDA so that any initial site specific requirements can be captured.
Waste treatment/packaging and storage requirements will be described for each stream of waste and SF
comprising:
9.1. Waste Treatment and Packaging Facilities
9.1.1. Low Level Waste
The LLW facilities are likely to be the same as is currently employed in Japan, subject to evaluation in
detail for compliance with UK practices and disposal criteria.
Aqueous slurry form wastes, such as ion exchange resins may be packaged in a facility, which can mix the
wastes with a suitable cement formulation to form a monolithic product of adequate mechanical properties
within a 200 litre drum. These drums would then be loaded into half height ISO containers for transport to
the disposal facility where the remainder of the space within the container would be grouted, prior to
emplacement within the repository.
Solid waste which required grouting within a drum, for example solid and metal objects which are not able
to be recycled or reused, may be packaged within a facility which can prepare a suitable grout infiltration
matrix (usually cement based), which can be delivered to each drum to form a monolithic product of
adequate mechanical properties.
These facilities are likely to be fixed, as is the current practice in Japan to ensure that dedicated plant is
available when required.
Where other processes, such as compaction or incineration are required for conditioning the waste it is
assumed that the facilities will be available off site, for example via the LLWR Ltd supplier services. In
this case a facility for drumming wastes and loading into full height, reusable ISO containers will be
provided so that they can be filled at site and dispatched to the disposal facility or a service supplier to
conduct the required processes.
9.1.2. Intermediate Level Waste
The facilities for treatment and packaging the ILW streams will depend upon the type of package chosen
from the two main options described in 8.1.3.1 above. For each package type the following are the main
facility requirements, which are described in more detail in Appendix D:
1) If thin walled stainless steel containers are to be used for immobilisation of the wastes in a cement
grout or formulation the shielded facility would need the following:
a. A facility which connects the container to the waste accumulation facility to allow retrieval
and loading the container with waste;
b. A means of introducing either a cement grout (for solid wastes) or a cement powder blend
(for aqueous slurry wastes) to encapsulate or solidify, respectively the solid or slurry waste;
c. An area for product curing to allow the wasteform to achieve adequate strength for
movement to the ILW store;
d. A transfer route to the ILW store either via a shielded transport vehicle or a shielded route
where the packaging facility is adjacent to the store.
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2) If robust shielded containers are to be used for direct packaging the following will be required:
a. A shielded facility which connects the container to the waste accumulation facility to allow
retrieval and loading the container with waste;
b. A means of size reduction (for solid wastes) or pump transfer (for aqueous slurry wastes) and
loading of the wastes into the containers;
c. A means of drying the wastes which may be a process external to the container for solid
wastes or in situ for the aqueous slurry waste. For the latter, standard vacuum drying
equipment is available which requires in situ heating of the waste whilst pulling a vacuum
and extracting the off gas via a suitable vent system, which may need a means of preventing
any volatile or soluble activity from entering the environment;
d. A transfer route to the ILW store which will be a simple standard transport vehicle capable of
carrying the load, but with no shielding required as this is supplied by the container.
9.1.3. Spent Fuel
The facility requirements for processing/packaging of SF will depend upon the management option chosen.
Whatever option is chosen there will be a common process required to dry the items. For the dry cask and
MPC options above this would need to be undertaken before loading into the relevant container, whilst for
the dry vault storage option drying would be undertaken after transport to the storage facility.
Apart from a drying facility, no other major dedicated facilities would be required; the standard reactor
design is expected to be adequate as it is assumed to contain equipment for loading SF elements into a cask
or other container.
At the end of interim storage, the current plan assumes that the SF would be removed from the dry storage
casks and transferred into final disposal containers. Such an operation, many decades after reactor closure,
would require a dry transfer cell facility to move spent fuel assemblies from the interim dry storage casks
to the final disposal containers.
At the time such a plant is needed, the site should have been fully decommissioned apart from the dry cask
spent fuel facility, and the ILW store.
9.2. Waste Stores
9.2.1. Low Level Waste
In order to accumulate a reasonable number of packages to make a consignment to the disposal facility, an
on-site buffer store will be required. The size of the store has been selected as two years’ worth of LLW
arisings to allow for disruption to the transport system or problems at the disposal facility which could
restrict or curtail the acceptance of waste consignments. The store will be located within the existing
Radioactive Waste treatment Building (RW/B) or other storage facilities and will be connected to the main
HVAC, utilities, drainage and other systems. Hence, for GDA it is assumed that there are no new facility
requirements, discharge points or additional doses arising from storage requirements for LLW.
9.2.2. Intermediate Level Waste
At this point in time Hitachi-GE are yet to determine the most appropriate method for storing the ILW on
the Generic Site for the purpose of GDA. However, the ILW will be stored on site in facilities, whose
characteristics will be suited to the specific packages/containers which are stored in them; for example
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whether they need to be shielded or unshielded. Such stores will be built to best UK practice standards and
guidance to ensure minimal impacts in respect of discharges and doses to the environment.
If the packages utilise thin walled stainless steel containers the store will need to be built to provide
adequate shielding and include all of the facilities necessary for handling the packages and for monitoring
their evolution over time. The use of remotely operated equipment is likely to be required for these
operations. If self-shielded containers are used (e.g. DCICs) then the quantity of shielding will be
significantly reduced. However, the store will require adequate facilities for handling the packages and
monitoring their evolution, although this is likely to be conducted by simpler, ‘hands-on’ techniques
because the dose rates will be designed with such in mind and taking account of ALARP considerations.
The NDA have produced a Guidance document for Packaged ILW stores which is directed at the NDA’s
legacy sites (22). This guidance could also be applicable where similar stores are necessary on the UK
ABWR sites.
The guidance includes:
? An integrated approach to the overall storage system by considering package and store
performance as a whole;
? Identification of good practice, principles, approaches and the use of ‘toolkits’ for specific
analyses; e.g. package evolution and performance models;
? Requirements for package care and management, identifying ideal, tolerable and failing
performance parameters.
9.2.3. Spent Fuel
The store arrangement for SF will depend upon the option chosen for its management, as noted above. This
facility will need to take account of the container/cask size and arrangements within the store, cooling and
shielding requirements and criticality and other safety requirements.
The SF assemblies would initially be kept in the existing fuel pool for a period to be determined to allow
them to cool down before being loaded into the chosen container system. The container would then be
sealed before being taken to the new storage building. The SF may need to be re-packaged for disposal
depending upon the option chosen for interim storage.
For the dry cask option the storage location would consist of a receiving area with rail/road-connection, a
cask maintenance platform (e.g. for seal replacement) and one or more cask storage areas, each equipped
with an overhead crane. Prefabricated concrete structures, in combination with reinforced poured concrete,
are flexible in terms of extension. Operation of the store would be totally passive and the facility would be
able to operate totally independently of the reactor site (after shutdown).
For each of the SF management options discussed in Section 8.2.2 the following are the main requirements
for long term storage:
1) Dry Cask
These casks would be capable of storing a number of different inner containers. The storage
requirements for these casks are likely to consist only of a weather proof facility which includes
adequate security and the ability to inspect the external surfaces of the casks. Further details of the
type of facility required are to be discussed at a later stage.
2) Multi-Purpose Container (MPC)
These containers are intended to be utilised for interim storage and eventual disposal and could hence
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be stored within the above noted dry casks for an interim period. The size of the MPCs will need to be
optimised to accommodate a sufficient number of SF elements whilst also allowing for eventual
transport and disposal at the GDF. As discussed in Section 8.2.2.2 above RWMD are currently
preparing some proposals on this aspect. The interim storage requirements, however, will be the same
as for the Dry Casks, as noted above.
3) Modular Vault Dry Storage System
This system is a ‘stand-alone’ store where the SF elements are stored in sealed containers within a
naturally cooled vault system, as described in Section 8.2.2.3 above. This store would require its own
facilities such as ventilation, security, discharge points and utility connections.
4) KBS-3 Container
This container is similar to the MPC described above with the exception that it accommodates a
smaller number of SF elements and hence its disposability is likely to be acceptable to RWMD,
subject to their assessments. As for the MPC these containers would be stored in Casks, for which the
interim storage requirements are as described above.
5) Extended Storage in On-site Fuel Pool
During the operational period the usual storage arrangements will apply. After the plant has ceased
generating storage in this separate facility will be an extension of the arrangements in place during the
operational period.
It is noted that Sizewell B has commenced building a dry SF store, which is due to be operational in 2015.
Hence, the experience gained from this project is likely to be of benefit, in terms of learning from
experience, if similar stores are to be built for the UK ABWR.
9.3. Waste Packages, Transport Systems and LoCs
RWMD have developed and published a number of specifications and guidance documents (13) which
define the requirements for a range of standard waste packages. These are largely based upon the use of
thin walled stainless steel containers within which a wasteform is created by combination of the waste
either with a cement powder blend for water based slurries or infiltration of a cement grout for solid waste
items. These specifications are based upon a GDF design and waste packaging proposals submitted by the
waste producers to RWMD are assessed against this design using a well-developed LoC process. This
process has three stages for package endorsement comprising ‘conceptual’, ‘interim’ and ‘final’ stages; the
last stage providing evidence of disposability to the ONR to enable them to grant a Licence Instrument to
the waste producer for implementation of waste packaging. These packages would need to be transported
within a reusable shielded transport container. This combination is usually classed as a Type B transport
package under the IAEA Transport Regulations.
As a number of waste producers have chosen to utilise the ‘robust shielded waste package’ concept, which
proposes the use of ductile cast iron containers for packaging wastes, where size reduction and drying are
used instead of waste immobilisation in cement, RWMD have also issued a Technical Note (14) which
provides guidance as to their use and acceptability in a GDF. As well as being disposal packages, these are
designed to qualify as transport packages in their own right without the need for additional outer packaging
to provide radiation shielding. They will be classed as Type A or Type B depending upon which container
is used and the characteristics of the contents.
In order to assess waste packaging arrangements proposed during the GDA process RWMD are using a
‘pre-conceptual’ assessment process, based upon the LoC process, which is designed to provide a
Disposability Assessment which is sufficient for the GDA process to show that waste packages generated
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by the subject design will be disposable.
10. Development of an Integrated Waste Strategy
The requirements for an Integrated Waste Strategy (IWS) have been developed by the NDA (23) for the
legacy wastes where its applicability is directed at operational sites or those undergoing decommissioning.
A successful IWS document will provide the following:
? Clearly identify what wastes need to be managed (both radioactive and non-radioactive);
? Describe how they are managed now and in the future, taking account of the full waste lifecycle;
? Demonstrate how the strategy is delivering against national policy and strategy;
? Sign post to key underpinning and justification information in an effective and accessible way
(including UK Radioactive Waste Inventory, TBURDs, etc.);
? Identify future problems and/or gaps, and the solutions to address them;
? Drive improvement by focusing on key outcomes;
? Show how the strategy will be implemented and how it factors into business decisions;
? Be an integral part of meeting regulatory requirements for waste management;
? Be available and accessible to people.
All of the above requirements will not be applicable at the GDA stage and those which are relevant are
incorporated within this arrangements document (for example much of the first two bullets above). These
arrangements do, however allow for an IWS to be subsequently developed at a later stage, for an
operational site, which will then be subject to periodic review and update.
11. Plan for Disposability Assessments
11.1. Introduction to the RWMD Disposability Assessment
Hitachi-GE is required to obtain a view from RWMD as to whether the HAW can be disposed of in line
with any plans for a GDF in the UK. As part of obtaining that view Hitachi-GE need to provide RWMD
with a suite of information to allow them to undertake their assessments. At this moment in time the
assessments are underway therefore no details are provided on the conclusion. The next issue of this
document will include such information.
The interactions with RWMD have been planned to achieve the submission of a disposability assessment
(by RWMD) to the Environment Agency, to comply with the requirements for Justification of the use of
the ABWR technology in the UK.
The information to be provided to RWMD will cover all HAW and SF for which there is currently no
disposal facility available and, for the latter where reprocessing is no longer to be undertaken. The HAW
will comprise Intermediate Level Wastes (ILW) as it is assumed that there will not be any vitrified HLW or
LLW which is unsuitable for disposal at the Low Level Waste Repository.
All relevant information required by RWMD will enable their assessment of the envisaged packages to be
conducted for suitability for disposal in a GDF. At this stage several options for packaging the SF for
interim storage and eventual disposal will be included because the preferred method has not yet been
defined. The final choice will involve many other aspects related to specific site requirements, including
interim storage arrangements, the need for long term cooling prior to disposal, etc.
For the purposes of the submission of information to RWMD, a number of assumptions will be included.
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11.2. Hitachi-GE Plan for the RWMD Disposability Assessment
The information to be provided to RWMD will include the following:
? Details of the characteristics of the waste and any proposed conditioning processed, (e.g. drying,
cementation, size reduction etc.) will be provided. These details will be supported by relevant
references from Hitachi-GE, which will include the physical, chemical and radionuclide properties
envisaged for the wastes together with quantities envisaged to arise during operation and from
future decommissioning activities;
? Specific parameters of the SF and any associated packaging which will be relevant to its condition
at the time of disposal will be addressed. The relevant parameters required to inform the
disposability assessment are also included;
? A number of packaging and interim storage options are described using a number of standard
containers, as currently used in the UK and known to RWMD. The means whereby each of the
wastes and SF may be rendered into a passively safe wasteform, within a disposable package, will
be outlined in concept, based upon current UK practice and experience of similar wastes and SF
arising on other UK sites;
? The ABWR ILWs will be compared to other UK ILWs, which have already been assessed and
found to be disposable by issue of final stage Letters of Compliance (LoC) to support the case that
the envisaged waste packages will be disposable.
In order to ensure that the information required by RWMD is as complete as possible and made available
in a timely manner to enable a full assessment to be completed, regular interactions during the provision of
the submission information and the subsequent assessment period will be undertaken.
The interactions between RWMD and Hitachi-GE/Horizon will include how their project addressing the
disposability and full life cycle implications of high-heat generating UK wastes, which includes a wide
range of UK wastes, is developing and how the approach for UK ABWR may be influenced.
A number of assumptions are included in the plan to focus the scope of the disposability assessment to
address GDA requirements for the current phase of the project.
12. Conclusions
This arrangements document sets out management arrangements for processing, interim storage and where
facilities are available the disposal of waste and SF generated by the UK ABWR in accordance with the
UK Government policy and regulatory constraints. The UK ABWR provides a high degree of confidence
that the challenges associated with the management of solid waste and SF are fully understood and that
solutions are available with the envelope of current UK and international experience. Specifically it
demonstrates that:
? The Waste Hierarchy will be applied and the production of radioactive waste will be avoided and
where this is not reasonably practicable, the quantity of waste produced will be minimised;
? The radioactive wastes generated by the UK ABWR are similar of those wastes generated by
operating ABWRs and all waste streams have process routes to interim storage or final disposal
solutions;
? A number of waste and SF management options which are based on nationally or internationally
proven technologies and the experience from ABWR projects are available for the UK ABWR.
The base case waste treatment facilities for each UK ABWR site include:
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? Waste Treatment Building for the receipt, segregation, treatment and conditioning of solid
radioactive wastes;
? Interim Storage Facility for solid ILW packages;
? Interim Storage Facility for receipt, packaging and safe interim storage of SF.
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Appendix A – Consideration of the REPs
The following is a list of the REPs listed in (2) which are considered relevant to Radioactive Waste
Management Arrangements & Disposability. Each REP has been addressed in turn and a comment added
to indicate where it has been considered in this arrangements document.
Table A-1 Compliance of the Document with the REPs
REP Description Comment
RSMDP1 Radioactive Substances Strategy This document
RSMDP5 Actions having Irreversible
Consequences
This document outlines the thorough and
considered philosophy underpinning all
radioactive waste management decisions. The
application of BAT will ensure a robust decision
making process
RSMDP6 Application of BAT The proposed practices detailed in this
document are considered to be BAT and the
decision making process for future
implementation will be informed by the
application of BAT
RSMDP8 Segregation of Wastes Covered in the arrangement of waste
accumulation facilities and that each waste is
packaged separately with no mixing. The
application of the Waste Hierarchy is
emphasised throughout the document
RSMDP9 Characterisation Covered in principles and packaging options to
ensure adequate knowledge of the waste
properties and that the most appropriate
management technique can be applied
RSMDP10 Storage Passive safe storage for packaged waste is
covered as a principle and mentioned
throughout the document
RSMDP11 Storage in a Passively Safe State Concentrating waste and relevant conditioning
technology to deliver a solid waste form is
outlined. The requirements for passive storage
and adequate management of waste packages
are covered.
RSMDP14 Record Keeping Requirements are covered for all radioactive
wastes e.g. RWMD requirement for packaged
HAWs.
RSMDP15 Requirements and Conditions for
Disposal of Wastes
Reference is made to WACs for VLLW and
LLW managed via LLWR Ltd and LoCs for
ILW managed via RWMD
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REP Description Comment
ENDP4 Environmental Protection Functions
and Measures
The defined management approaches for waste
conditioning and storage and consignment to
other facilities will ensure adequate
environmental protection
ENDP8 Ageing and Degradation Not explicitly covered for plant systems.
However, adoption of BAT will lead to
implementation of this. The RWMD LoC
process, which is referred to, requires the
longevity of waste packages and wasteforms
with predictable evolution
ENDP 15 Mechanical Containment Systems for
Liquids and Gases
This is applicable to raw waste accumulation
facilities which are referred to. Adoption of
BAT will lead to implementation of this for
waste packaging
ENDP16 Ventilation Systems The HVAC filters are included as the solid
waste stream which will arise from these
systems.
DEDP1 Decommissioning Strategy Waste management aspects only are covered.
Decommissioning Strategy is covered in a
separate document
DEDP2 Decommissioning Plan Waste management aspects only covered.
Decommissioning Plan is covered in a separate
document
DEDP3 Considering Decommissioning during
Design and Operation
Waste management aspects only covered.
Considerations during design and operation is
covered in a separate document
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Appendix B - Waste Hierarchy
The Waste Hierarchy as illustrated in Figure B1 below aims to encourage the management of waste
materials in order to reduce the amount of waste materials produced, and to recover maximum value from
the wastes that are produced. It is not applied as a strict hierarchy as many complex factors influence the
optimal management for any given waste material. However, as a guide, it encourages the prevention of
waste, followed by the reuse and refurbishment of goods, then value recovery through recycling.
The next option is energy recovery, an important level in the hierarchy as many materials have significant
embedded energy that can be recovered. Waste prevention, reuse, recycling and recovery are collectively
defined by the Organisation for Economic Co-operation and Development (OECD) as waste minimisation.
Finally, waste disposal should only be used when no option further up the hierarchy is possible.
Figure B-1 - Waste Hierarchy
5
5
SEPA Briefing Note - Ozone Depleting Substances (ODS) and the Construction Industry; Version 1; July
2008
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Appendix C - Waste Stream and Spent Fuel Descriptions
1. Introduction
This Appendix provides currently available details of the wastes and SF which it is envisaged will be
generated during the operational life of the UK ABWR and the details of the assumed decommissioning
wastes which will arise at the end of life. Where additional details are required this is shown as a To Be
Advised (TBA) item.
2. Nature and Quantity of the Wastes and Spent Fuel
A number of reports have been produced which describe the wastes envisaged to arise from the UK ABWR.
Specifically (24) provides a description of the various waste treatment systems currently used in Japan and
which are likely to form the basis of systems for use in the UK. (25) provides a brief summary of all of the
solid wastes which are envisaged to arise from operation and decommissioning of the UK ABWR. The
arisings of spent filters from various HVAC systems are also included in this document.
(26) has been produced to provide more realistic inventories which are relevant to waste management and
disposal. The data within this report excludes a number of radionuclides whose half-lives are too short to be
relevant for short term LLW or much longer term ILW disposal. The report identifies 19 radionuclides
which are considered relevant for L2 disposal in Japan. As this category is approximately equivalent to the
lower activity range specified in the UK for ILW this list of radionuclides will be used for the initial
inventories for LLW and ILW for the GDA.
The following tables summarise the information for each category of waste envisaged to arise from the UK
ABWR during its operational life (24), (25) and (26) and from decommissioning activities (25).
Information is also included for the SF arisings during the operational period.
The information provided for activities has been updated in line with (26) to include the main radionuclides
contributing to the activities. This information is included in the following summary tables. The activity
levels have also been adjusted to include only the radionuclides with a half-life greater than 3 months.
The wastes are categorized as concentrated liquid waste (by an evaporator), waste sludge from filters, spent
resins from the demineralisers and miscellaneous solid wastes which are activated or contaminated.
Further general information of the various systems within the UK ABWR is similar to that described in
(27).
Liquids and gases which are discharged after abatement are not included in this document. Wastes which
can be treated and disposed of to current licensed disposal facilities (e.g. the LLWR or licensed landfill site)
are included. Wastes which currently have no available disposal facility, which comprise ILW and SF
(assumed to be waste and not subject to reprocessing) are included, which will be consigned to the GDF
when it is available.
2.1. Operational VLLW
This category currently includes very few items and it is likely that further will arise during operations. Any
items arising will be managed according to the Waste Hierarchy and it is currently assumed that the
services offered by LLWR Ltd will be employed to optimise the management of the wastes and any
disposals necessary.
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Table C2.1-1 Details of VLLW Miscellaneous Solid Items
Parameter Description
Waste Origin Arising from operations and maintenance
Waste physical/chemical description Low activity piping, motors and heat insulators from
maintenance activities
Key parameters for conditioning Solid items
Nature of radioactive material Low levels of contamination
Main radionuclides/activity No details available
Annual operational arising No details available
Total operational arising No details available
Hazardous substances None
2.2. Operational LLW
Note that the LLW limit is 1.20E+04 Bq/g for beta/gamma and 4.00E+03 Bq/g alpha activities and that the
following activities are for un-processed/packaged wastes. Depending upon the packaging technique used
the stated specific activities for wastes, to consign for disposal, may decrease; for example where the best
technique for packaging is to mix with a cement formulation. Where a technique is chosen which may
concentrate the waste, and hence the total activity may move into the ILW range this factor will be taken
into account when choosing the best technique.
The following details are based upon wastes arising from current Japanese sites and there may be some
difference for operational plants in the UK. For example the number and type of HVAC filters may be
different for UK ABWR, depending upon safety analysis and the criteria which are used for filter
replacement. The values used are based upon Japanese experience, where the recommendations of the filter
manufacturer are used. This will be a bounding case for the number of filters arising.
Table C2.2-1 Details of LLW HVAC Filters
Parameter Description
Waste Origin Arising from filter changing in air treatment facilities:
? Reactor Building exhaust
? Turbine Building exhaust
? Turbine Building high radiation exhaust
? Radwaste Building exhaust
? Service Building exhaust
Waste physical/chemical description Medium efficiency filters (bag type) - 6.3 kg/filter
HEPA filters – 13 kg/filter
Key parameters for conditioning Solid items comprising filter media and support frames
Nature of radioactive material Low levels of contamination
Main radionuclides/activity (Bq/g) Co-60, Ni-63, Fe-55, Ag-110m; 1.3E+03
Annual operational arising Medium efficiency filters – 1.9 ton/year
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Parameter Description
HEPA filters – 0.3 ton/year
Total operational arising Medium efficiency filters – 114 ton
HEPA filters – 18 ton)
Hazardous substances None
Table C2.2-2 Details of LLW Bead Resins
Parameter Description
Waste Origin Arising from the CD, the LCW and HCW demineraliser
Systems. The CD removes radionuclides from the main
reactor coolant water after is has passed through the
condenser. The origin of the LCW is as described above. The
HCW comprises wastes with a higher particulate content and
clean-up systems may include filtration, evaporators and
demineralisers. Bead resins are used in the demineralisers to
extract soluble radionuclides prior to reuse or discharge of the
liquors
Waste physical/chemical description LCW & HCW Demineraliser; Cation Exchange Bead Resin,
Anion Exchange Bead Resin. These resins have a cross linked
polystyrene matrix
CD; Cation Exchange Bead Resin Anion Exchange Bead
Resin. These resins have a Styrene divinylbenzene copolymer
matrix
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, incineration, compaction
Nature of radioactive material Soluble species arising from the noted systems; including
soluble activated corrosion products and fission products
Main radionuclides/activity (Bq/g) CD – Co-60, Ag-110m; 3.8E+03
LCW demineraliser – Co-60, Mn-54, Ag-110m; 2.6E+04;
HCW demineraliser – H-3, C-14; 7.9E+02
Annual operational arising 14.5 m
3
/year
Total operational arising ~880 m
3
/60 years
Hazardous substances None
Table C2.2-3 Details of LLW Concentrated Waste
Parameter Description
Waste Origin Arising from the HCW evaporator and stored in the CONW
tanks for a period of decay before processing
Waste physical/chemical description Concentrates from various sources containing organic and
inorganic particulate and soluble species
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, compaction
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Parameter Description
Nature of radioactive material Particulate and soluble species arising from the noted systems
Main radionuclides/activity (Bq/g) H-3, Co-60, Ag-110m; 1.3E+02
Annual operational arising 2.4 m
3
/year
6
Total operational arising ~150 m
3
/60 years
Hazardous substances None
Table C2.2-4 Details of LLW Miscellaneous Combustible Waste
Parameter Description
Waste Origin Dry active waste generated through routine and maintenance
operations. The waste consists of CF and LCW spent hollow
fiber filter membrane, polythene (sheet and bag), paper, wood,
cloth (wipes and gloves), rubber gloves, spent active carbon
Waste physical/chemical description Heterogeneous non-metallic solid dry wastes, apart from the
oil waste
Key parameters for conditioning All wastes could be incinerated, if appropriate. A significant
proportion would be compactible, if that route was required
Nature of radioactive material Particulate contamination
Main radionuclides/activity TBA
Annual operational arising CF spent hollow fibre media; 2.2 m
3
/year
LCW spent hollow fibre media; 0.06 m
3
/year
Spent active carbon; depend on maintenance
Miscellaneous; depend on maintenance
Total operational arising TBA
Hazardous substances None
Table C2.2-5 Details of LLW Miscellaneous Non-Combustible Waste
Parameter Description
Waste Origin Arises during maintenance operations within the nuclear
island. The waste mainly consists of contaminated RCA
plant. Comprising metal plate, pipes, bulk, cables, lagging
material, gas filters, concrete and glass
Waste physical/chemical description Metallic and other solids
Key parameters for conditioning Dry largely non-compactible
Nature of radioactive material Particulate contamination
Main radionuclides/activity (Bq/g) Co-60, Ag-110m; 3.0E+03
Annual operational arising Depend on maintenance and replacement
6
Design basis value; actual close to zero
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Parameter Description
Total operational arising TBA
Hazardous substances None
2.3. Operational ILW
The following streams are those whose activities are envisaged to be greater than the LLW activity limits of
1.20E+04 Bq/g for beta/gamma and 4.00E+03 Bq/g alpha and are not heat generating. The stated activities
are for the raw unprocessed wastes and the specific activities will change depending upon the packaging
technique used and may also be influenced by radioactive decay of some radionuclides, in particular those
which have a short half-life. For GDA it will be assumed that the categories will remain the same for
conditioned and unconditioned wastes. However, if during site operations some ILW streams are
re-categorised to LLW the application of relevant LLW management techniques will ensure that the streams
are adequately dealt with. No issues are envisaged with being able to treat any ILW stream as LLW, subject
to the noted activity limits.
Table C2.3-1 Details of ILW Filter Sludge (Crud)
Parameter Description
Waste Origin Arising from backwashing of CF and LCW filters. LCW
arises from the drain sumps of the various components in the
R/B, the drywell, the T/B and the RW/B. It may also be
collected in low-conductivity waste collecting pools
Waste physical/chemical description Corrosion product Fe2O3, Fe3O4, FeOOH
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, compaction
Nature of radioactive material Corrosion product particulate slurry and some dissolved
species
Main radionuclides/activity (Bq/g) CF – Co-60, Mn-54, Fe-55, Zn-65; 1.1E+05
LCW filters - Co-60, Mn-54, Fe-55, Zn-65, Ag-110m;
4.5E+05
Annual operational arising 1.9 m
3
/year
Total operational arising ~114 m
3
/60 years
Hazardous substances None
Table C2.3-2 Details of ILW Powder Resin
Parameter Description
Waste Origin Arising from the CUW and FPC filter demineralisers.
Waste physical/chemical description CUW Cation Exchange Powder Resin. ANION Exchange
Powder Resin. These resins have a cross linked polystyrene
matrix.
This stream also includes any particulate sludge which is
filtered from the CUW and FPC systems
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Parameter Description
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, incineration, compaction
Nature of radioactive material Soluble species arising from the noted systems; including
particulate and soluble activated corrosion products and
fission products
Main radionuclides/activity (Bq/g) CUW filter demineraliser – Co-60, Mn-54, Fe-55, Zn-65,
Ag-110m; 3.6E+07
FPC filter demineraliser – Co-60, Mn-54, Fe-55, Zn-65,
Ag-110m; 1.3E+07
Annual operational arising 4.9 m
3
/year
Total operational arising ~300 m
3
/60 years
Hazardous substances None
Table C2.3-3 Details of ILW Higher Activity Metals – Control Rods
Parameter Description
Waste Origin Cruciform shape and inserted between each group of 4 fuel
elements external to the channel boxes and hence within
non-boiling water. Performs the dual functions of power
distribution shaping and reactivity control
Waste physical/chemical description Metallic construction. The control rods may contain Boron
carbide or Hafnium as alternative materials for reactivity
control
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual operational arising 5 units; (0.5 ton; approx. 1.5 m
3
)/year
Total operational arising 30 ton (approx. 90 m
3
)/60 years
Hazardous substances None
Table C2.3-4 Details of ILW Higher Activity Metals – Channel Boxes
Parameter Description
Waste Origin Channels which contain each fuel element within the reactor
core to direct the coolant flow and contain the boiling regions
Waste physical/chemical description Zircaloy box which surrounds the fuel bundle. Approx. 4.3 m
long and 15 x 15 cm square
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual operational arising Approx. 170 units; (6.8 ton; approx. 34 m
3
)/year
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Parameter Description
Total operational arising 408 ton/60 years
Hazardous substances None
Table C2.3-5 Details of ILW Higher Activity Metals – Reactor Components, etc
Parameter Description
Waste Origin Various reactor core components.
Waste physical/chemical description Start-up Range Neutron Monitoring system (SRNM including
dry tubes), Local Power range Monitoring System (LPRM),
Traversing In-core Probes (TIP) and neutron source units
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual/periodic operational arising LPRM – approx. 1.8 ton (LPRM: 0.5 ton, Basket: 1.3 ton)
SRNM – approx. 0.9 ton/10years (SRNM: 0.3 ton, Basket: 0.6
ton)
TIP – approx. 0.3 ton/10-20 years (TIP: 0.05 ton, Basket: 0.25
ton)
Neutron source unit – 0.3 ton/life (Unit: 0.05 ton, Basket: 0.25
ton)
Total operational arising Approx. 116 ton/60 years
Hazardous substances None
2.4. Decommissioning Non-Radioactive Waste
The following waste details have been included for completeness. However, as they are not radioactive
wastes they will not be addressed in detail.
Table C2.4-1 Details of Non-Radioactive Decommissioning Wastes
Parameter Description
Waste Origin Metal and concrete
Waste physical/chemical description Metal and concrete
Key parameters for conditioning Large items needing size reduction for disposal
Total arising 632,900 ton
Hazardous substances TBA
2.5. Decommissioning VLLW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
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Table C2.5-1 Details of VLLW Decommissioning Wastes
Parameter Description
Waste Origin Contaminated and irradiated metal and concrete including
arisings from decontamination of some of the LLW. Arising
from various sources during decommissioning
Waste physical/chemical description Metal and concrete
Key parameters for conditioning Large items needing size reduction for disposal to licensed
disposal site
Nature of radioactive material Largely contaminated with some activation of metallic items
Main radionuclides/activity TBA
Total arising 25,983 ton
Hazardous substances TBA
2.6. Decommissioning LLW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
Table C2.6-1 Details of LLW Decommissioning Metal and Concrete Wastes
Parameter Description
Waste Origin Contaminated and irradiated metal and concrete. Arising from
various sources during decommissioning
Waste physical/chemical description Metallic and concrete
Key parameters for conditioning Large items needing size reduction for disposal to Low Level
Waste Repository
Nature of radioactive material TBA
Main radionuclides/activity TBA
Total arising 8,660 ton
Hazardous substances TBA
Table C2.6-2 Details of LLW Decommissioning Process Wastes
Parameter Description
LCW Demineraliser Bead Resin 80 m
3
arising over 20 years from operations ceasing
HCW Demineraliser Bead Resin 18 m
3
arising over 20 years from operations ceasing
HCW Evaporator Concentrated
Waste
48 m
3
arising over 20 years from operations ceasing
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2.7. Decommissioning ILW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
Table C2.7-1 Details of ILW Decommissioning Irradiated Metals
Parameter Description
Waste Origin Reactor core components and from areas subject to activation.
Waste physical/chemical description Activated metal components
Key parameters for conditioning Dry, metallic, non-combustible. Some large items needing size
reduction for packaging and disposal
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) 1.30E+04~2.00E+10
Total arising 920 ton
Hazardous substances None
Table C2.7-2 Details of ILW Decommissioning Process Wastes
Parameter Description
CUW Filter Demineraliser Powder
resin/crud
17 m
3
arising over 5 years from operations ceasing
FPC Filter Demineraliser Powder
resin/crud
7.5 m
3
arising over 5 years from operations ceasing
LCW Filter Crud 12 m
3
arising over 20 years from operations ceasing
2.8. Spent Fuel
The following table summarises the general information currently available on the SF. (28) indicates the
detailed parameters which are required for GDA, and which will be needed in due course. This information
is currently on hold, pending the availability of an export licence. The table below also summarises specific
information which RWMD have requested to support the disposability assessment.
Table C2.8-1 Details of Spent Fuel Assemblies
Parameter Description
Waste Origin Uranium dioxide fuel pellets within Zircaloy cladding form
the fuel assemblies, which undergo fission and produce heat.
A proportion of this heat is ultimately converted to electricity.
The fission products and actinides produced during the fission
process are considered as waste. In addition to the fission
products and actinides, the structures of the fuel assemblies
are activated and at discharge from the reactor contain
activation products
Waste physical/chemical description The reactor core consists of fuel rods held in bundles by
spacer grids and top and bottom fittings. The fuel rods consist
of uranium dioxide pellets stacked in a cladding tube, plugged
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Parameter Description
and seal welded. Fuel element design is GE14. Materials of
construction are Zircaloy, stainless steel, Inconel, ceramic
UO
2
and Gd
2
O
3
Specific information required by
RWMD
1. How the SF is believed to evolve during interim storage
period. Initially this may be based upon current Japanese
practice, but may need to be modified in the future if UK
interim storage conditions are different;
2. SF data comprising a full radionuclide inventory of the
fuel and any non-fuel core components. To include initial
enrichment, burn up (and variability), any clad
impurities, burnable poisons. 1 year cooled data would
be helpful and provision of heat and A2 information if
available;
3. Details of Uranium type and specifically if any
reprocessed material was used as this can affect the
U236 profile;
4. Variability’s in burn up, for example a max core burn up
value may include variability’s for individual fuel
elements as a result of fuel shuffling;
5. Operational strategy for the core to provide
understanding of the burn up variations;
6. Information on the strategy to be adopted for failed fuel.
Key parameters for conditioning Dry, metallic, non-combustible, heat generating. Approx. 4.5
m long
Nature of radioactive material Fission products, activation products and actinides
Annual discharge from core 150 assemblies (approximate)
Total discharge from core 9600 assemblies (approximate)
Hazardous substances None
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Appendix D - Descriptions of Conditioning Options
1. Introduction
This Appendix provides detailed concepts for a number of options for management of the radioactive
wastes (VLLW, LLW and ILW) and the SF. The Environment Agency has requested that this should be
provided as a part of the GDA submission; hence the following is proposed to meet this requirement.
It should be noted that at this stage all available options are not listed and the options shown below are
current, commonly used techniques which are considered to be BAT and are included as viable options for
GDA. They are currently in use or planned to be used on specific legacy sites (e.g. Magnox Limited) and at
some operating sites (e.g. Sizewell B) and hence considered to be viable for inclusion.
For VLLW and LLW (as shown in Figure D1) the use of the LLWR is assumed as an example as there are
other facilities available, which have not been considered at this stage. For the ILW streams disposal to a
UK national GDF is assumed as, currently, this is the only proposed disposal route.
At the appropriate time, for specific waste streams all available options will be considered as part of a BAT
analysis. At this time there may also be other alternative technologies which are not available today and
have been developed to sufficient maturity for application.
The following diagrams and notes describe the processes for dealing with each of the waste stream and SF
categories which are discussed in the main document and where details are available in Appendix C.
The options comprise:
1 Very and Low Level Wastes
2 Intermediate Level Waste; Solidification of Sludge/Crud and IX Resins
3 Intermediate Level Waste; Encapsulation of Activated Metals
4 Intermediate Level Waste; Drying of Sludge/Crud and IX Resins
5 Intermediate Level Waste; Drying of Activated Metals
6 Spent Fuel management
Each of the above options is described for each of the process steps. As some of the steps are common to a
number of options the descriptions have been repeated for completeness so that the description of each
option will be complete.
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Waste
Accumulation
Slag/
secondary
waste
Sort
Decontamination/
Size reduce
Retrieve
Dry,
Compactable
(non-
incinerable)
Metallic
Combustible
Characterisation
and application of
the Waste
Hierarchy
VLLW
Dry, non-compactable &
non- incinerable, e.g. soils
& concrete. Solidified IX
resins, sludges and
concentrates in cement
matrix
Metal
Recycling
Consign to
VLLW Disposal
Facility
Combustible
Waste
Treatment
Super
compaction
VLLW
Disposal
Consign to
Recycling
Facility
Load into
suitable
transport
container
LLW Disposal
Load into
suitable
transport
container
Ash residues
Load into
suitable
transport
container
LLW Disposal
LLW Disposal
LLW Disposal
Load into
suitable
transport
container
Load into
suitable
transport
container
C
o
n
s
i
g
n
v
i
a
L
L
W
R
S
e
r
v
i
c
e
s
C
o
n
t
r
a
c
t
B
u
f
f
e
r
S
t
o
r
e
o
n
s
i
t
e
i
f
r
e
q
u
i
r
e
d
4
3
2
1
10
11
5
6
7
8
9
11
11
13
14
15
12
11
11
Figure D1 – Very and Low Level Wastes
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Notes for Figure D1 – Very and Low Level Wastes
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles.
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste item is dealt with in
an optimal manner according to its category and to enable the provision of adequate data for the
applicable waste route; for example to enable adequate justification for compliance with the applicable
Waste Acceptance Criteria (e.g.) of the LLWR.
3 The waste items are retrieved from their accumulation facilities using appropriate standard UK
practices. Solid waste items would be retrieved using tools and grabs in facilities which would mostly
be manually operated and able to ensure adequate protection to an operator from direct radiation dose
or contamination. Retrieval would be into a sorting facility or may be directly into a suitable container,
depending upon the waste item, for subsequent treatment.
4 Where waste items are mixed, a facility for sorting the waste will enable the streams to be segregated
and directed to the most appropriate process for subsequent treatment in accordance with the Waste
Hierarchy.
5 Dry, Compactable (non-incinerable) wastes will include heterogeneous items such items as those
which arise during maintenance operations within the nuclear island. The waste mainly consists of
contaminated RCA plant. Comprising metal plate, pipes, bulk, cables, lagging material, gas filters,
concrete and glass. HVAC filters (medium efficiency and HEPA type) may also be included in this
category.
6 Metallic wastes may arise from the above stream and would be segregated during the sorting process
7 Combustible wastes will comprise dry active waste generated through routine and maintenance
operations. The waste consists of CF and LCW spent hollow fibre filter membrane, polythene (sheet
and bag), paper, wood, cloth (wipes and gloves), rubber gloves, spent bead resin, spent active carbon.
8 VLLW items will arise from operations and maintenance and include such items as low activity piping,
motors and heat insulators from maintenance activities
9 Dry, non-compactable & non-incinerable wastes will include such items as soils & concrete and other
similar items which will be segregated from other wastes, as described above. This category will also
include the wet LLW streams which arise, such as ion exchange resins, sludges and concentrates
which will be rendered as dry and non-compactible, typically by mixing with a suitable solidification
medium (e.g. a cement powder formulation) in a 200 litre disposal drum. There are a number of
existing wastes from UK legacy sites which have previously been solidified in this manner and hence
a suitable formulation will be based upon these.
10 Where applicable to specific waste streams, decontamination and size reduce will be undertaken. This
step would typically be applicable where metals are prepared for subsequent melting and recycling or
where other items (e.g. concrete) may be decontaminated to allow for the bulk of the waste to be
recycled as aggregate or to be disposed of to a permitted land fill site rather than to the LLWR. Size
reduction may also be necessary solely to ensure that the waste items may be suitable for packaging in
a standard container such as a 200 litre drum.
11 When the waste is in a suitable form (e.g. via solidification of wet wastes, size reduction etc.) they will
be loaded into a suitable transport container. This may comprise the standard half height ISO container,
which is specified by and used at the LLWR. In this case, after transport to the LLWR, the container
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will be infiltrated with a cement grout and allowed to cure prior to emplacement within the LLWR
vault. LLWR also specify a number of other standard containers which may be used for transport of
wastes for subsequent treatment at the LLWR or at one of its suppliers where other processes are
applicable for treatment of the waste; for example super compaction or incineration.
12 The wastes, as accumulated into suitable transport containers, may be buffer stored on site for a period
of time. This may be required to comply with the Joint Waste Management Plan, agreed between the
site operator and LLWR which would include a schedule of waste deliveries to LLWR for onward
processing or direct disposal, depending upon the waste type.
13 It is assumed for the GDA submission that all VLLW and LLW will be consigned via a LLWR
Services Contract and will utilise the well-established facilities available for dealing with the wide
range of wastes specified. This will ensure that the Waste Hierarchy is adequately applied in directing
wastes for treatment via the most appropriate route to ensure that the quantities of waste consigned for
disposal in the LLWR vaults is minimised. Hence, the inputs from the buffer storage area will be dealt
with in a manner appropriate to the waste, which could be any one of the 5 outputs shown.
14 Slag secondary waste residues which arise from metal recycling will be treated according to their
characteristics. Their radionuclide content will be carefully determined to ensure that ‘concentration’
effect from the melting process has not rendered this waste as ILW. In the event that it has the residues
will be treated as ILW and its processing with follow the appropriate route. Typically, these LLW
residues may be packaged in a similar manner to wet wastes as identified above for ion exchange
resins, sludges and concentrates. In this case the residues may either be mixed with a prepared cement
grout in a mixing drum or by mixing the residues with water and adding a suitable cement blend
within a mixing drum, typically 200 litres. A process and formulation for solidification of ash and
dusts has previously been developed for UK legacy sites and hence these secondary wastes will be
dealt with using a similar process.
15 The ash residues arising from the incineration process can also be dealt with in a similar manner to the
above noted for the slag secondary residues. The comment for the slag residues, where the
‘concentration’ effect of incineration may have rendered the waste as ILW will also apply to this waste
stream.
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Figure D2 – Intermediate Level Waste; Solidification of Sludge/Crud and IX Resins
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Transfer to
buffer storage
as required
Pre treatment
Hydraulic
transfer
Solidify in
cement blend
formulation in
lost paddle
container
Cure, cap add
lid, inspect and
complete
package
Transfer to
shielded store
Monitored
storage period
Transfer to
GDF
1
2
3 4 5
6
7
8
9 10
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Notes for Figure D2 – Intermediate Level Waste; Solidification of Sludge/
Crud and IX Resins
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (25) as Sludge (Crud) and Powder Resin (which also contains particulate corrosion
product).
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specifications and guidance
as published by RWMD (13) to adequately inform the LoC process such that a LoC can be issued in
due course, subject to RWMD’s disposability assessment. The characterisation data may be obtained at
an appropriate point in the process where the waste stream is homogenised by a mixing/recirculation
technique to ensure that samples which are representative of the bulk of the stream are obtained.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. This technique will utilise equipment which is installed within the accumulation facility
and may comprise mixers to ensure that the wastes are homogeneous and suction devices which are
capable of extraction of the waste stream slurry at the solids concentration which would be determined
by the characterisation information.
4 Buffer storage of a batch of waste may be necessary to ensure that the subsequent process is
adequately controlled, or this facility may be required to allow for a waste solidification campaign to
be undertaken when a sufficient quantity of waste has been accumulated. This will then allow further
quantities of waste to be added to the accumulation facility to ensure that normal operations are not
held up. This buffer storage facility may also be the point at which the sampling is undertaken to
obtain the characterisation data noted in step 2 above.
5 Transfer of the waste stream from the buffer tank to the waste packaging process will use an
appropriate UK standard technique for delivering a measured quantity of waste of defined properties
to the waste container. The control of the transfer equipment will be integrated with the downstream
equipment to ensure that waste delivery complies with the overall process to produce a predictable
waste package (defined as the combination of the disposal container and the wasteform produced
within it) in accordance with the requirements of the LoC. In this case the overall integrity of the
waste package is mostly provided by the properties of the cemented contents with a minor
contribution claimed for the container.
6 A pretreatment step may be necessary to adjust and/or optimise the properties of the waste stream to
adjust such properties as (for example) the solids concentration or the pH. This is to ensure
compatibility with the packaging requirements and solidification formulation which has been
demonstrated to RWMD via the LoC process to produce satisfactory, disposable, wasteform properties.
The pretreatment step will be conducted either in the buffer tank or disposal container depending on
what is required and the waste properties. For example, reducing the water content of the IX resin
stream can be undertaken via a submerged filter within the disposal container whereas to reduce water
content within a sludge stream may require some external equipment comprising such equipment as a
settling tank or filter system.
7 Solidification of the waste stream will be conducted within a UK standard container, which is
compliant with the appropriate RWMD specification. These containers will be manufactured from
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austenitic stainless steel (typical grade 316L or EN 1.4404 to BS EN 10088-2:2005) and incorporate
an integral mixing paddle which will remain with the wasteform once it has been generated within the
container. Typical container sizes, as used in the UK for many legacy site wastes, include 500 litre and
3 m
3
drums.
The cement blend and quantity added will be as defined for the specific waste stream. This will be
based upon formulations as used in Japan for these wastes or as for similar wastes, which have arisen
at legacy sites and have been solidified in the UK. The formulation will have been previously
endorsed by RWMD via the LoC process as producing a wasteform of adequate properties, in
compliance with the relevant specifications, as noted above. The long term evolution of the wasteform
will also have been demonstrated. Both of these aspects will have been demonstrated by suitable
supporting development trials with inactive simulants or based upon directly applicable supporting
data from (for example) Japanese or UK practice.
8 After production and curing of the wasteform within the disposal container the whole package is
completed by addition of an inactive grout cap to minimise the residual voidage within the package
and fitting of the container lid by a sealed, bolted closure. The package will then be inspected via
visual means and swabbing the external surface to ensure that any contamination is within the limits
required for onward transport and interim storage on site.
9 The packages will be transferred to an on-site ILW store using a suitable shielded route. Depending
upon the site layout this may be via a shielded transporter or a shielded route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and will include addressing of
the guidance published for ILW stores for UK legacy sites (22).
This document provides guidance for an integrated approach to the overall storage system by
considering package and store performance as a whole; identification of good practice, principles,
approaches and the use of ‘toolkits’ for specific analyses; e.g. package evolution and performance
models; requirements for package care and management, including monitoring, identifying ideal,
tolerable and failing performance parameters. Where monitoring indicates deterioration, such that
some packages may not be suitable for eventual disposal, suitable ‘reworking’ schemes will be
developed which may, in exceptional circumstances require re packaging of specific waste packages.
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Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Figure D3 – Intermediate Level Waste; Encapsulation of Activated Metals
Size reduce
Add grout to
infiltrate waste
Load into
disposal
container
Cure, cap add
lid, inspect and
complete
package
Transfer to
shielded store
Monitored
storage period
Transfer to
GDF
8
7
6
5 4
3
2
1 9
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Notes for Figure D3 – Intermediate Level Waste; Encapsulation of
Activated Metals
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (24) as Higher Activity Metals; Control Rods, Channel Boxes
7
& Others. The method of
accumulation in Japan is within a ‘wet bunker’ where the items will also undergo size reduction prior
to placing in containers and grouting. The means of accumulation for UK ABWR is to be determined
and may follow UK practice for gas reactor legacy sites where such items are accumulated in dry
shielded vault stores.
2 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. The method of retrieval will be determined by the requirements of the overall process and
will be significantly influenced by how the items are to be accumulated (i.e. in a wet or a dry facility)
and where it is appropriate to conduct subsequent operations, especially size reduction.
3 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specifications and guidance
as published by RWMD (13) to adequately inform the LoC process such that an LoC can be issued in
due course, subject to RWMD’s disposability assessment. The characterisation data may be obtained at
an appropriate point in the process. For example, a gamma assay station may be incorporated within
the facility where the items are loaded into a disposal container.
4 The waste items will be size reduced in a suitable facility which will ensure that the operations can be
conducted according to ALARP principles. The design of the facility will depend upon the
accumulation method chosen. Where accumulation is in a wet bunker the necessary shielding will be
provided by the depth of water. In this case underwater techniques will be used to size reduce. Where
the items have been accumulated in a dry storage environment the necessary shielding will be
provided by a suitable shielded cell or the size reduction may be undertaken within the dry storage
facility. The size reduction techniques will comprise cutting (in a manner to limit secondary waste; e.g.
shearing) and compaction. The latter technique will take into account the need for adequate grout
infiltration during the subsequent packaging operation.
5 The items will be loaded into a suitable UK standard container, which is compliant with the
appropriate RWMD specification. These containers will be manufactured from austenitic stainless
steel (typical grade 316L or EN 1.4404 to BS EN 10088-2:2005). Typical container sizes, as used in
the UK for many legacy site wastes, include 500 litre drums and 3m
3
boxes.
6 Encapsulation of the waste stream will be conducted within the container by addition of a suitable
cement grout to produce a predictable waste package (defined as the combination of the disposal
container and the wasteform produced within it) in accordance with the requirements of the LoC. In
this case the overall integrity of the waste package is mostly provided by the properties of the
cemented contents with a minor contribution claimed for the container.
The cement blend and quantity added will be as defined for the specific waste stream. This will be
based upon formulations as used in Japan for these wastes or as for similar wastes, which have arisen
7
Note it is assumed for GDA that these items remaining with the spent fuel assemblies and are managed/packaged
as part of this stream
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at legacy sites and have been encapsulated in the UK. The formulation will have been previously
endorsed by RWMD via the LoC process as producing a wasteform of adequate properties, in
compliance with the relevant specifications, as noted above. The long term evolution of the wasteform
will also have been demonstrated. Both of these aspects will have been demonstrated by suitable
supporting development trials with inactive simulants or based upon directly applicable supporting
data from (for example) Japanese or UK practice.
7 After production and curing of the wasteform within the disposal container the whole package is
completed by addition of an inactive grout cap to minimise the residual voidage within the package
and fitting of the container lid by a sealed, bolted closure. The package will then be inspected via
visual means and swabbing the external surface to ensure that any contamination is within the limits
required for onward transport and interim storage on site.
8 The packages will be transferred to an on-site ILW store using a suitable shielded route. Depending
upon the site layout this may be via a shielded transporter or a shielded route/tunnel which is part of
the site infrastructure.
9 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and will include addressing of
the guidance published for ILW stores for UK legacy sites (22).
This document provides guidance for an integrated approach to the overall storage system by
considering package and store performance as a whole; identification of good practice, principles,
approaches and the use of ‘toolkits’ for specific analyses; e.g. package evolution and performance
models; requirements for package care and management, including monitoring, identifying ideal,
tolerable and failing performance parameters. Where monitoring indicates deterioration, such that
some packages may not be suitable for eventual disposal, suitable ‘reworking’ schemes will be
developed which may, in exceptional circumstances require re packaging of specific waste packages.
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Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud and IX Resins
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Transfer to
buffer storage
as required
Hydraulic
transfer into
robust
shielded
container
Fit bolted lid,
test seals,
swab &
monitor
Transfer to
drying plant
Retest
closures
where
required, swab
& monitor
Transfer to
intermediate
storage facility
Monitored
storage period
Transfer to
GDF
1
8
7
6
5
4 3
2
9
10
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Notes for Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud
and IX Resins
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (24) as Sludge (Crud), Powder resin (which also contains particulate corrosion product).
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specification as published
by RWMD (14) to adequately inform the LoC process such that a LoC can be issued in due course,
subject to RWMD’s disposability assessment. The characterisation data may be obtained at an
appropriate point in the process where the waste stream is homogenised by a mixing/recirculation
technique to ensure that samples which are representative of the bulk of the stream are obtained.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. This technique will utilise equipment which is installed within the accumulation facility
and may comprise mixers to ensure that the wastes are homogeneous and suction devices which are
capable of extraction of the waste stream slurry at the solids concentration which would be determined
by the characterisation information.
4 Buffer storage of a batch of waste may be necessary to ensure that the subsequent process is
adequately controlled, or this facility may be required to allow for a waste packaging campaign to be
undertaken when a sufficient quantity of waste has been accumulated. This will then allow further
quantities of waste to be added to the accumulation facility to ensure that normal operations are not
held up. This buffer storage facility may also be the point at which the sampling is undertaken to
obtain the characterisation data noted in step 2 above.
5 Transfer of the waste stream from the buffer tank to the waste packaging process will use an
appropriate UK standard technique for delivering a measured quantity of waste of defined properties
to the waste container. The control of the transfer equipment will be integrated with the downstream
equipment to ensure that waste delivery complies with the overall process to produce a predictable
waste package (defined as the combination of the disposal container and the wasteform produced
within it) in accordance with the requirements of the LoC. In this case the overall integrity of the
waste package is provided by the robust container and none would be claimed for the waste contents.
The robust container would be one of a number of designs currently being considered within the UK
by some legacy sites (e.g. Magnox Limited) and at the Sizewell B PWR. They are manufactured from
ductile cast iron, a material which is readily formed into the required shape by casting/machining and
whose properties are enhanced by the metal composition and controlled cooling to increase the
material’s ductility. The currently available designs can be supplied by Gesellschaft für
Nuklear-Service mbH (GNS - http://www.gns.de/language=en/4877) and Croft
(http://www.croftltd.com/).
6 Depending upon the layout of the retrieval equipment and the rest of the process plant the container
may then be sealed at this point by addition of a sealed lid secured by bolts. If the containers location
was in a contamination controlled area the container’s external surfaces would also be swabbed and
monitored prior to transfer to later stages in the process to ensure that contamination levels are within
allowable limits.
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7 The waste within the container will be dried by removal of bulk water and as much interstitial
moisture as required in order to achieve the required dryness level. This process will require
penetrations to be made into the container for the application of a vacuum, and some external heating
will also be applied to achieve this dryness. The process plant will include appropriate monitoring
equipment to measure the residual moisture content and hence determine an end point for the drying
process.
8 After the required dryness level is achieved the process equipment will be disconnected and the
container resealed by insertion of appropriate closures, which will be tested for leak tightness. The
container’s external surfaces would again be swabbed and monitored to ensure that contamination
levels are within allowable limits.
9 The packages will be transferred to an on-site ILW interim storage facility using a suitable route.
Depending upon the site layout this may be via a suitable transporter or a route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and may include addressing of
the guidance published for ILW stores for UK legacy sites (22). However, the applicability of this
guidance will be reviewed as it was developed by the NDA for shielded stores which are intended for
the long term interim storage of unshielded ILW packages, mostly comprising cemented wasteform
within thin walled stainless steel containers.
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Figure D5 – Intermediate Level Waste; Drying of Activated Metals
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Size reduce
and sort as
necessary
Transfer into
robust
shielded
container
Fit bolted lid,
test seals,
swab &
monitor
Transfer to
drying plant
Retest
closures
where
required, swab
& monitor
Transfer to
intermediate
storage facility
Monitored
storage period
Transfer to
GDF
1
8
7
6
5
4
3
2
9
10
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Notes for Figure D5 – Intermediate Level Waste; Drying of Activated
Metals
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. Higher Activity
Metals; Control Rods, Channel Boxes
8
& Others. The method of accumulation in Japan is within a
‘wet bunker’ where the items will also undergo size reduction prior to placing in containers and
grouting. The means of accumulation for UK ABWR is to be determined and may follow UK practice
for gas reactor legacy sites where such items are accumulated in dry shielded vault stores.
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specification as published
by RWMD (14) to adequately inform the LoC process such that an LoC can be issued in due course,
subject to RWMD’s disposability assessment. The characterisation data may be obtained at an
appropriate point in the process. For example, a gamma assay station may be incorporated within the
facility where the items are loaded into a disposal container.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. The method of retrieval will be determined by the requirements of the overall process and
will be significantly influenced by how the items are to be accumulated (i.e. in a wet or a dry facility)
and where it is appropriate to conduct subsequent operations, especially size reduction.
4 The waste items will be size reduced in a suitable facility which will ensure that the operations can be
conducted according to ALARP principles. The design of the facility will depend upon the
accumulation method chosen. Where accumulation is in a wet bunker the necessary shielding will be
provided by the depth of water. In this case underwater techniques will be used to size reduce. Where
the items have been accumulated in a dry storage environment the necessary shielding will be
provided by a suitable shielded cell or the size reduction may be undertaken within the dry storage
facility. The size reduction techniques will comprise cutting (in a manner to limit secondary waste; e.g.
shearing) and compaction.
5 Transfer of the waste stream to the waste packaging process will use an appropriate UK standard
technique for delivering a measured quantity of waste of defined properties to the waste container.
At this stage the waste may be characterised by gamma assay or similar. The control of the transfer
equipment will be integrated with the downstream equipment to ensure that waste delivery complies
with the overall process to produce a predictable waste package (defined as the combination of the
disposal container and the wasteform produced within it) in accordance with the requirements of the
LoC. In this case the overall integrity of the waste package is provided by the robust container and
none would be claimed for the waste contents.
The robust container would be one of a number of designs currently being considered within the UK
by some legacy sites (e.g. Magnox Limited) and at the Sizewell B PWR. They are manufactured from
ductile cast iron, a material which is readily formed into the required shape by casting/machining and
whose properties are enhanced by the metal composition and controlled cooling to increase the
material’s ductility. The currently available designs can be supplied by Gesellschaft für
8
Note it is assumed for GDA that these items remaining with the spent fuel assemblies and are managed/packaged
as part of this stream
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Nuklear-Service mbH (GNS - http://www.gns.de/language=en/4877) and Croft
(http://www.croftltd.com/).
6 Depending upon the layout of the retrieval equipment and the rest of the process plant the container
may then be sealed at this point by addition of a sealed lid secured by bolts. If the containers location
was in a contamination controlled area the container’s external surfaces would also be swabbed and
monitored prior to transfer to later stages in the process to ensure that contamination levels are within
allowable limits.
7 The waste within the container will be dried by removal of any residual bulk water and as much
interstitial/adherent moisture as required in order to achieve the required dryness level. This process
will require penetrations to be made into the container for the application of a vacuum, and some
external heating will also be applied to achieve this dryness. The process plant will include appropriate
monitoring equipment to measure the residual moisture content and hence determine an end point for
the drying process.
8 After the required dryness level is achieved the process equipment will be disconnected and the
container resealed by insertion of appropriate closures, which will be tested for leak tightness. The
container’s external surfaces would again be swabbed and monitored to ensure that contamination
levels are within allowable limits.
9 The packages will be transferred to an on-site ILW interim storage facility using a suitable route.
Depending upon the site layout this may be via a suitable transporter or a route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and may include addressing of
the guidance published for ILW stores for UK legacy sites (22). However, the applicability of this
guidance will be reviewed as it was developed by the NDA for shielded stores which are intended for
the long term interim storage of unshielded ILW packages, mostly comprising cemented wasteform
within thin walled stainless steel containers.
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Accumulate in
Fuel Pool of
Reactor Building
1
Retrieve
Load into Dry
Cask
Load into
Multi Purpose
Container (MPC)
Load into
Sealed Channel
Transport by
Wet Cask
Load into
KBS-3 Container
On-site
Dry Cask Storage
On-site
MPC Storage
Modular Vault
Dry Storage
On-site
Fuel Pool
Storage
On-site Dry
Container
Storage
R
e
-
p
a
c
k
a
g
e
f
o
r
D
i
s
p
o
s
a
l
(
i
f
n
e
c
e
s
s
a
r
y
)
T
r
a
n
s
f
e
r
t
o
G
D
F
OPTIONS
2
3
5
7
9
11
4
6
8
10
12
13
14
Figure D6 – Spent Fuel Management Options
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Notes for Figure D6 – Spent Fuel Management Options
1 After discharge from the reactor the spent fuel assemblies (each comprising the fuel element bundle
and its surrounding channel box) will be accumulated in the spent fuel pool for a period of decay
storage to allow them to cool. This period is usually several years (believed to be approx. 5 years for
typical BWR spent fuel) and the spent fuel pool water will be conditioned to maintain the correct
temperature, cleanliness and chemistry conditions.
2 The method used for retrieval of the spent fuel/channel box will depend upon the subsequent storage
option chosen. The usual method for removal of the spent fuel assemblies is to load them vertically
into a flask which is submerged into the spent fuel pool, followed by lidding and removal of the cask.
Operations to decontaminate the flask exterior and condition the internal water are also undertaken
prior to transport of the flask to its intended location. This is usually a reprocessing facility but for the
UK ABWR spent fuel reprocessing will not be assumed and hence the retrieval method is likely to be
different. The envisaged retrieval method, as part of each of the noted options will be described in the
relevant sections below.
3 Loading the spent fuel assemblies into a dry cask will require a number of steps which will include the
following:
a. Load the spent fuel assemblies, under water, into an inner container, which is compatible with
the chosen dry cask. This container may be an MPC, as described in 5 below.
b. Drain the water from the container and seal with a lid closure. This would be completed in a
facility which provides appropriate shielding, containment for contamination control and
remotely operated equipment.
c. Dry the internal environment and backfill with an inert gas.
d. Transfer the containers into a dry cask.
4 The dry casks are stored in a suitable facility which is likely to be similar to an industrial type
warehouse which will provide weather and security protection. The casks are likely to be transported
into and out of the facility using specialist heavy equipment, probably similar to that offered by Holtec,
as shown at http://rampac.energy.gov/docs/education/Q8.pdf. It is unlikely that the store will require
any conditioning of the internal environment as the casks can be inspected externally and any
necessary maintenance undertaken.
5 Use of a Multi-Purpose Container (MPC) similar to the type used by Holtec (as above) inside the dry
cask may allow the spent fuel elements within to be disposed of directly to a GDF without further re
packaging. This would be subject to granting of a LoC, which will include a disposability assessment,
from RWMD and other decisions related to logistics and commercial/business requirements. For
example, a container suitable for GDF disposal is likely to be manufactured from relatively expensive
materials and to more stringent requirements than required for a period of on-site interim storage
inside the cask. This could entail greater costs at this stage than if packaging for disposal was delayed
until a GDF was available. The time period from the start of spent fuel interim storage to the
availability of a GDF may also allow for the development of, as yet unknown, more cost effective
techniques for producing a disposable package. Or the requirements for disposability at the GDF may
have been modified to allow for a more cost effective method of spent fuel packaging to be used.
These potential benefits would need to be weighed against the requirements for repackaging of the
spent fuel, when the GDF was available, assuming that an MPC was not initially used.
6 Storage of the MPC would be within a suitable dry cask in facilities as described in 4 above.
7 The spent fuel assemblies may be transferred into sealed channels, either singly or in groups after
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being removed from the spent fuel pool and dried. The sealed channels would be part of a Modular
Vault Dry Store (MVDS) facility, similar to that described in the following link:
http://www.wmsym.org/archives/1988/V2/80.pdf. The operations to remove, dry and transport the
assemblies to the MVDS would require specialist equipment, probably remotely operated within a
shielded containment facility and utilising a shielded transport vehicle to transfer the assemblies from
the reactor building to the, probably separate, MVDS building.
8 Interim storage within a MVDS system can continue until a GDF is available as the cooling is by
passive means, which increases its reliability. The link in 7 above describes one such system, which is
based upon the proven method for dry storage of spent Magnox fuel at the Wylfa power station.
9 Use of a wet cask for removal of the spent fuel assemblies, as described in 2 above, will facilitate the
transport of the assemblies to an on-site fuel pool storage facility, which would be separate from the
reactor’s operational spent fuel pool.
10 Fuel pool storage would continue for an extended period until a GDF was available. This would
include separate facilities for conditioning, cooling and cleaning the water to maintain the correct
quality.
11 The KBS-3 container is a developed container which may be suitable for disposal to a GDF, subject to
granting of a LoC by RWMD. Hence, the spent fuel assemblies may be loaded directly into these
containers using a procedure similar to as described in 3 above. However, as these containers are
likely to be more expensive than those used for interim storage only, their use will need to be carefully
considered taking account of factors such as discussed in 5 above. Use of these containers may also
eliminate the opportunity of utilising any improved methods or take account of any changes in the
Regulatory and disposability regimes during the interim storage period.
12 The KBS-3 containers would be stored during the interim period within a suitable dry cask in facilities
as described in 4 above.
13 Where appropriate, the spent fuel assemblies will need to be re-packaged for transport to and disposal
at a GDF. As described in 5/6 &11/12 above, where the assemblies are initially stored in an MPC or
the KBS-3 container these are likely to be acceptable for disposal, subject to granting of a LoC by
RWMD. Hence, re-packaging should not be required for these options.
For options 3/4 and 7/8, where the spent fuel assemblies have been stored for the interim period in
dry facilities, re-packaging by transfer into a suitable disposal container will require a suitable facility
which incorporates shielding and containment for contamination control and remotely operated
equipment.
For option 9/10 above transfer into a suitable disposal container will require similar facilities as above
plus draining and drying facilities to ensure that the spent fuel assemblies contain minimal (or
preferably zero) residual moisture prior to loading into a suitable disposal container.
These latter options will require the granting of appropriate LoCs by RWMD prior to re-packaging for
disposal.
14 Transfer of the packaged spent fuel assemblies to a GDF will be subject to granting of the appropriate
LoCs, as described above. Depending upon the packaging adopted for disposal a suitable transport
method will need to be adopted. For example, the Holtec system referenced above includes various
transport casks. However, within the UK the use of these or any other transport cask will need to be
assessed by the appropriate authority as well as the noted requirement for LoCs. The Office for
Nuclear Regulation (ONR) currently covers Nuclear Transport aspects within their remit under its
Radioactive Materials Transport (RMT) function.
doc_159014982.pdf
These Radioactive Waste Management Arrangements will fulfil the requirements of the Environment Agency Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant Designs (1), together with consideration of the Environment Agency REPs (2) and the applicable requirements of the Office for Nuclear Regulation (ONR) for the management of wastes and spent fuel (SF) from the Hitachi-GE UK ABWR.
Form10/00
Hitachi-GE Nuclear Energy, Ltd.
UK ABWR
UK ABWR Generic Design Assessment
Radioactive Waste Management Arrangements
Document ID : GA91-9901-0022-00001
Document Number : WE-GD-0001
Revision Number : C
Form10/00
Hitachi-GE Nuclear Energy, Ltd.
UK ABWR
DISCLAIMERS
Proprietary Information
This document contains proprietary information of Hitachi-GE Nuclear Energy, Ltd. (Hitachi-GE), its
suppliers and subcontractors. This document and the information it contains shall not, in whole or in part,
be used for any purpose other than for the Generic Design Assessment (GDA) of Hitachi-GE’s UK ABWR.
This notice shall be included on any complete or partial reproduction of this document or the information it
contains.
Copyright
No part of this document may be reproduced in any form, without the prior written permission of
Hitachi-GE Nuclear Energy Ltd. Copyright (C) 2014 Hitachi-GE Nuclear Energy, Ltd. All Rights
Reserved.
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Table of Contents
1. Acronyms ................................................................................................................................................ 1
2. References ............................................................................................................................................... 6
3. Introduction ............................................................................................................................................. 8
4. Waste Management Objectives and Principles........................................................................................ 8
4.1. Objectives ....................................................................................................................................... 8
4.2. Principles ........................................................................................................................................ 9
5. Assumptions ............................................................................................................................................ 9
6. Regulatory Context ............................................................................................................................... 10
6.1. Consideration of the REPs ............................................................................................................ 11
7. The Decision Making Process ............................................................................................................... 11
8. Management of the Radioactive Waste Streams ................................................................................... 12
8.1. Wastes Arising .............................................................................................................................. 15
8.1.1. Very Low Level Waste ............................................................................................................. 15
8.1.2. Low Level Waste ...................................................................................................................... 16
8.1.3. Intermediate Level Waste ......................................................................................................... 19
8.1.3.1. General Packaging Options .................................................................................................. 20
8.1.3.2. Sludge (Crud) ....................................................................................................................... 21
8.1.3.3. Powder Ion Exchange Resins ............................................................................................... 22
8.1.3.4. Higher Activity Metals ......................................................................................................... 22
8.1.3.5. Irradiated Metal (Decommissioning) ................................................................................... 23
8.1.3.6. Borderline Wastes ................................................................................................................. 23
8.2. Spent Fuel ..................................................................................................................................... 24
8.2.1. Fuel Characteristics .................................................................................................................. 24
8.2.2. Packaging Options .................................................................................................................... 26
8.2.2.1. Dry Cask............................................................................................................................... 26
8.2.2.2. Multi-Purpose Container ...................................................................................................... 27
8.2.2.3. Modular Vault Dry Storage System ...................................................................................... 27
8.2.2.4. KBS-3 Container .................................................................................................................. 27
8.2.2.5. On-Site Fuel Pool Option ..................................................................................................... 28
8.3. Summary of Management Options ............................................................................................... 28
9. Waste Management Infrastructure ......................................................................................................... 29
9.1. Waste Treatment and Packaging Facilities ................................................................................... 30
9.1.1. Low Level Waste .................................................................................................................. 30
9.1.2. Intermediate Level Waste ..................................................................................................... 30
9.1.3. Spent Fuel............................................................................................................................. 31
9.2. Waste Stores ................................................................................................................................. 31
9.2.1. Low Level Waste .................................................................................................................. 31
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9.2.2. Intermediate Level Waste ..................................................................................................... 31
9.2.3. Spent Fuel............................................................................................................................. 32
9.3. Waste Packages, Transport Systems and LoCs ............................................................................. 33
10. Development of an Integrated Waste Strategy ...................................................................................... 34
11. Plan for Disposability Assessments ....................................................................................................... 34
11.1. Introduction to the RWMD Disposability Assessment ............................................................. 34
11.2. Hitachi-GE Plan for the RWMD Disposability Assessment ..................................................... 35
12. Conclusions ........................................................................................................................................... 35
Appendix A – Consideration of the REPs ..................................................................................................... 37
Appendix B - Waste Hierarchy ...................................................................................................................... 39
Appendix C - Waste Stream and Spent Fuel Descriptions ............................................................................ 40
1. Introduction ........................................................................................................................................... 40
2. Nature and Quantity of the Wastes and Spent Fuel ............................................................................... 40
2.1. Operational VLLW ....................................................................................................................... 40
2.2. Operational LLW .......................................................................................................................... 41
2.3. Operational ILW ........................................................................................................................... 44
2.4. Decommissioning Non-Radioactive Waste .................................................................................. 46
2.5. Decommissioning VLLW ............................................................................................................. 46
2.6. Decommissioning LLW ................................................................................................................ 47
2.7. Decommissioning ILW ................................................................................................................. 48
2.8. Spent Fuel ..................................................................................................................................... 48
Appendix D - Descriptions of Conditioning Options .................................................................................... 50
1. Introduction ........................................................................................................................................... 50
Notes for Figure D1 – Very and Low Level Wastes ...................................................................................... 52
Notes for Figure D2 – Intermediate Level Waste; Solidification of Sludge/ Crud and IX Resins ................ 55
Notes for Figure D3 – Intermediate Level Waste; Encapsulation of Activated Metals ................................. 58
Notes for Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud and IX Resins ............................ 61
Notes for Figure D5 – Intermediate Level Waste; Drying of Activated Metals ............................................ 64
Notes for Figure D6 – Spent Fuel Management Options .............................................................................. 67
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1. Acronyms
ABWR Advanced Boiling Water Reactor
AC Atmospheric Control System
ALARA As Low As Reasonably Achievable
ALARP As Low As Reasonably Practicable
BAT Best Available Technique
BPEO Best Practicable Environmental Option
BPM Best Practicable Means
Bq Becquerel
BSS Basis Safety Standards Directive
BWR Boiling Water Reactor
C&I Control and Instrumentation
CAD Controlled Area Drain
CCI Commercially Confidential Information
CD Condensate Demineraliser
CDL Calculated Detection Limit
CF Condensate Filter
COMAH Control of Major Accident Hazards
CONW Concentrated Waste System
CP Corrosion Product
CSG Combustion Sector Guidance Note
CST Condensate Storage Tank
CUW Reactor Water Clean-up System
CW Circulating Water System
CWP Circulating Water Pump
D/W Dry Well
DAW Dry Active Waste
DCD Design Control Document
DECC Department of Energy and Climate Change
DEFRA Department for Environment, Food and Rural Affairs
DF Decontamination Factor
DORIS The marine dispersion model used in PC-CREAM 08
®
DPUR Dose Per Unit Release
EIA Environmental Impact Assessment
EMCLs Environmental Media Concentration Limits
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EPR/EPR10 Environmental Permitting (England and Wales) Regulations 2010
EQS Environment Quality Standards
ERICA Environmental Risk from Ionising Contaminants: Assessment and Management
ESE Environmentally Sensitive Equipment
EU European Union
f-value Fuel leakage rate
F/D Filter Demineraliser
FAP Forward Action Plan
FDP Funded Decommissioning Programme
FDW Feedwater System
FP Fission Product
FPC Fuel Pool Cooling and Clean-up System
GDA Generic Design Assessment
GDF Geological Disposal Facility
GEP Generic Environmental Permit
GNF Global Nuclear Fuel
GSD Generic Site Description
HAW Higher Activity Waste
HCEP How to Comply with your Environmental Permit
HCW High Conductivity Waste System
HEPA High Efficiency Particulate Air (Filter)
HFE Human Factors Engineering
HFF Hollow Fibre Filter
HLW High Level Waste
HNCW HVAC Normal Cooling Water System
HOP Hydrazine, Oxalic acid, Potassium permanganate
HS Heating Steam System
HSCR Heating Steam and Condensate Water Return System
HSD Hot Shower Drain
HSE Health and Safety Executive (UK)
HVAC Heating Ventilation and Air Conditioning System
HWC Hydrogen Water Chemistry
I&C Instrumentation and Control
IA Instrument Air System
IAEA International Atomic Energy Agency
ICRP International Commission on Radiological Protection
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IEX Ion-Exchange (demineraliser) system
ILW Intermediate Level Waste
IPPC Integrated Pollution Prevention and Control
IRA Initial Radiological Assessment
IWS Integrated Waste Strategy
J ABWR Japanese ABWR
KK-6 Kashiwazaki-Kariwa Nuclear Power Station Unit 6
KK-7 Kashiwazaki-Kariwa Nuclear Power Station Unit 7
LCW Low Conductivity Waste System
LD Laundry Drain System
LLW Low Level Waste
LLWR Low Level Waste Repository
LoC Letter of Compliance
LOCA Loss of Coolant Accident
LPRM Local Power Range Neutron Monitor
LS Laundry System
LWR Light Water Reactor
MCERTS Monitoring Certification Scheme
MS Main Steam System
NDA Nuclear Decommissioning Authority
NHS Non Human Species
NMCA Noble Metal Chemical Addition
NPP Nuclear Power Plant
NRW Natural Resources Wales
NUREG Nuclear Regulatory Commission Regulation (US)
OG Off gas
ONR Office for Nuclear Regulation
OSPAR Oslo and Paris Convention on Protection of the Marine Environment of the North East
Atlantic
P&D Plumbing and Drainage System
P&ID Process and Information Document for Generic Assessment of Candidate Nuclear
Power Plant Design
P/C Power Centre
PCI Pellet Cladding Interaction
PCSR Pre-Construction Safety Report
PI Personal Information
ppb parts per billion
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PWR Pressurised Water Reactor
QA Quality Assurance
QAP Quality Assurance Plan
QC Quality Control
QMP Quality Management Plan
QMS Quality Management System
R/B Reactor Building
RCLEA Radioactively Contaminated Land Exposure Assessment
RCW Reactor Building Cooling Water System
REP Radioactive Substances Regulation – Environmental Principle
RGP Relevant Good Practice
RP Requesting Party
RPDP Radiation Protection Developed Principle
RQ Risk Quotient
RSA Radioactive Substances Act
RSR Radioactive Substances Regulation
RSW Reactor Building Service Water System
RW/B Radwaste Building
RWMA Radioactive Waste Management Arrangement
RWMD Radioactive Waste Management Directorate
S/B Service Building
S/P Suppression Pool
SA Station Service Air System
SAM Sampling System
SAP Safety Assessment Principle
SF Spent Fuel
SFAIRP So Far As Is Reasonably Practicable
SFP Spent Fuel Pool
SGTS Standby Gas Treatment System
SJAE Steam Jet Air Ejector
SLC Standby Liquid Control System
SoDA Statement of Design Acceptability
SPCU Suppression Pool Clean-up System
SQEP Suitably Qualified and Experienced Person (UK)
SRNM Start-up Range Neutron Monitor
SS Spent Sludge System
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Sv Sievert
TBURD Technical Baseline and Underpinning Research and Development
T/B Turbine Building
TIP Traversing In-core Probe
TCW Turbine Building Cooling Water System
TSW Turbine Building Service Water System
TV Tank Vent Treatment System
UF Uncertainty Factor
UK United Kingdom
US United States
VLLW Very Low Level Waste
WENRA Western European Nuclear Regulators' Association
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2. References
1 Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant
Designs, Version 2, March 2013.
2 Regulatory Guidance Series, No RSR 1, Radioactive Substances Regulation – Environmental
Principles, Version 2, April 2010.
3 Demonstration of BAT, GA91-9901-0023-00001, XE-GD-0097, Rev C, March 2014.
4 Fundamentals of the management of radioactive waste: An introduction to the management of
higher-level radioactive waste on nuclear licensed sites: Guidance from the Health and Safety
Executive, the Environment Agency and the Scottish Environment Protection Agency to nuclear
licensees, December 2007.
5 An Overview of NDA Higher Activity Waste, February 2012.
6 UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear
Industry, August 2010.
7 Waste Acceptance Criteria – Very Low Level Waste Disposal, LLW Repository Ltd,
WSC-WAC-VER – Version 3.0 – April 2012.
8 Waste Acceptance Criteria – Low Level Waste Disposal, LLW Repository Ltd, WSC-WAC-LOW
– Version 3.3 –April 2012 (currently subject to revision).
9 Waste Acceptance Procedure – Overview, LLW Repository, WSC-WAP-OVR – Version 2.0 –
January 2011.
10 Justification for no active oils generated for ABWR operations; WE-GD-0005; Rev.0; February
2014.
11 Waste Package Specification for 2 metre box waste package, NDA (RWMD), WPSG No. 350/3,
January 2013.
12 Waste Package Specification for 4 metre box waste package, NDA (RWMD), WPSG No. 330/3,
January 2013.
13 http://www.nda.gov.uk/rwmd/producers/detail.cfm#specifications.
14 NDA Technical Note no. 13403461, Geological Disposal, Generic specification for robust shielded
waste packages November 2010.
15 Magnox Care and Maintenance Preparation Wastes in Ductile Cast Iron Containers (Conceptual
stage) Summary of Assessment Report, Issue date of assessment report, 19 May 2010.
16 Croft Rectangular Box Safstores® for the packaging of ILW (Conceptual stage) Summary of
Assessment Report Issue date of assessment report, 24 February 2012.
17 Guidance on decision making for management of wastes close to the LLW and ILW categorisation
boundary that could potentially cross the LLW boundary, Helen Cassidy, NWP/REP/016, Issue 2 –
February 2013.
18 A White Paper on Nuclear Power, Department for Business, Enterprise and Regulatory Reform,
CM 7296, January 2008.
19 http://www.wnfc.info/proceedings/2004/presentations/woodwardppt.pdf.
20 Packaging of Sizewell B Spent Fuel (Pre-Conceptual stage), Summary of Assessment Report,
Issue date of assessment report 23 December 2011.
21 http://www.skb.se/Templates/Standard____24109.aspx.
22 Industry Guidance, Interim Storage of Higher Activity Waste Packages – Integrated Approach,
Effective from November 2012.
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23 ENG01 - Specification and Guidance on the Content and Format of an Integrated Waste Strategy,
Revision 3, 19 October 2012.
24 Latest Liquid Waste and Solid Waste Treatment System and Development, WJ-GD-0017, to be
provided at a later date.
25 Solid waste generation arising from operation and decommissioning, WJ-GD-0018, to be provided
at a later date.
26 The Activities of Solid wastes, WJ-GD-0019, to be provided at a later date.
27 ABWR Plant General Description, December 2006.
28 Parameters for spent fuel assessment in GEP, UE-UK-0001, April 2013.
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3. Introduction
These Radioactive Waste Management Arrangements will fulfil the requirements of the Environment
Agency Process and Information Document for Generic Assessment of Candidate Nuclear Power Plant
Designs (1), together with consideration of the Environment Agency REPs (2) and the applicable
requirements of the Office for Nuclear Regulation (ONR) for the management of wastes and spent fuel
(SF) from the Hitachi-GE UK ABWR. The Hitachi-GE Radioactive Waste Management Arrangements
have been produced with potential site operators in mind to ensure that the site specific Integrated Waste
Strategy (IWS) can follow on from these generic arrangements without major changes to the fundamental
approach.
This document describes the radioactive waste management arrangements, including how radioactive
wastes and SF will arise throughout the facility’s lifecycle (including decommissioning), and how they will
be managed and disposed of.
This document will focus on those wastes which will be treated and packaged in some way either for
immediate disposal or on-site interim storage as ‘solid’ waste, pending disposal at a later date. The systems
in place for the management and treatment of Gaseous and Aqueous Radioactive Discharges are described
in the suite of common documents, and dedicated chapters provided as part of the PCSR and will be
underpinned by a demonstration of BAT (3).
The Hitachi-GE Radioactive Waste Management Arrangements have been developed with the UK
regulatory system in mind for the UK ABWR. Where applicable, the current Japanese practice for dealing
with a specific waste stream will be outlined and discussed in the context of current UK practices. Where
UK practice differs this will be noted and management options will be discussed for the specific stream. In
such circumstances a recommendation for additional work will be included to develop appropriate
solutions, which may, at this stage comprise a number of suitable options. These arrangements will be
updated to reflect any changes or on-going developments.
4. Waste Management Objectives and Principles
The following high level objectives and principles are used within Hitachi-GE to inform decision making
and design development for systems having a bearing on the generation or treatment of radioactive wastes.
These have been developed within Hitachi-GE with future site operators in mind. The waste management
objectives and principles are used to guide the development of strategies and in due course, their
implementation on specific sites. These take due consideration of the objectives and principles published
by HSE (4) and Environment Agency (2).
4.1. Objectives
The objectives for the Hitachi-GE and UK ABWR Radwaste Management Arrangements are to:
? Safely control and account for radioactive waste;
? Protect human health and the environment both now and in the future. Where it is reasonably
practicable, deliver this objective by concentrating, containing and packaging the waste and
isolating it from the accessible environment by disposing of it in appropriate geological facilities
at the earliest opportunity;
? Balance the use of environmental, social and economic resources in an optimal way;
? Ensure undisturbed power production from the reactor, provided health, safety and environmental
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protection are not compromised.
4.2. Principles
The principles to be applied when developing and ultimately implementing waste management and
decommissioning strategies will be:
? The safety and protection of the public, the workforce and the environment, are of prime
importance and wastes should be managed to ensure an optimal level of protection. Sustainability
should be considered such that unduly restrictive burdens are not left for future generations;
? Strategies and practices will be compliant with UK Government and relevant international policies,
legislation and standards, including the policy aim of sustainable development;
? Mature and/or standardised options should be given preference;
? All radioactive wastes should be managed within strategies that are integrated, take account of
interactions and dependencies, and consider the full lifecycle of the wastes, including the
possibility of long term storage prior to disposal;
? Strategies will take account of the “Waste Hierarchy” as described in Appendix B in an attempt to
avoid waste generation and where this is not possible, to minimise the quantities and activities of
waste, so far as is reasonably practicable;
? Characterisation and segregation of wastes should be used to help ensure subsequent management
is safe, efficient and effective;
? Strategies should be informed by the best scientific knowledge. Technologies and techniques
which are planned to be deployed should be readily available, or be capable of development with
planned and realistic levels of R&D;
? The accumulation of wastes should be minimised and they should not intentionally be stored in
their raw, as produced, form for any longer than is necessary to organise their efficient packaging
in a way that is acceptable to the organisation responsible for disposal. The waste treatment and
packaging process should ensure that waste is processed into a passively safe condition as soon as
practicable;
? Appropriate documentation and records will be produced, stored and maintained covering the
selection and design of waste treatment processes, their operation and the characteristics of the
raw and packaged waste;
? Waste packages should be accessible during interim storage.
5. Assumptions
The development of these waste management arrangements is based on a number of assumptions of which
the following are the main high level ones:
? The site comprises one reactor unit;
? The design basis for the ABWR fuel is to use the GE14 type (Ref Global Nuclear Fuel BWR
design);
? Government Policy, standards, legislative and regulatory environments remain unchanged, or
changes pending have no significant impact;
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? Strategies detailed within these arrangements will reflect only currently available technologies;
? Definitions of waste categories will remain unchanged;
? The waste management arrangements take as a starting point that the wastes that will be generated
from the reference reactor design and the processes, equipment and facilities comprising that
design;
? It is assumed that SF will not be reprocessed but will be treated as a waste stream and stored,
packaged and disposed of accordingly;
? Because the GDF for HAW (ILW and HLW, including SF) is still only at a conceptual stage and
detailed packaging requirements are not available, the focus of these arrangements is to ensure that
ILW and SF can be safely stored on-site for an extended interim period of many decades;
? Materials from the abatement of liquid and gaseous discharges (such as ion exchange resins and
charcoal filter media) are included within these arrangements, but the actual discharges to the
environment are discussed in a separate paper;
? LLWR (or a successor) is available throughout the operational and decommissioning phases and
its “Conditions for Acceptance” (CFA) continue to apply unchanged. LLWR Ltd offers a
well-recognised service for the whole UK Nuclear industry. It is currently adopted by many
operational sites in the UK and also by the legacy sites. However, there are a number of other
facilities which can be used for waste disposals which are not listed in this document. The
references to LLWR are assumed for GDA and to enable the obtaining of an ‘Acceptance in
Principle’ for management/disposal of VLLW and LLW based upon the services undertaken by
LLWR Ltd. When appropriate the application of BAT will be implemented by the licensee to
determine the best route for specific wastes, when other facilities and techniques for management
of VLLW and LLW will be included;
? These arrangements take into account the liquid waste streams produced by a UK ABWR which
are delivered to the Waste Treatment Facility for treatment and solidification. It does not address
liquid and gaseous waste discharges, non-radioactive waste generated at the plant or wastes
generated from the operation of the interim storage facilities;
? These arrangements assume that the options presented can be used by utilities to demonstrate that
the design makes use of Best Available Techniques (BAT). The intention is to preserve as much
flexibility as possible whilst providing confidence that BAT solutions can be identified by utilities
in the future. This is judged to be sensible given the range of uncertainties at this stage. In
particular, issues relating to the future treatment and transport of certain wastes may need to be
resolved via consultations with RWMD.
6. Regulatory Context
The regulatory framework will be as currently applicable to UK Nuclear facilities. This framework is
largely non-prescriptive and a future operator must demonstrate to the Regulators that they fully
understand the hazards and risks associated with their operations and know how to control and reduce
them. Similarly, they are obliged to assess and minimise their impact on the environment.
To achieve high standards of nuclear safety and to restrict environmental impact, a UK operator must
adhere to the principle of continuous improvement, which is embedded in UK law, and requires nuclear
designers and operators to reduce risks as far as is reasonably practicable.
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In managing radioactive wastes in the UK, the main stakeholders for Hitachi-GE and their responsibilities
are:
? Government – To determine policy, in the light of international agreements and guidance, and
prepare statutory legislation;
? Regulators – To enforce the law and pay cognisance to Government policy and publish guidance
which interprets Government policy;
? Operators – To implement appropriate waste management strategies, in compliance with
Government policies and legislation and regulatory requirements/guidance.
The main policies, principles, guidance and regulations that relate to waste management and
decommissioning are identified below:
? International Directives and initiatives:
o The OSPAR Convention;
o The Basic Safety Standards Directive (Council Directive 96/29/EURATOM);
o Euratom Treaty Article 37.
? National Policy:
o The Review of Radioactive Waste Management Policy: Final Conclusions (Cmnd 2919)
1995, as amended;
o UK Strategy for Radioactive Discharges 2011-2020;
o Managing Radioactive Waste Safely (MRWS) – Proposals for developing a policy for
managing solid radioactive waste in the UK.;
o Policy for the Long Term Management of Solid Low Level Radioactive Waste in the
United Kingdom, By Defra, DTI and the Devolved Administrations, March 2007.
? National Legislation:
o Nuclear Installations Act 1965;
o Environmental Permitting Regulations 2010/2011;
o Ionising Radiation Regulations 1999;
o Pollution Prevention and Control 2000.
? Regulatory Guidance:
o Regulation Environmental Principles (2);
o Guidance to Inspectors on the Management of Radioactive Materials and Radioactive
Waste on Nuclear Licensed Sites.
6.1. Consideration of the REPs
As noted above, the REPs, which are listed in (2), have been considered during the development of these
arrangements. In particular, the extent to which the relevant REPs listed in the Environment Agency
Process and Information Document (1) has been addressed is shown in Appendix A.
7. The Decision Making Process
The decision making process will be as currently adopted within Hitachi-GE and as informed by the
Demonstration of BAT (3). The decision making process has been developed, with future site operators in
mind – to allow them to demonstrate full understanding of why certain techniques have been adopted, and
to allow for site specific criteria to be included.
Hitachi-GE recognises that there is a need to underpin the selection of optimal disposal routes by the
demonstration that those routes represent the BAT. These arrangements are not supported by a full BAT
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assessment as this will be conducted at a site specific stage. However, this Radioactive Waste Management
Arrangements document refers to the requirement to apply the BAT process (3), which addresses the
Claims, Arguments and the main Evidence required to select the optimal disposal route for GDA, as
described in Claim 4 of the Demonstration of BAT (3).
Where required, the appropriate technique which includes Cost/Benefit analysis will be undertaken as part
of Hitachi-GE’s decision making process.
The noted BAT assessment will also take account of application of the Waste Hierarchy and ALARP
considerations to determine optimal disposal routes and levels of discharge. Hence, such analysis will not
be repeated in this document.
The decision making process within Hitachi-GE includes making a judgement on whether a specific waste
route is considered BAT. As part of the Hitachi-GE Management arrangements information will be
gathered in the following areas to support the decision making process:
? The number of people (workers and the public) and other environmental targets that may be
exposed to radiological risk;
? The likelihood that they could be exposed to radiation, where exposure is not certain to happen;
? The magnitude and distribution in time and space of radiation doses that they will or could
receive;
? Nuclear security and safeguards requirements;
? Issues similar to those above, but relating to non-radiological hazards;
? Economic, societal and environmental factors;
? Technical viability;
? Uncertainties in any of the above.
Any decision, and the evidence used to make that decision, will be presented as Evidence in the BAT case.
8. Management of the Radioactive Waste Streams
In the UK, radioactive wastes are classified in terms of the nature and quantity of radioactivity that they
contain and their heat-generating capacity, as follows:
? High Level Wastes (HLW) which includes SF are wastes in which the temperature may rise
significantly as a result of their radioactivity, so this factor has to be taken into account in the
design of storage or disposal facilities;
? Intermediate Level Wastes (ILW) are wastes exceeding the upper boundaries for LLW, but which
do not require heat generation to be taken into account in the design of storage or disposal
facilities;
? Low Level Wastes (LLW) are wastes having a radioactive content not exceeding 4 GBq per tonne
of alpha activity, or 12 GBq per tonne of beta/gamma activity;
? Additionally, Very Low Level Waste (VLLW) is a subset of LLW and there is also a category of
exempt waste which does not require an authorisation for disposal.
Higher Activity Wastes (HAW) comprises all HLW and ILW, and a small fraction of LLW with a
concentration of specific radionuclides that prohibits its disposal at existing and planned future disposal
facilities for LLW. The definitions of these categories are contained within (5).
The radioactive wastes and SF that Hitachi-GE expects the UK ABWR to generate are grouped in Table
8-1 into 13 waste streams and one SF stream. The quantities of these wastes have been minimised as
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described in Claim 3 of the Demonstration of BAT (3). They are described in more detail in Appendix C.
Each waste type is described in turn below and they will be managed according to their categorisation as
VLLW, LLW, ILW or SF.
Table 8-1 Summary of Waste and Spent Fuel Streams
No. Title Description Category Form Arising during
1 Dry active waste Miscellaneous dry, low
activity wastes in various
forms; including metals,
concrete cloths, paper etc.
VLLW
Solid
Operations &
Decommissioning
2 HVAC Filters Arising from filter changing
in air treatment facilities from
exhausts from Reactor,
Turbine (including high
radiation), Radwaste and
Service buildings
LLW Solid Operations
3 Bead resin Arising from the CD, LCW
and HCW demineralisers;
Styrene divinylbenzene
copolymer matrix
LLW Wet Operations
4 Concentrates Arising from the HCW
evaporators comprises
particulate and dissolved
species.
LLW Wet Operations
5 Miscellaneous
combustible
Includes spent hollow fibre
filter membrane, plastic
sheets, paper, wood cloth,
spent bead resin
LLW Solid Operations
6 Miscellaneous
non-combustible
Metal, pipes, cables, lagging,
gas filters, concrete, glass
LLW Solid Operations
7 Sludge (crud) Arising from backwashing of
various filters from the CF
and the LCW systems
ILW Wet Operations
8 Powder resin Arising from the CUW and
FPC filter demineralisers;
cross linked polystyrene
matrix. Contains particulate
corrosion product
ILW Wet Operations
9 Higher activity
metals – control
rods
Cruciform shape metallic
construction containing
stainless steel tubes in each
wing of the cruciform filled
with boron carbide powder.
Hafnium may also be
employed to perform the
same function of reactivity
control
ILW Solid Operations
10 Higher activity Zircaloy box which ILW Solid Operations
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No. Title Description Category Form Arising during
metals – channel
boxes
1
surrounds the fuel bundle.
Approx. 4.3 m long and 15 x
15 cm square
11 Higher activity
metals - others
Various reactor core
components from SRNM and
LPRM systems.
ILW Solid Operations
12 Contaminated
and irradiated
metal and
concrete
Non-combustible, largely
metal and concrete items.
Some very large and
requiring size reduction
LLW Solid Decommissioning
13 Irradiated metal Reactor core components and
from areas subject to
activation
ILW
Solid
Decommissioning
14 Spent Fuel Used fuel elements HLW Solid Operations
Following its generation, radioactive waste will undergo a number of management stages, before it is
finally disposed of. Joint guidance from the HSE, Environment Agency and SEPA identifies six stages (4)
comprising pre-treatment, treatment, packaging, storage, retrieval and disposal. The Regulators note that
implementation depends on the type of waste and the strategy selected for its management.
For managing the waste streams from UK ABWR, the following stages will be used:
? Short term temporary buffer storage – segregated waste may be temporarily stored, for
example, in a buffer tank in order to accumulate an adequate batch size for treatment. At this stage
or during subsequent treatment stages the waste stream will be characterised to determine its
physical, chemical and radionuclide properties. Characterisation methods will include the use of
sampling and analysis (typically for wet wastes) and radiological assay (typically for solid
wastes);
? Treatment – involves changing the characteristics of the waste by processes such as volume
reduction (using drying, cutting, compaction or incineration), cleaning/decontamination by
filtration or ion exchange, or precipitation;
? Packaging for Interim Storage - involves transforming the waste into a form that is suitable for
handling, transfer, storage and, potentially, disposal. This might involve immobilisation of the
waste and placing it into steel drums or other engineered containers to create a waste package. In
the case of VLLW, LLW and ILW, such a package should be suitable for disposal but for SF it is
possible that additional packaging may be needed at the end of interim storage and before
disposal;
? Transfer to Interim Store – movement of the package, within a shielded overpack if necessary,
from the packaging facility to the on-site interim store;
? Interim Storage – on-site interim storage will be necessary for ILW and SF until the planned
Geological Disposal Facility (GDF) is available to receive the waste. Waste may need to be stored
in a passively safe condition for many decades. In the case of LLW, temporary storage of
packaged waste may be required in order to accommodate the requirements of the transport
system and the receipt arrangements at a disposal site (e.g. the Low Level Waste Repository);
1
Classified here as ILW, although for disposal it will be assumed that they remain with the spent fuel
elements
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? Packaging for Disposal – it is possible that SF may need to be re-packaged in order to make it
suitable for disposal. It is likely that such packaging could not be defined confidently until an
actual GDF is available;
? Transport for Disposal – movement of the packaged waste and SF, within suitably shielded and
protected flasks or containers, from the reactor site to the GDF (for SF and ILW), and to a suitably
permitted disposal site for VLLW and LLW (e.g. LLWR);
? Disposal - packages of radioactive waste are deposited in a disposal facility with no intention of
retrieval.
Adequate records will need to be kept during the whole management cycle, as outlined above. In particular
NDA/Radioactive Waste Management Directorate (RWMD) require that records are kept to ensure that the
history of the wastes and specific package properties are preserved to inform their disposability and that
their long term evolving properties will remain within acceptable limits.
8.1. Wastes Arising
The following Sections provide additional information of each category of waste arising together with the
options for their management.
For all VLLW and LLW streams, “Agreements in Principle” will be sought to demonstrate that the wastes
can be managed and, where appropriate, disposed of in the UK. For the ILW and spent fuel streams a
disposability assessment will be sought from RWMD to confirm acceptability of these, when suitably
packaged and subject to detailed LoC assessments, into a future GDF.
The following Sections describe each category of waste envisaged to arise from a UK ABWR and the
options for packaging, interim storage and disposal. These options are currently based upon good practice
as applied to similar wastes on some UK legacy sites which would be considered as BAT at this time.
However, the selection of the appropriate option by the operator, when developing their plans for
packaging, interim storage and disposal, will be subject to the application of BAT within their decision
making process.
8.1.1. Very Low Level Waste
VLLW is a subset of LLW and falls into two distinct categories:
? Low Volume VLLW (‘dustbin disposal’) - wastes that can be safely disposed of to an unspecified
destination with municipal, commercial or industrial waste, each 0.1 cubic metre of material
containing less than 400kBq of total activity, or single items containing less than 40 kBq of total
activity. There are additional limits for carbon (C-14) and tritium (H-3) in wastes containing these
radionuclides.
The radioactive risk from such material is low enough that controls on disposal, after it has been
removed from the premises where it arose, are not necessary;
? High Volume VLLW (bulk disposals) – wastes with maximum concentrations of 4 MBq per tonne
of total activity or 40 MBq per tonne for tritium total activity that can be disposed to specified
landfill sites. There is an additional limit for tritium (40 MBq/t) in wastes containing this
radionuclide (7).
After the waste is removed from its site of origin, it will be subject to controls on its disposal, which will
be specified by the environmental Regulators.
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For a nuclear power station the “dustbin disposal” sub-category is not relevant and only “bulk disposals”
are considered in this document. This is because the Environment Agency does not allow disposal of low
volume VLLW originating from a nuclear site via this route.
VLLW will be dealt with according to current UK practices (6) and subject to, for example, the VLLW
acceptance criteria for the LLWR (7). The main criteria are as noted above with specific restrictions being
determined through for example, the Waste Enquiry Process, as determined by the LLWR.
Only one VLLW stream has been identified for UK ABWR, Dry Active Waste, which is mixed waste that
will arise during reactor operations and decommissioning. The waste consists of contaminated personal
protection equipment, monitoring swabs, plastic, equipment, structures and contaminated plant.
These wastes will require specific removal, handling, sorting and size reduction techniques depending on
their physical form and characteristics prior to treatment. The preferred strategy for this mixed dry active
VLLW, produced during operations and during decommissioning, is to recycle the metallic portion where
practicable and to dispose of the remainder, following pre-compaction if possible, to permitted disposal
sites within the UK.
No credible alternative options have been identified for this waste stream at this time.
8.1.2. Low Level Waste
Most LLW in the UK today arises from the operation of nuclear power stations and nuclear fuel
reprocessing facilities, as well as the decommissioning and clean-up of nuclear sites. Operational LLW is
principally lightly contaminated miscellaneous waste, arising from maintenance and monitoring, such as
plastic, paper and metal. LLW from decommissioning comprises building materials and metal plant and
equipment. Most LLW from nuclear licensed sites is currently managed via the LLWR Ltd organisation,
which provides a service for the range of wastes produced, as described below, and where applicable
specific wastes are disposed of at the LLWR site near Drigg in Cumbria.
LLW is waste having a radioactive content not exceeding 4 GBq per tonne of alpha, or 12 GBq per tonne
of beta/gamma activity.
As an example, based upon current practices LLW can be dealt with in accordance with the strategy to
manage waste through the LLW Joint Waste Management Plan, which is agreed between the waste
producing site and LLWR Ltd. This organisation encourages an approach based upon the Waste Hierarchy
and a number of recycling options are offered for waste producers to use. Where solid wastes are produced
these can be disposed of to the LLWR, in accordance with their published acceptance criteria (8). LLWR
Ltd offers a ‘complete’ service to waste producers which is assumed to be the case for the GDA submission.
However, any future site operator may choose their own arrangements based on the site specific
requirements and the demonstration that they have selected the optimal disposal route at the time. The
service from LLWR Ltd comprises four main options, as detailed in (9) and briefly described below:
a. Metallic Waste Treatment Service
This service is a recycling option. Waste can be treated by decontamination, blasting or melting to
remove the radiological content. The vast majority of the metal can then be recycled as Exempt
Waste thus reducing the volume for disposal. The remaining Secondary Waste is either consigned
to the LLWR for disposal as LLW or VLLW, depending upon its activity, or disposed of by the
Service Supplier, depending upon contract arrangements between the waste producer, LLWR Ltd
and the Service Supplier.
b. Combustible Waste Treatment Service
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This service is a volume reduction option. Waste can be incinerated to reduce its volume. The
remaining Secondary Waste is either disposed of by the Service Supplier or consigned to the Low
Level Waste Repository for disposal as either LLW or VLLW, depending upon waste
characteristics and the contractual arrangements as noted above.
c. Supercompactable Waste Treatment Service
This service is a volume reduction option. Waste can be treated by shredding and / or high force
compaction to reduce its volume. The Secondary Waste product is consigned to the LLWR for
disposal.
d. Low Level Waste Disposal Service
This service is a disposal option. Waste that is not suitable or selected for treatment, has already
been treated or is Secondary Waste from a treatment process is consigned to the LLWR for
disposal. LLWR Ltd offers a range of standard disposal containers which are mostly centred on the
use of the half height ISO container. The containers can be filled with waste in various forms,
compliant with the applicable waste acceptance criteria, either at the waste producer’s site or at the
repository. For the containers consigned from the waste producer, after transport to the repository
the interspaces within the ISO container are filled with a cement grout prior to emplacement in the
repository trench.
A site operator will need to make business decisions, which will also include commercial and logistical
requirements which may indicate that there are other ways of dealing with specific wastes. For example,
the site operator may decide to contract directly with a waste Service Supplier rather than utilising the
LLWR contract arrangements. It is not envisaged that such alternative arrangements would change the
outcome as far as waste management was concerned.
One of the sources of decommissioning LLW in the UK ABWR will be from materials within the
controlled area, such as concrete.
For IX resins, incineration has been used for some sites in Japan however this approach is not being
developed for the UK ABWR GDA submission.
Six distinct LLW streams have been identified as arising during operations only and some additionally
during decommissioning. They are described briefly below together with the options available for
treatment, packaging and disposal. The UK practice generally is to consign packaged LLW to the LLWR
via the Joint Waste Management Plan, as noted above. Significant on site storage facilities are not
necessary but a facility for buffer storage, likely to be for approximately 2 years’ worth of LLW arisings,
will be provided to allow flexibility to manage consignments without adversely affecting generation rates,
which could otherwise cause bottlenecks. For each LLW stream, after accumulation of a number of drums
of packaged waste in the buffer store they will be consigned to the LLWR in accordance with the Joint
Waste Management Plan.
The following Sections briefly describe each stream and propose a preferred method for dealing with the
waste. This could include some recycling or other processes aimed at a reduction in the wastes which will
need to be disposed of to the LLWR, in accordance with the Waste Hierarchy. Where a preferred method is
identified this will be based upon current UK practice for the specific waste, where this is well established.
However, for site specific application, which will occur in several years’ time, when the plans for dealing
with actual wastes are developed, there may be other means available for dealing with specific streams.
Where this occurs, such alternative means will be subject to the applicable BAT assessments and other
optioneering requirements needed to develop the overall business case for implementation by the operator.
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1) HVAC Filters
These filters are installed in the exhaust outlets of various buildings, as listed in Table 8-1 and will
included a number of medium efficiency, bag type filters and several High Efficiency Particulate
Air (HEPA) filters. The detailed type and construction of these filters will depend upon the
operational and safety case requirements.
The construction of the medium efficiency filters usually comprise a metal frame with a bag type
glass fibre filter medium attached.
The construction of the HEPA filters proposed for the UK ABWR, as currently used in Japanese
reactors, comprises a wooden frame, glass fibre element and aluminium separators between
adjacent elements.
The design and application of each type of filter for the UK ABWR is yet to be decided and will
follow UK practice for these types of filters for both performance/safety requirements and
design/construction. The filters can be supplied in a range of shapes, sizes and materials, including
wood and metal, mostly with glass fibre media. It is envisaged that whatever materials of
construction are used the currently available techniques of compaction and cement grouting are
likely to be adequate for packaging of filters for disposal.
2) Bead Resin
These are used within demineraliser beds to remove soluble radioactive species from various
water circuits from the Condensate Demineraliser (CD), Low Conductivity Waste (LCW) and
High Conductivity Waste (HCW) systems. The precise resins used by any UK site operator will be
determined by the operator themselves; however the following are assumed for the GDA
submission. The resins used in the Japanese ABWR systems comprise a styrene divinylbenzene
copolymer matrix which is chosen for optimal removal of the specific radionuclides present. When
the resins are spent the beds are periodically discharged to buffer storage tanks to allow
accumulation in order to implement efficient periodic packaging campaigns.
Common practice in the UK is to store the spent resins under water for a period and then apply a
solidification process; hence it will be assumed that bead resins are solidified with polymer or a
suitable cement formulation in disposal drums, typically 200 litre.
3) Concentrates
The Concentrated Waste System (CONW) treats wastes arise from the HCW evaporators, which
comprises wastes with such a high conductivity because of particulate and dissolved species that
they are not suitable for demineralisation. The HCW streams arise from sources such as RW/B
sumps, Service Building (S/B) sumps, CD drains and other similar sources.
It has been determined that as cementation is common practice in the UK it is assumed to be
applied to the UK ABWR. Japanese practice for these wastes is to solidify with a suitable cement
formulation in disposal drums, typically 200 litre.
4) Miscellaneous Combustible
These wastes are generated through routine operations, maintenance and decommissioning in
radioactive areas. They consist mainly of contaminated personal protective equipment, CF and
LCW spent hollow fibre membrane, polyethylene (sheet, bag), paper, wood, cloth, rubber gloves,
turbine oil waste, spent bead resin, spent active carbon filter media.
However, it is expected that there will be no radioactive oils in this category because of the
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location of lubricated items and methods of operation of the turbines and other equipment where
lubricating oils are used (10).
The treatment options include incineration, followed by encapsulation in cement formulation and
disposal of ash, or compaction within a drum and direct disposal within a container suitable for the
disposal facility. This will be assumed to be undertaken via the LLWR Ltd services and recycling/
management options
5) Miscellaneous Non-combustible
These wastes are generated as above for the combustible streams, through routine operations,
maintenance and decommissioning in radioactive areas comprising such materials as metals,
concrete, lagging, glass etc. The metals category, during decommissioning, with consist mainly of
removed plant items and is included as a separate category below.
These wastes may well include some items which could be dealt with in ways other than direct
disposal and the requirements of the Waste Hierarchy will be applied. For example contaminated
metal items could be decontaminated and the metal recycled via a suitable route. LLWR Ltd
currently offers this as a service to waste producers and would be a viable route for this recycling.
Treatment options will be influenced by the quantities arising and during normal operations it is
likely that size reduction followed by grouting or compaction within a drum and direct disposal to
a suitable facility will be the best option for wastes requiring disposal.
6) Contaminated and Irradiated Metal and Concrete (Decommissioning)
These wastes will be generated during the decommissioning phase and will comprise large
volumes of metal and concrete items. Many will be very large and requiring size reduction. The
strategy for management of these wastes is described in Chapter 27 of the PCSR.
However, the means of dealing with these wastes arising in the LLW category will be consistent
with the Waste Hierarchy and (for example) the Joint Waste Management Plans developed with
LLWR Ltd. For some contaminated metals the option of decontamination and recycling will be
adopted whilst some irradiated metals may need to be packaged and disposed of to the repository.
Concrete is likely to be subject to decontamination and possible reuse as infill or aggregate for
other constructions whilst the removed contaminated portion or any activated portion (e.g.
reinforcing bar) would be disposed of to the repository.
Processes for dealing with these portions will largely comprise size reduction and packaging in a
suitable disposal container. As noted above the LLWR is largely based on the use of the half height
ISO container but the RWMD standard 2m or 4m box with appropriate shield thickness (11) &
(12) may also be applicable for some items which would also be grouted in place with a suitable
cement formulation. The use of this route and how much of the waste is consigned in this way will
be dependent upon the application of the Waste Hierarchy and whether some items may be
suitable for recycling or reuse via the (for example) LLWR waste management services.
8.1.3. Intermediate Level Waste
ILW has radioactivity levels that are higher than LLW but which do not generate enough heat to require
special storage or disposal facilities. However, like other radioactive waste, it needs to be contained to
protect workers and the public from the radiation and contamination. ILW includes metal items such as
fuel cladding and redundant reactor components, and sludges and resins from the treatment of radioactive
liquid effluents. The main waste types are described below and comprise:
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? Sludge, also referred to as crud, arising from filtration of water streams
? Powder ion exchange resins, arising from water stream filter/demineralisers. This stream will also
contain some sludge
? Activated metals arising from locations within the reactor which are subjected to irradiation
At this point in time (i.e. GDA) several viable options exist to manage radioactive waste over the longer
term in the UK. A site operator will make the final decision as to how they intend to manage their wastes;
however for GDA all the viable options have been described. In order to demonstrate that the waste
produced by the UK ABWR can be disposed of and managed in the UK regulatory system a single option
has been selected to enable more detailed studies and analysis. At this stage the option of cement
encapsulation (for solid items) and solidification (for wet/slurry wastes) into unshielded stainless steel
containers will be selected as the packaging method to be adopted for a disposability assessment by
RWMD.
The ILW will be dealt with according to the requirements of RWMD and will be subject to disposability
assessments during the GDA process. For later, site specific waste streams and packaging methods a Letter
of Compliance (LoC) process is used to obtain endorsement. The adoption of the noted solution for GDA
will not preclude a future operator from selecting another method, subject to completion of relevant
optioneering and BAT assessments, business case analysis and LoC assessments.
The physical, chemical and radionuclide properties of the wastes will be assessed by RWMD to determine
compatibility with the current GDF concept. This process has been developed for the arisings of legacy
wastes from the currently operating and decommissioning sites and it is envisaged will be applicable to
wastes arising from the UK ABWR, when site specific information is available.
The management of ILW streams generated from operating a single UK ABWR are likely to fall within a
number of currently applied processes, as discussed in the following Sections.
8.1.3.1. General Packaging Options
There are currently a range of different containers and associated packaging processes which could be used
for packaging these wastes in the UK. They fall into two broad categories:
1) Thin walled containers
These containers are usually manufactured from stainless steel within which a wasteform is
produced, either by addition of an immobilisation matrix or by intimate mixing of the waste,
where it is water based as a slurry, with a suitable cement based formulation to form a wasteform.
For these containers there would be a need for a shielded onsite store to shield against direct dose
arising from the packages. In this case the combination of container and wasteform produce a
package which complies with the specification requirements of RWMD (13) to ensure durability
and predictable evolution over the periods of on-site interim storage and eventual transport to and
disposal in a GDF. Alternative immobilisation matrices are also available for specific wastes; for
example certain polymers have been used for specific, often ‘niche’ or unusual wastes. For
transport, these packages will be overpacked in a shielded transport container to protect the public
from radiation dose, in compliance with the relevant transport regulations and the package from
the requirements of the hypothetical accident requirements which are defined within those
regulations.
2) Robust shielded waste containers
In these containers, the waste contents have been subject to simpler processes like size reduction
and drying without the introduction of an immobilising matrix. In this case the requirements for
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the package for interim storage, transport and disposal to a GDF are largely met by the container
alone with little credit being claimed for the wasteform inside. These containers would also
provide sufficient external shielding and integrity to comply with the relevant transport regulations
depending upon the properties of the radioactive contents. It is also envisaged that interim stores
for these packages would be much simpler, and hence cheaper, than for the other option, above, as
they would only need to be weather proof and secure, although for long periods. (14) provides a
description of how such packages could meet the requirements of the RWMD specification
requirements noted above.
For the purposes of GDA, either option would be considered as viable, subject to a satisfactory
disposability assessment by RWMD. A number of LoCs for packages comprising stainless steel containers
and cemented wasteforms have been issued by RWMD for a range of wastes which are directly analogous
to the ILW streams arising from the UK ABWR; hence adoption of this approach is based upon well
recognised and informed approaches. RWMD have also assessed two submissions for conceptual LoCs
using robust shielded containers (15) and (16). The assessments show that this type of package can be
made to be disposable provided certain specific underpinning information is submitted as summarised
below:
a. Substantiate the performance of the packages under the accident scenarios for GDF operations,
either demonstration of complete containment or provide refined Release Fraction values;
b. Understand the constraints on the water content of the waste and any associated risks due to
pressurisation of the containers;
c. Provide confidence that the functionality of the lid bolts can be maintained during storage prior to
transport, allowing the seals to be changed as necessary; and
d. Demonstrate that the requirements for Data Recording and Management Systems are understood.
Where wastes are packaged during the operational period they would then be stored on site in an ILW store,
pending the availability of a GDF. The store type (shielded or unshielded) would be determined in
conjunction with the choice of waste packaging option; whether the waste packages are unshielded or
self-shielded as noted in 1 & 2 above. Until a waste container is selected for the UK ABWR – the on-site
interim storage requirements are unknown. However, the storage requirements for the alternative package
types are briefly described in later Sections of this document.
8.1.3.2. Sludge (Crud)
Sludges are forms of wet ILW that result from the filtration of liquids, from fuel cooling pools, active
drains and from the settlement of process residues and corrosion products. They are treated in the UK
ABWR in the Spent Resin and Sludge system (SS) arising from backwashing of various filters from the
Condensate Filter facility (CF) and the Low Conductivity Waste (LCW) systems. The waste from the CF
system will be within the ILW category as it arises from the condensate system after the steam has passed
through the turbines. The particulate activity will comprise small quantities of corrosion product which
may have been carried over with the steam.
The LCW system collects wastes from various sources including the Reactor Building (R/B), Turbine
Building (T/B) and Radwaste Building (RW/B) drains, and processes them via filters, demineraliser and
sampling tanks. The sludge will arise from backwashing the filters.
The main options for packaging sludges are either to dry using heat and vacuum, and store within a robust
shielded container or solidification in a suitable cement formulation intimately mixed within a thin walled
stainless steel container to form a homogeneous monolithic wasteform. Typically for cemented wasteforms
3m
3
or 500 litre drums are applicable, using a ‘lost paddle’ mixing system. The latter method is common
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practice within the UK, although the drying method is being adopted by some waste producers (e.g.
Magnox Limited) as their analysis shows that significant savings in short term capital expenditure will
result because shielded stores will not need to be built. Hence, a site operator may wish to consider the
option of using robust shielded containers when evaluating their business case for ILW packaging and
storage.
8.1.3.3. Powder Ion Exchange Resins
The Reactor Water Clean-up System (CUW) and Fuel Pool Cooling Clean-up System (FPC) filter
demineralisers process the main reactor circuit feed water and the water used in the SF cooling pool. This
stream comprises the powder resins which are used for removal of dissolved radioactivity and will include
any particulate sludge which is filtered from the CUW and FPC systems comprising particles of metal
oxide (iron oxide and others) in the reactor water/fuel pool water. These resins are discharged, together
with any entrained particulate sludge, when the particulate capture capacity of the resin is exhausted.
The powder resins as used in Japan have a cross linked polystyrene matrix and hence it will be assumed
that a similar material will be used for the UK ABWR as this resin material type is similar to those
currently used for radioactive water clean-up in other UK reactor systems; for example, for clean-up of
cooling pool water in the Magnox reactors.
The two main options for packaging these streams is the same as for the sludges above, comprising drying
and storage within robust shielded containers or solidification in a suitable cement formulation intimately
mixed within a thin walled stainless steel container to form a homogeneous monolithic wasteform.
8.1.3.4. Higher Activity Metals
These wastes arise during operations primarily, from the Zircaloy channel boxes which surround each of
the fuel element bundles in order to contain the boiling water region. These items are approximately 4.3 m
long and 15 x 15 cm square and used to be removed from the fuel element assembly after a period of
cooling in the fuel pool. Currently, these items are stored with the spent fuel assemblies in the spent fuel
pool in Japan. The fuel element and channel box are later dispatched together for further storage prior to
reprocessing.
For the UK ABWR, for GDA, it will be assumed that the channel boxes will be consigned with the SF and
not removed from the fuel assembly on discharge from the reactor. As it will be assumed that the UK
practice is not to reprocess the SF, this option could be advantageous. It will eliminate any need to store
and process or package as ILW, with attendant benefits in reducing operator dose and saving cost and
environmental impact of any facilities needed for processing/packaging to be undertaken. Adoption of this
approach will be subject to a satisfactory disposability assessment by RWMD and future discussion with
the site operator.
The other main contribution to activated metals during the operational period are the control rods, which
are of cruciform shape metallic construction containing stainless steel tubes in each wing of the cruciform
filled with boron carbide powder. The use of Hafnium may also be employed, depending upon the control
rod design. This material has a larger neutron capture cross section and longer lifetime which may offer
advantages and reduce the quantity of waste generated from replacement and management of control rods.
They are approximately the same length as the channel boxes and the activity is currently assumed to be
the same as for the channel boxes. There is, however a smaller volume of these items arising than the
channel boxes, at 5 units/year, as the control rods remain in the reactor for several fuel cycles. Their
management will need to be discussed further with the site operator to determine the optimum approach
which satisfies an acceptable business case.
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Japanese practice for these items is also to store in a water filled bunker. For the UK ABWR the control
rods may be managed in a number of ways. The method of interim storage of these items prior to
packaging is to be decided as UK practice with gas reactors is to store activated metals in a dry, usually
shielded ‘vault’ facility. This will allow a period of decay to minimise dose to operatives and to simplify
facility requirements when packaging is eventually undertaken. The logistics of this will need further
detailed analysis by the site operator as it will also involve development of an acceptable business case for
interim raw waste storage and a packaging facility. For this option it will be assumed that the control rods
will be size reduced to cut to length and reduce voidage possibly by shearing, to minimise secondary waste
and compaction prior to loading into a suitable disposal container. The type of control rod to be used for
the UK ABWR has not yet been determined. However, if they incorporate sealed tubes containing Boron
Carbide as the neutron absorber the size reduction will need to be undertaken to avoid breaching them. If
Hafnium is used as the neutron absorber instead, then the implications of the presence of this material will
also need to be considered. The envisaged quantity of these items over the operational life is approximately
30ton; hence further consideration by the site operator will be necessary to develop an acceptable business
case for the optimum solution.
The main container options for packaging these items are stainless steel boxes of similar size (3 m
3
) to
those used for sludge/resin aqueous slurry solidification, 500 litre stainless steel drums or robust containers
which are self-shielded and are also qualified as transport containers. For the latter category there are a
number of ductile cast iron container (DCIC) products available (e.g. by Crofts
2
or GNS
3
) of different
sizes.
The items would be grouted in to the stainless steel containers but not in the robust DCIC type. The
standard RWMD 2 m and 4 m box may also be suitable for these activated items after a decay storage
period. These boxes are primarily designed for packaging of decommissioning wastes of Low Specific
Activity and hence it may be logical to store the control rods until end of operations and deal with them
during decommissioning. Within these containers the wastes may be grouted or un-grouted, subject to
disposability assessment advice from RWMD. The interim storage on site of the ‘raw’ items could be in a
dry shielded environment to allow some decay (probably in the region of one order of magnitude) prior to
packaging and disposal or under water storage could also be implemented.
A number of other items have been identified. These include various reactor core components from the
SRNM and LPRM systems. These items could be dealt with on a campaign basis during the operational
period either by grouting into a stainless steel container, as described above or loading into a robust
shielded container.
8.1.3.5. Irradiated Metal (Decommissioning)
Where irradiated metal arises during decommissioning, which is determined to be ILW, this can be dealt
with using any of the containers noted in Section 8.1.3.4, depending upon their physical (shape, size and
weight), chemical and radiological characteristics and how much processing (e.g. size reduction) is
necessary. As for other ILW streams the acceptability of the packages to the GDF will be determined via
the LoC process, as noted in Section 8.1.3.1 above.
8.1.3.6. Borderline Wastes
Where a specific waste stream is likely to be considered as ‘borderline’ in that it may be close to the LLW
and ILW categorisation boundary, these will be assessed using an agreed methodology. LLWR Ltd have
2
http://www.croftltd.com/about/news_items/Developement-of-shielded-packages.php
3
http://www.gns.de/language=en/4877
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published a guidance document (17) which contains a number of decision making factors for consideration.
This document has been developed in conjunction with RWMD and its application will be linked with the
LoC process and the waste acceptance criteria adopted by LLWR.
RWMD are currently considering several aspects related to borderline wastes which include:
? Use of decay storage for up to 300 years for some short lived radionuclides;
? Further development of the LLWR boundary wastes report (17) to better optimise the balance
between LLW and ILW;
? Specific consideration of Tritiated wastes and whether to decay store or remove the Tritium;
? Consideration, in conjunction with LLWR Ltd to manage the wastes via the Safety case rather than
purely the Waste Acceptance Criteria.
Developments in this area by LLWR Ltd and RWMD will be monitored, possibly by the site operator via
the relevant User Groups for the LLWR and the GDF, when waste management practices are being planned
for implementation, to ensure that the optimal route for specific waste streams are chosen.
8.2. Spent Fuel
The current assumption, based upon the Government White Paper ‘Meeting the Energy Challenge’ (18) for
the management of Spent Nuclear Fuel from new build is to treat it as a ‘HAW’ and not reprocess, as is
undertaken in other countries This document specifically states that ‘unless otherwise stated, references to
the Government position on waste refer to “higher activity waste”, which includes intermediate level waste
waste and spent fuel from any new nuclear power stations’. The current convention is to define HAW as
comprising all HLW and ILW, and a small fraction of LLW. Hence, the terms ‘HAW’ and ‘HAW and
spent fuel’ are likely to be used interchangeably. Therefore, following a period of cooling in the SF Pool
(SFP) the SF will be considered as a waste and will be stored and packaged appropriately and disposed of
once a GDF is available.
SF is produced only during the operational phase of the reactor. However, all of the SF stored in the fuel
cooling pool, after final shutdown, is assigned to the decommissioning phase. From a technical perspective
there is no difference between these two phases and all SF will be managed using the same strategy.
A small number of fuel assemblies may experience defects. The characteristics of these will be discussed
with RWMD as part of the disposability assessment and any specific measures to manage them agreed.
The main defects that are likely to occur during a 60 year operational lifetime are fuel rod defects caused
by debris fretting or pellet-cladding interaction, for which the mitigation technologies have been
introduced.
At the highest level there are several options available for the long term on-site storage of SF in the UK.
The following Sections describe the spent fuel characteristics from a UK ABWR and the options for
packaging, interim storage and disposal. These options are based upon good practice as currently applied
in other countries (including one option planned for implementation at Sizewell B) and which would be
considered as BAT at this time. However, the selection of the appropriate option by the operator, when
developing their plans for packaging, interim storage and disposal will be subject to the application of
BAT within their decision making process.
8.2.1. Fuel Characteristics
The GE14 fuel assembly consists of a fuel bundle (composed of fuel rods, water rods, spacers, and upper
and lower tieplates and channel fasteners), and a channel that surrounds the fuel bundle (Figure 8.2-1).
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This fuel type has 92 fuel rods (of which fourteen are partial length rods) in a 10x10 rod array and two
large central water rods. The fuel rods and water rods are spaced and supported by the upper and lower
tieplates with intermediate spacing provided by eight spacers. The upper and lower tieplates are fixed by
eight tie rods, which hold the fuel bundle together. The upper tieplate has a handle for transferring the fuel
assembly.
Figure 8.2-1 – GE14 Fuel Assembly
Based upon preliminary advice from RWMD the assumed cooling period before geological disposal is
likely to be of the order of 100 years. General information for fuel and plant operation are shown in Table
8.2-1 below. Potential fuel failure mechanisms are debris fretting or pellet-cladding interaction, for which
the mitigation technologies have been introduced.
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Table 8.2-1 General information for fuel and plant operation
Item Value
Power (MWt) 3926
Operational cycle (month) 18
Average discharged fuel burn-up (GWd/t) 50
Weight of UO2 per assembly (kg) 200*
Total number of fuel assemblies discharged over
60 years operation
9600*
*:approximate
8.2.2. Packaging Options
The SF assemblies will be stored in the SF pool in order that the heat output and activity levels can reduce.
The SF will then be transferred to another on site storage facility prior to it being re-packaged, if required,
for disposal in the GDF.
The option to store the channel boxes with the SF assemblies has been considered for minimising waste
volumes and avoiding the potential operator dose associated with removal of them and subsequent size
reduction. It is assumed for GDA that the channel boxes will remain with the SF and hence the
treatment/packaging of them will be the same as for the SF.
Management of any defective/failed fuel rods/assemblies is to be determined.
Options for the on-going management of the SF, will be subject to a satisfactory disposability assessment
by RWMD, and will include the following candidate packaging/storage options:
8.2.2.1. Dry Cask
This is current Japanese practice of removal of SF elements from the fuel cooling pool after an appropriate
period and loading a number of them into a large shielded cask, which is capable of storing the SF. This
cask could then be used for transport, after fitting of suitable high performance shock absorbers, to a
central interim storage site, if appropriate. For GDA it is assumed that the SF remains on site in a suitable
facility.
This method stores the spent nuclear fuel which has already been cooled in a fuel storage pool for a period
of time. The SF assemblies are sealed within a large leak-tight container and surrounded by an inert gas.
There are two types of cask system that can be used:
? Concrete – a number of SF elements are placed in a stainless steel canister which is welded shut
before being placed into a concrete outer casing;
? Metal – the SF elements are placed into a metal cask which is sealed and bolted shut.
Both systems are designed to provide protection against external hazards and provide adequate radiation
shielding to both the site workforce and members of the public. There are a number of dry storage
container designs in use around the world and different designs can be used for storage, transportation or
both. In some designs, the steel cylinders containing the fuel are placed vertically in a concrete cask; in
other designs the cylinders are placed horizontally. The concrete casks provide radiation shielding. Other
cask designs place the steel cylinder vertically on a concrete pad at a dry cask storage site and use either
metal or a combination of metal and concrete outer cylinders for radiation shielding.
SF assemblies would be loaded into the chosen cask system in the Reactor Building. The cask would then
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be sealed and decontaminated before being taken to the dry cask storage facility which would provide
weather protection.
The SF elements would need to be prepared for transport to and disposal in a GDF by removal from the
cask storage into another container which would be suitable for transport and acceptable for disposal. Such
a container may be as described in 8.2.2.4 below.
8.2.2.2. Multi-Purpose Container
To avoid the need for repackaging the SF into a disposal container, after a period of storage for cooling (for
example in a cask storage system, as described above) the concept of using a Multi-Purpose Container
(MPC) has been proposed. This would involve the packaging of a smaller number of SF elements into a
container which can be used for extended on site storage and subsequent disposal to a GDF.
There are systems available for this which are currently used in the USA. For example, the Holtec system,
similar to that shown in (19) has been proposed for Sizewell B PWR SF, which accommodates 24 elements
in a single MPC. Use of this container has been assessed by RWMD (20) which concluded that use of this
MPC was not optimum when taking into account all of the management stages to the GDF and that re
packaging for disposal may still be necessary. The current situation is that RWMD plan to propose an
MPC
4
which would be suitable for on-site interim storage and subsequent disposal when a GDF is
available. This MPC will then be considered for use with the UK ABWR SF and hence, use of this system
is considered to be a viable option, subject to the RWMD disposability assessment.
8.2.2.3. Modular Vault Dry Storage System
Dry vault storage is a method, already used in many other countries, for storing spent nuclear fuel that has
already been cooled in the SF pool. In this concept individual fuel assemblies are stored within sealed
channels within the shielded vault. Current examples of these vaults have been built in such a way that
additional storage vault capacity can be constructed when needed as new modules.
After removal from the reactor the SF assemblies would be loaded into a flask in the Reactor Building. The
flask would then be sealed and decontaminated before being taken to the vault store. Here, the fuel
assemblies would be lifted from the flask and dried. A shielded fuel handling machine would then transfer
the fuel assemblies to the allocated storage channel in the vault, which would then be filled with inert gas
and sealed.
The SF assemblies are stored vertically within the channels and cooled by natural circulation. This acts as a
self-regulating system in which higher SF temperatures cause increased airflow through the vault, thereby
increasing heat removal.
The SF elements would need to be prepared for transport to and disposal in a GDF by removal from the
store into another container which would be suitable for transport and acceptable for disposal. Such a
container may be as described in 8.2.2.4 below.
8.2.2.4. KBS-3 Container
The reference conceptual design assumption for spent UK PWR fuel is that it is packaged in a KBS-3V
type canister, as used by SKB in Sweden (21) and this is considered to be a viable option for UK ABWR
SF with suitable modifications to take account of the different SF element dimensions and other
characteristics.
For PWR SF use of this concept is currently subject to the assumed temperature constraint applied to the
4
RWMD plan to issue a report in due course which will be referenced in this document.
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inner bentonite buffer surface within the GDF and that a canister contains four SF assemblies. If all the SF
were to be subject to the maximum burn-up considered, which is 65 GWd/tU, then the SF might need of
the order of 100 years of cooling in interim storage before it could be disposed of in a GDF. This aspect
will be discussed with RWMD during the disposability assessment for the UK ABWR where it is likely
that a more realistic case for cooling time would be developed which may include (for example) a mixture
of SF elements within a single container whose burn-ups would not all be at the stated maximum.
This container may be used for disposal of SF elements after a period of interim storage in any one of the
previously described storage systems, unless the MPC system is one where the MPC itself would be
acceptable to RWMD for disposal to the GDF.
Hence, use of the KBS-3 container is considered to be a viable option, subject to the RWMD disposability
assessment.
8.2.2.5. On-Site Fuel Pool Option
On-Site Fuel Pool Storage within an independent building is a method, already used in many countries, for
storing spent fuel that has been stored in the SFP in the Reactor Building. On-Site Fuel Pool Storage is the
method that is adopted in UK-EPR and is a viable option to consider for UK ABWR. The characteristics
are that cooling performance is high and, the condition of the spent fuel can be confirmed by visual
inspection from above the water, which will provide the necessary shielding.
After the period of storage for cooling in the SFP in the Reactor Building, the spent fuel would be transfer
to a separate On-Site Fuel Pool by a transport cask. The spent fuel would be stored in racks in the SFP. The
spent fuel elements would need to be prepared for transport to and disposal in a GDF by removal from the
pool into another container for transport and the package would be made acceptable for disposal. Such a
container may be as described in para. 8.2.2.4 above.
The use of this option will require cooling and clean-up equipment for managing the SF pool water quality.
8.3. Summary of Management Options
The available management options for each waste category and the SF are summarised in Table 8.3-1
below. These are based upon existing practices within the UK for legacy sites (e.g. Magnox, Sellafield,
Dounreay) and some operational sites (e.g. Sizewell B) and from international experience, especially for
the management of SF. The options have not been selected by a rigorous BAT process but are assumed to
be so at this stage of GDA because of their current applications, as noted above.
Further details of each option are included in Appendix D.
Table 8.3-1 Summary of Waste and Spent Fuel Streams
No. Title Category Form Management options
1 Dry active
waste
VLLW Solid Recycle metals where practicable, compaction,
where possible and direct disposal of remainder
to permitted disposal site
2 HVAC Filters LLW Solid To be determined, depending upon filter details
utilising a combination of compaction and
cement grouting
3 Bead resin LLW Wet Solidify with polymer or cement formulation in
disposal drum, typically 200 litre
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No. Title Category Form Management options
4 Concentrates LLW Wet Solidify with cement formulation in disposal
drum, typically 200 litre
5 Miscellaneous
combustible
LLW Solid ? Incineration (off site)
? Compaction within a drum
6 Miscellaneous
non-combustibl
e
LLW Solid ? Size reduce and grout within a disposal
drum
? Compaction within a drum
7 Sludge (crud) ILW Wet ? Dry and store within robust shielded
container
? Encapsulate in cement formulation
? Encapsulate in polymer formulation
8 Powder resin ILW Wet ? Dry and store within robust shielded
container
? Encapsulate in cement formulation
? Encapsulate in polymer formulation
9 Higher activity
metals –
control rods
ILW Solid
Store on site in dry shielded facility OR
Size reduce and:
? Dry and store within robust shielded
container
? Place in disposal container (typically 3m
3
box) and grout with cement formulation
? Channel boxes stay with SF
10 Higher activity
metals –
channel boxes
ILW Solid
11 Higher activity
metals – others
ILW Solid
12 Contaminated
and irradiated
metal and
concrete
LLW Solid
? Size reduce and place in disposal
container (typically 2m or 4m box) and
grout with cement formulation
? Decommissioning ILW only – may also use
robust shielded containers; DCIC or 4 m/2
m box
13 Irradiated metal
(from
decommissioni
ng)
ILW
Solid
14 Spent Fuel HLW Solid ? Dry cask storage with future packaging for
disposal
? Multi-Purpose Container storage for future
disposal without repackaging
? Dry vault storage ( Modular Vault Dry
Store)
? Package and store in SKB type container
? Extended storage in on-site fuel pool
9. Waste Management Infrastructure
A number of facilities will be required to conduct the various activities of waste management, conditioning,
packaging and storage of each of the waste and SF categories described above. The detailed descriptions of
each facility will depend, to some extent upon the conditioning and/or packaging method chosen for each
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specific waste stream.
The infrastructure requirements will be developed by Hitachi-GE, including consultation with Horizon,
during the initial phase of GDA so that any initial site specific requirements can be captured.
Waste treatment/packaging and storage requirements will be described for each stream of waste and SF
comprising:
9.1. Waste Treatment and Packaging Facilities
9.1.1. Low Level Waste
The LLW facilities are likely to be the same as is currently employed in Japan, subject to evaluation in
detail for compliance with UK practices and disposal criteria.
Aqueous slurry form wastes, such as ion exchange resins may be packaged in a facility, which can mix the
wastes with a suitable cement formulation to form a monolithic product of adequate mechanical properties
within a 200 litre drum. These drums would then be loaded into half height ISO containers for transport to
the disposal facility where the remainder of the space within the container would be grouted, prior to
emplacement within the repository.
Solid waste which required grouting within a drum, for example solid and metal objects which are not able
to be recycled or reused, may be packaged within a facility which can prepare a suitable grout infiltration
matrix (usually cement based), which can be delivered to each drum to form a monolithic product of
adequate mechanical properties.
These facilities are likely to be fixed, as is the current practice in Japan to ensure that dedicated plant is
available when required.
Where other processes, such as compaction or incineration are required for conditioning the waste it is
assumed that the facilities will be available off site, for example via the LLWR Ltd supplier services. In
this case a facility for drumming wastes and loading into full height, reusable ISO containers will be
provided so that they can be filled at site and dispatched to the disposal facility or a service supplier to
conduct the required processes.
9.1.2. Intermediate Level Waste
The facilities for treatment and packaging the ILW streams will depend upon the type of package chosen
from the two main options described in 8.1.3.1 above. For each package type the following are the main
facility requirements, which are described in more detail in Appendix D:
1) If thin walled stainless steel containers are to be used for immobilisation of the wastes in a cement
grout or formulation the shielded facility would need the following:
a. A facility which connects the container to the waste accumulation facility to allow retrieval
and loading the container with waste;
b. A means of introducing either a cement grout (for solid wastes) or a cement powder blend
(for aqueous slurry wastes) to encapsulate or solidify, respectively the solid or slurry waste;
c. An area for product curing to allow the wasteform to achieve adequate strength for
movement to the ILW store;
d. A transfer route to the ILW store either via a shielded transport vehicle or a shielded route
where the packaging facility is adjacent to the store.
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2) If robust shielded containers are to be used for direct packaging the following will be required:
a. A shielded facility which connects the container to the waste accumulation facility to allow
retrieval and loading the container with waste;
b. A means of size reduction (for solid wastes) or pump transfer (for aqueous slurry wastes) and
loading of the wastes into the containers;
c. A means of drying the wastes which may be a process external to the container for solid
wastes or in situ for the aqueous slurry waste. For the latter, standard vacuum drying
equipment is available which requires in situ heating of the waste whilst pulling a vacuum
and extracting the off gas via a suitable vent system, which may need a means of preventing
any volatile or soluble activity from entering the environment;
d. A transfer route to the ILW store which will be a simple standard transport vehicle capable of
carrying the load, but with no shielding required as this is supplied by the container.
9.1.3. Spent Fuel
The facility requirements for processing/packaging of SF will depend upon the management option chosen.
Whatever option is chosen there will be a common process required to dry the items. For the dry cask and
MPC options above this would need to be undertaken before loading into the relevant container, whilst for
the dry vault storage option drying would be undertaken after transport to the storage facility.
Apart from a drying facility, no other major dedicated facilities would be required; the standard reactor
design is expected to be adequate as it is assumed to contain equipment for loading SF elements into a cask
or other container.
At the end of interim storage, the current plan assumes that the SF would be removed from the dry storage
casks and transferred into final disposal containers. Such an operation, many decades after reactor closure,
would require a dry transfer cell facility to move spent fuel assemblies from the interim dry storage casks
to the final disposal containers.
At the time such a plant is needed, the site should have been fully decommissioned apart from the dry cask
spent fuel facility, and the ILW store.
9.2. Waste Stores
9.2.1. Low Level Waste
In order to accumulate a reasonable number of packages to make a consignment to the disposal facility, an
on-site buffer store will be required. The size of the store has been selected as two years’ worth of LLW
arisings to allow for disruption to the transport system or problems at the disposal facility which could
restrict or curtail the acceptance of waste consignments. The store will be located within the existing
Radioactive Waste treatment Building (RW/B) or other storage facilities and will be connected to the main
HVAC, utilities, drainage and other systems. Hence, for GDA it is assumed that there are no new facility
requirements, discharge points or additional doses arising from storage requirements for LLW.
9.2.2. Intermediate Level Waste
At this point in time Hitachi-GE are yet to determine the most appropriate method for storing the ILW on
the Generic Site for the purpose of GDA. However, the ILW will be stored on site in facilities, whose
characteristics will be suited to the specific packages/containers which are stored in them; for example
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whether they need to be shielded or unshielded. Such stores will be built to best UK practice standards and
guidance to ensure minimal impacts in respect of discharges and doses to the environment.
If the packages utilise thin walled stainless steel containers the store will need to be built to provide
adequate shielding and include all of the facilities necessary for handling the packages and for monitoring
their evolution over time. The use of remotely operated equipment is likely to be required for these
operations. If self-shielded containers are used (e.g. DCICs) then the quantity of shielding will be
significantly reduced. However, the store will require adequate facilities for handling the packages and
monitoring their evolution, although this is likely to be conducted by simpler, ‘hands-on’ techniques
because the dose rates will be designed with such in mind and taking account of ALARP considerations.
The NDA have produced a Guidance document for Packaged ILW stores which is directed at the NDA’s
legacy sites (22). This guidance could also be applicable where similar stores are necessary on the UK
ABWR sites.
The guidance includes:
? An integrated approach to the overall storage system by considering package and store
performance as a whole;
? Identification of good practice, principles, approaches and the use of ‘toolkits’ for specific
analyses; e.g. package evolution and performance models;
? Requirements for package care and management, identifying ideal, tolerable and failing
performance parameters.
9.2.3. Spent Fuel
The store arrangement for SF will depend upon the option chosen for its management, as noted above. This
facility will need to take account of the container/cask size and arrangements within the store, cooling and
shielding requirements and criticality and other safety requirements.
The SF assemblies would initially be kept in the existing fuel pool for a period to be determined to allow
them to cool down before being loaded into the chosen container system. The container would then be
sealed before being taken to the new storage building. The SF may need to be re-packaged for disposal
depending upon the option chosen for interim storage.
For the dry cask option the storage location would consist of a receiving area with rail/road-connection, a
cask maintenance platform (e.g. for seal replacement) and one or more cask storage areas, each equipped
with an overhead crane. Prefabricated concrete structures, in combination with reinforced poured concrete,
are flexible in terms of extension. Operation of the store would be totally passive and the facility would be
able to operate totally independently of the reactor site (after shutdown).
For each of the SF management options discussed in Section 8.2.2 the following are the main requirements
for long term storage:
1) Dry Cask
These casks would be capable of storing a number of different inner containers. The storage
requirements for these casks are likely to consist only of a weather proof facility which includes
adequate security and the ability to inspect the external surfaces of the casks. Further details of the
type of facility required are to be discussed at a later stage.
2) Multi-Purpose Container (MPC)
These containers are intended to be utilised for interim storage and eventual disposal and could hence
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be stored within the above noted dry casks for an interim period. The size of the MPCs will need to be
optimised to accommodate a sufficient number of SF elements whilst also allowing for eventual
transport and disposal at the GDF. As discussed in Section 8.2.2.2 above RWMD are currently
preparing some proposals on this aspect. The interim storage requirements, however, will be the same
as for the Dry Casks, as noted above.
3) Modular Vault Dry Storage System
This system is a ‘stand-alone’ store where the SF elements are stored in sealed containers within a
naturally cooled vault system, as described in Section 8.2.2.3 above. This store would require its own
facilities such as ventilation, security, discharge points and utility connections.
4) KBS-3 Container
This container is similar to the MPC described above with the exception that it accommodates a
smaller number of SF elements and hence its disposability is likely to be acceptable to RWMD,
subject to their assessments. As for the MPC these containers would be stored in Casks, for which the
interim storage requirements are as described above.
5) Extended Storage in On-site Fuel Pool
During the operational period the usual storage arrangements will apply. After the plant has ceased
generating storage in this separate facility will be an extension of the arrangements in place during the
operational period.
It is noted that Sizewell B has commenced building a dry SF store, which is due to be operational in 2015.
Hence, the experience gained from this project is likely to be of benefit, in terms of learning from
experience, if similar stores are to be built for the UK ABWR.
9.3. Waste Packages, Transport Systems and LoCs
RWMD have developed and published a number of specifications and guidance documents (13) which
define the requirements for a range of standard waste packages. These are largely based upon the use of
thin walled stainless steel containers within which a wasteform is created by combination of the waste
either with a cement powder blend for water based slurries or infiltration of a cement grout for solid waste
items. These specifications are based upon a GDF design and waste packaging proposals submitted by the
waste producers to RWMD are assessed against this design using a well-developed LoC process. This
process has three stages for package endorsement comprising ‘conceptual’, ‘interim’ and ‘final’ stages; the
last stage providing evidence of disposability to the ONR to enable them to grant a Licence Instrument to
the waste producer for implementation of waste packaging. These packages would need to be transported
within a reusable shielded transport container. This combination is usually classed as a Type B transport
package under the IAEA Transport Regulations.
As a number of waste producers have chosen to utilise the ‘robust shielded waste package’ concept, which
proposes the use of ductile cast iron containers for packaging wastes, where size reduction and drying are
used instead of waste immobilisation in cement, RWMD have also issued a Technical Note (14) which
provides guidance as to their use and acceptability in a GDF. As well as being disposal packages, these are
designed to qualify as transport packages in their own right without the need for additional outer packaging
to provide radiation shielding. They will be classed as Type A or Type B depending upon which container
is used and the characteristics of the contents.
In order to assess waste packaging arrangements proposed during the GDA process RWMD are using a
‘pre-conceptual’ assessment process, based upon the LoC process, which is designed to provide a
Disposability Assessment which is sufficient for the GDA process to show that waste packages generated
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by the subject design will be disposable.
10. Development of an Integrated Waste Strategy
The requirements for an Integrated Waste Strategy (IWS) have been developed by the NDA (23) for the
legacy wastes where its applicability is directed at operational sites or those undergoing decommissioning.
A successful IWS document will provide the following:
? Clearly identify what wastes need to be managed (both radioactive and non-radioactive);
? Describe how they are managed now and in the future, taking account of the full waste lifecycle;
? Demonstrate how the strategy is delivering against national policy and strategy;
? Sign post to key underpinning and justification information in an effective and accessible way
(including UK Radioactive Waste Inventory, TBURDs, etc.);
? Identify future problems and/or gaps, and the solutions to address them;
? Drive improvement by focusing on key outcomes;
? Show how the strategy will be implemented and how it factors into business decisions;
? Be an integral part of meeting regulatory requirements for waste management;
? Be available and accessible to people.
All of the above requirements will not be applicable at the GDA stage and those which are relevant are
incorporated within this arrangements document (for example much of the first two bullets above). These
arrangements do, however allow for an IWS to be subsequently developed at a later stage, for an
operational site, which will then be subject to periodic review and update.
11. Plan for Disposability Assessments
11.1. Introduction to the RWMD Disposability Assessment
Hitachi-GE is required to obtain a view from RWMD as to whether the HAW can be disposed of in line
with any plans for a GDF in the UK. As part of obtaining that view Hitachi-GE need to provide RWMD
with a suite of information to allow them to undertake their assessments. At this moment in time the
assessments are underway therefore no details are provided on the conclusion. The next issue of this
document will include such information.
The interactions with RWMD have been planned to achieve the submission of a disposability assessment
(by RWMD) to the Environment Agency, to comply with the requirements for Justification of the use of
the ABWR technology in the UK.
The information to be provided to RWMD will cover all HAW and SF for which there is currently no
disposal facility available and, for the latter where reprocessing is no longer to be undertaken. The HAW
will comprise Intermediate Level Wastes (ILW) as it is assumed that there will not be any vitrified HLW or
LLW which is unsuitable for disposal at the Low Level Waste Repository.
All relevant information required by RWMD will enable their assessment of the envisaged packages to be
conducted for suitability for disposal in a GDF. At this stage several options for packaging the SF for
interim storage and eventual disposal will be included because the preferred method has not yet been
defined. The final choice will involve many other aspects related to specific site requirements, including
interim storage arrangements, the need for long term cooling prior to disposal, etc.
For the purposes of the submission of information to RWMD, a number of assumptions will be included.
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11.2. Hitachi-GE Plan for the RWMD Disposability Assessment
The information to be provided to RWMD will include the following:
? Details of the characteristics of the waste and any proposed conditioning processed, (e.g. drying,
cementation, size reduction etc.) will be provided. These details will be supported by relevant
references from Hitachi-GE, which will include the physical, chemical and radionuclide properties
envisaged for the wastes together with quantities envisaged to arise during operation and from
future decommissioning activities;
? Specific parameters of the SF and any associated packaging which will be relevant to its condition
at the time of disposal will be addressed. The relevant parameters required to inform the
disposability assessment are also included;
? A number of packaging and interim storage options are described using a number of standard
containers, as currently used in the UK and known to RWMD. The means whereby each of the
wastes and SF may be rendered into a passively safe wasteform, within a disposable package, will
be outlined in concept, based upon current UK practice and experience of similar wastes and SF
arising on other UK sites;
? The ABWR ILWs will be compared to other UK ILWs, which have already been assessed and
found to be disposable by issue of final stage Letters of Compliance (LoC) to support the case that
the envisaged waste packages will be disposable.
In order to ensure that the information required by RWMD is as complete as possible and made available
in a timely manner to enable a full assessment to be completed, regular interactions during the provision of
the submission information and the subsequent assessment period will be undertaken.
The interactions between RWMD and Hitachi-GE/Horizon will include how their project addressing the
disposability and full life cycle implications of high-heat generating UK wastes, which includes a wide
range of UK wastes, is developing and how the approach for UK ABWR may be influenced.
A number of assumptions are included in the plan to focus the scope of the disposability assessment to
address GDA requirements for the current phase of the project.
12. Conclusions
This arrangements document sets out management arrangements for processing, interim storage and where
facilities are available the disposal of waste and SF generated by the UK ABWR in accordance with the
UK Government policy and regulatory constraints. The UK ABWR provides a high degree of confidence
that the challenges associated with the management of solid waste and SF are fully understood and that
solutions are available with the envelope of current UK and international experience. Specifically it
demonstrates that:
? The Waste Hierarchy will be applied and the production of radioactive waste will be avoided and
where this is not reasonably practicable, the quantity of waste produced will be minimised;
? The radioactive wastes generated by the UK ABWR are similar of those wastes generated by
operating ABWRs and all waste streams have process routes to interim storage or final disposal
solutions;
? A number of waste and SF management options which are based on nationally or internationally
proven technologies and the experience from ABWR projects are available for the UK ABWR.
The base case waste treatment facilities for each UK ABWR site include:
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? Waste Treatment Building for the receipt, segregation, treatment and conditioning of solid
radioactive wastes;
? Interim Storage Facility for solid ILW packages;
? Interim Storage Facility for receipt, packaging and safe interim storage of SF.
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Appendix A – Consideration of the REPs
The following is a list of the REPs listed in (2) which are considered relevant to Radioactive Waste
Management Arrangements & Disposability. Each REP has been addressed in turn and a comment added
to indicate where it has been considered in this arrangements document.
Table A-1 Compliance of the Document with the REPs
REP Description Comment
RSMDP1 Radioactive Substances Strategy This document
RSMDP5 Actions having Irreversible
Consequences
This document outlines the thorough and
considered philosophy underpinning all
radioactive waste management decisions. The
application of BAT will ensure a robust decision
making process
RSMDP6 Application of BAT The proposed practices detailed in this
document are considered to be BAT and the
decision making process for future
implementation will be informed by the
application of BAT
RSMDP8 Segregation of Wastes Covered in the arrangement of waste
accumulation facilities and that each waste is
packaged separately with no mixing. The
application of the Waste Hierarchy is
emphasised throughout the document
RSMDP9 Characterisation Covered in principles and packaging options to
ensure adequate knowledge of the waste
properties and that the most appropriate
management technique can be applied
RSMDP10 Storage Passive safe storage for packaged waste is
covered as a principle and mentioned
throughout the document
RSMDP11 Storage in a Passively Safe State Concentrating waste and relevant conditioning
technology to deliver a solid waste form is
outlined. The requirements for passive storage
and adequate management of waste packages
are covered.
RSMDP14 Record Keeping Requirements are covered for all radioactive
wastes e.g. RWMD requirement for packaged
HAWs.
RSMDP15 Requirements and Conditions for
Disposal of Wastes
Reference is made to WACs for VLLW and
LLW managed via LLWR Ltd and LoCs for
ILW managed via RWMD
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REP Description Comment
ENDP4 Environmental Protection Functions
and Measures
The defined management approaches for waste
conditioning and storage and consignment to
other facilities will ensure adequate
environmental protection
ENDP8 Ageing and Degradation Not explicitly covered for plant systems.
However, adoption of BAT will lead to
implementation of this. The RWMD LoC
process, which is referred to, requires the
longevity of waste packages and wasteforms
with predictable evolution
ENDP 15 Mechanical Containment Systems for
Liquids and Gases
This is applicable to raw waste accumulation
facilities which are referred to. Adoption of
BAT will lead to implementation of this for
waste packaging
ENDP16 Ventilation Systems The HVAC filters are included as the solid
waste stream which will arise from these
systems.
DEDP1 Decommissioning Strategy Waste management aspects only are covered.
Decommissioning Strategy is covered in a
separate document
DEDP2 Decommissioning Plan Waste management aspects only covered.
Decommissioning Plan is covered in a separate
document
DEDP3 Considering Decommissioning during
Design and Operation
Waste management aspects only covered.
Considerations during design and operation is
covered in a separate document
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Appendix B - Waste Hierarchy
The Waste Hierarchy as illustrated in Figure B1 below aims to encourage the management of waste
materials in order to reduce the amount of waste materials produced, and to recover maximum value from
the wastes that are produced. It is not applied as a strict hierarchy as many complex factors influence the
optimal management for any given waste material. However, as a guide, it encourages the prevention of
waste, followed by the reuse and refurbishment of goods, then value recovery through recycling.
The next option is energy recovery, an important level in the hierarchy as many materials have significant
embedded energy that can be recovered. Waste prevention, reuse, recycling and recovery are collectively
defined by the Organisation for Economic Co-operation and Development (OECD) as waste minimisation.
Finally, waste disposal should only be used when no option further up the hierarchy is possible.
Figure B-1 - Waste Hierarchy
5
5
SEPA Briefing Note - Ozone Depleting Substances (ODS) and the Construction Industry; Version 1; July
2008
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Appendix C - Waste Stream and Spent Fuel Descriptions
1. Introduction
This Appendix provides currently available details of the wastes and SF which it is envisaged will be
generated during the operational life of the UK ABWR and the details of the assumed decommissioning
wastes which will arise at the end of life. Where additional details are required this is shown as a To Be
Advised (TBA) item.
2. Nature and Quantity of the Wastes and Spent Fuel
A number of reports have been produced which describe the wastes envisaged to arise from the UK ABWR.
Specifically (24) provides a description of the various waste treatment systems currently used in Japan and
which are likely to form the basis of systems for use in the UK. (25) provides a brief summary of all of the
solid wastes which are envisaged to arise from operation and decommissioning of the UK ABWR. The
arisings of spent filters from various HVAC systems are also included in this document.
(26) has been produced to provide more realistic inventories which are relevant to waste management and
disposal. The data within this report excludes a number of radionuclides whose half-lives are too short to be
relevant for short term LLW or much longer term ILW disposal. The report identifies 19 radionuclides
which are considered relevant for L2 disposal in Japan. As this category is approximately equivalent to the
lower activity range specified in the UK for ILW this list of radionuclides will be used for the initial
inventories for LLW and ILW for the GDA.
The following tables summarise the information for each category of waste envisaged to arise from the UK
ABWR during its operational life (24), (25) and (26) and from decommissioning activities (25).
Information is also included for the SF arisings during the operational period.
The information provided for activities has been updated in line with (26) to include the main radionuclides
contributing to the activities. This information is included in the following summary tables. The activity
levels have also been adjusted to include only the radionuclides with a half-life greater than 3 months.
The wastes are categorized as concentrated liquid waste (by an evaporator), waste sludge from filters, spent
resins from the demineralisers and miscellaneous solid wastes which are activated or contaminated.
Further general information of the various systems within the UK ABWR is similar to that described in
(27).
Liquids and gases which are discharged after abatement are not included in this document. Wastes which
can be treated and disposed of to current licensed disposal facilities (e.g. the LLWR or licensed landfill site)
are included. Wastes which currently have no available disposal facility, which comprise ILW and SF
(assumed to be waste and not subject to reprocessing) are included, which will be consigned to the GDF
when it is available.
2.1. Operational VLLW
This category currently includes very few items and it is likely that further will arise during operations. Any
items arising will be managed according to the Waste Hierarchy and it is currently assumed that the
services offered by LLWR Ltd will be employed to optimise the management of the wastes and any
disposals necessary.
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Table C2.1-1 Details of VLLW Miscellaneous Solid Items
Parameter Description
Waste Origin Arising from operations and maintenance
Waste physical/chemical description Low activity piping, motors and heat insulators from
maintenance activities
Key parameters for conditioning Solid items
Nature of radioactive material Low levels of contamination
Main radionuclides/activity No details available
Annual operational arising No details available
Total operational arising No details available
Hazardous substances None
2.2. Operational LLW
Note that the LLW limit is 1.20E+04 Bq/g for beta/gamma and 4.00E+03 Bq/g alpha activities and that the
following activities are for un-processed/packaged wastes. Depending upon the packaging technique used
the stated specific activities for wastes, to consign for disposal, may decrease; for example where the best
technique for packaging is to mix with a cement formulation. Where a technique is chosen which may
concentrate the waste, and hence the total activity may move into the ILW range this factor will be taken
into account when choosing the best technique.
The following details are based upon wastes arising from current Japanese sites and there may be some
difference for operational plants in the UK. For example the number and type of HVAC filters may be
different for UK ABWR, depending upon safety analysis and the criteria which are used for filter
replacement. The values used are based upon Japanese experience, where the recommendations of the filter
manufacturer are used. This will be a bounding case for the number of filters arising.
Table C2.2-1 Details of LLW HVAC Filters
Parameter Description
Waste Origin Arising from filter changing in air treatment facilities:
? Reactor Building exhaust
? Turbine Building exhaust
? Turbine Building high radiation exhaust
? Radwaste Building exhaust
? Service Building exhaust
Waste physical/chemical description Medium efficiency filters (bag type) - 6.3 kg/filter
HEPA filters – 13 kg/filter
Key parameters for conditioning Solid items comprising filter media and support frames
Nature of radioactive material Low levels of contamination
Main radionuclides/activity (Bq/g) Co-60, Ni-63, Fe-55, Ag-110m; 1.3E+03
Annual operational arising Medium efficiency filters – 1.9 ton/year
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HEPA filters – 0.3 ton/year
Total operational arising Medium efficiency filters – 114 ton
HEPA filters – 18 ton)
Hazardous substances None
Table C2.2-2 Details of LLW Bead Resins
Parameter Description
Waste Origin Arising from the CD, the LCW and HCW demineraliser
Systems. The CD removes radionuclides from the main
reactor coolant water after is has passed through the
condenser. The origin of the LCW is as described above. The
HCW comprises wastes with a higher particulate content and
clean-up systems may include filtration, evaporators and
demineralisers. Bead resins are used in the demineralisers to
extract soluble radionuclides prior to reuse or discharge of the
liquors
Waste physical/chemical description LCW & HCW Demineraliser; Cation Exchange Bead Resin,
Anion Exchange Bead Resin. These resins have a cross linked
polystyrene matrix
CD; Cation Exchange Bead Resin Anion Exchange Bead
Resin. These resins have a Styrene divinylbenzene copolymer
matrix
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, incineration, compaction
Nature of radioactive material Soluble species arising from the noted systems; including
soluble activated corrosion products and fission products
Main radionuclides/activity (Bq/g) CD – Co-60, Ag-110m; 3.8E+03
LCW demineraliser – Co-60, Mn-54, Ag-110m; 2.6E+04;
HCW demineraliser – H-3, C-14; 7.9E+02
Annual operational arising 14.5 m
3
/year
Total operational arising ~880 m
3
/60 years
Hazardous substances None
Table C2.2-3 Details of LLW Concentrated Waste
Parameter Description
Waste Origin Arising from the HCW evaporator and stored in the CONW
tanks for a period of decay before processing
Waste physical/chemical description Concentrates from various sources containing organic and
inorganic particulate and soluble species
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, compaction
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Parameter Description
Nature of radioactive material Particulate and soluble species arising from the noted systems
Main radionuclides/activity (Bq/g) H-3, Co-60, Ag-110m; 1.3E+02
Annual operational arising 2.4 m
3
/year
6
Total operational arising ~150 m
3
/60 years
Hazardous substances None
Table C2.2-4 Details of LLW Miscellaneous Combustible Waste
Parameter Description
Waste Origin Dry active waste generated through routine and maintenance
operations. The waste consists of CF and LCW spent hollow
fiber filter membrane, polythene (sheet and bag), paper, wood,
cloth (wipes and gloves), rubber gloves, spent active carbon
Waste physical/chemical description Heterogeneous non-metallic solid dry wastes, apart from the
oil waste
Key parameters for conditioning All wastes could be incinerated, if appropriate. A significant
proportion would be compactible, if that route was required
Nature of radioactive material Particulate contamination
Main radionuclides/activity TBA
Annual operational arising CF spent hollow fibre media; 2.2 m
3
/year
LCW spent hollow fibre media; 0.06 m
3
/year
Spent active carbon; depend on maintenance
Miscellaneous; depend on maintenance
Total operational arising TBA
Hazardous substances None
Table C2.2-5 Details of LLW Miscellaneous Non-Combustible Waste
Parameter Description
Waste Origin Arises during maintenance operations within the nuclear
island. The waste mainly consists of contaminated RCA
plant. Comprising metal plate, pipes, bulk, cables, lagging
material, gas filters, concrete and glass
Waste physical/chemical description Metallic and other solids
Key parameters for conditioning Dry largely non-compactible
Nature of radioactive material Particulate contamination
Main radionuclides/activity (Bq/g) Co-60, Ag-110m; 3.0E+03
Annual operational arising Depend on maintenance and replacement
6
Design basis value; actual close to zero
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Parameter Description
Total operational arising TBA
Hazardous substances None
2.3. Operational ILW
The following streams are those whose activities are envisaged to be greater than the LLW activity limits of
1.20E+04 Bq/g for beta/gamma and 4.00E+03 Bq/g alpha and are not heat generating. The stated activities
are for the raw unprocessed wastes and the specific activities will change depending upon the packaging
technique used and may also be influenced by radioactive decay of some radionuclides, in particular those
which have a short half-life. For GDA it will be assumed that the categories will remain the same for
conditioned and unconditioned wastes. However, if during site operations some ILW streams are
re-categorised to LLW the application of relevant LLW management techniques will ensure that the streams
are adequately dealt with. No issues are envisaged with being able to treat any ILW stream as LLW, subject
to the noted activity limits.
Table C2.3-1 Details of ILW Filter Sludge (Crud)
Parameter Description
Waste Origin Arising from backwashing of CF and LCW filters. LCW
arises from the drain sumps of the various components in the
R/B, the drywell, the T/B and the RW/B. It may also be
collected in low-conductivity waste collecting pools
Waste physical/chemical description Corrosion product Fe2O3, Fe3O4, FeOOH
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, compaction
Nature of radioactive material Corrosion product particulate slurry and some dissolved
species
Main radionuclides/activity (Bq/g) CF – Co-60, Mn-54, Fe-55, Zn-65; 1.1E+05
LCW filters - Co-60, Mn-54, Fe-55, Zn-65, Ag-110m;
4.5E+05
Annual operational arising 1.9 m
3
/year
Total operational arising ~114 m
3
/60 years
Hazardous substances None
Table C2.3-2 Details of ILW Powder Resin
Parameter Description
Waste Origin Arising from the CUW and FPC filter demineralisers.
Waste physical/chemical description CUW Cation Exchange Powder Resin. ANION Exchange
Powder Resin. These resins have a cross linked polystyrene
matrix.
This stream also includes any particulate sludge which is
filtered from the CUW and FPC systems
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Parameter Description
Key parameters for conditioning In aqueous slurry form. Possible candidate for cementation,
drying, incineration, compaction
Nature of radioactive material Soluble species arising from the noted systems; including
particulate and soluble activated corrosion products and
fission products
Main radionuclides/activity (Bq/g) CUW filter demineraliser – Co-60, Mn-54, Fe-55, Zn-65,
Ag-110m; 3.6E+07
FPC filter demineraliser – Co-60, Mn-54, Fe-55, Zn-65,
Ag-110m; 1.3E+07
Annual operational arising 4.9 m
3
/year
Total operational arising ~300 m
3
/60 years
Hazardous substances None
Table C2.3-3 Details of ILW Higher Activity Metals – Control Rods
Parameter Description
Waste Origin Cruciform shape and inserted between each group of 4 fuel
elements external to the channel boxes and hence within
non-boiling water. Performs the dual functions of power
distribution shaping and reactivity control
Waste physical/chemical description Metallic construction. The control rods may contain Boron
carbide or Hafnium as alternative materials for reactivity
control
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual operational arising 5 units; (0.5 ton; approx. 1.5 m
3
)/year
Total operational arising 30 ton (approx. 90 m
3
)/60 years
Hazardous substances None
Table C2.3-4 Details of ILW Higher Activity Metals – Channel Boxes
Parameter Description
Waste Origin Channels which contain each fuel element within the reactor
core to direct the coolant flow and contain the boiling regions
Waste physical/chemical description Zircaloy box which surrounds the fuel bundle. Approx. 4.3 m
long and 15 x 15 cm square
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual operational arising Approx. 170 units; (6.8 ton; approx. 34 m
3
)/year
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Parameter Description
Total operational arising 408 ton/60 years
Hazardous substances None
Table C2.3-5 Details of ILW Higher Activity Metals – Reactor Components, etc
Parameter Description
Waste Origin Various reactor core components.
Waste physical/chemical description Start-up Range Neutron Monitoring system (SRNM including
dry tubes), Local Power range Monitoring System (LPRM),
Traversing In-core Probes (TIP) and neutron source units
Key parameters for conditioning Dry, metallic, non-combustible
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) Co-60, Ni-63; 5.6E+08
Annual/periodic operational arising LPRM – approx. 1.8 ton (LPRM: 0.5 ton, Basket: 1.3 ton)
SRNM – approx. 0.9 ton/10years (SRNM: 0.3 ton, Basket: 0.6
ton)
TIP – approx. 0.3 ton/10-20 years (TIP: 0.05 ton, Basket: 0.25
ton)
Neutron source unit – 0.3 ton/life (Unit: 0.05 ton, Basket: 0.25
ton)
Total operational arising Approx. 116 ton/60 years
Hazardous substances None
2.4. Decommissioning Non-Radioactive Waste
The following waste details have been included for completeness. However, as they are not radioactive
wastes they will not be addressed in detail.
Table C2.4-1 Details of Non-Radioactive Decommissioning Wastes
Parameter Description
Waste Origin Metal and concrete
Waste physical/chemical description Metal and concrete
Key parameters for conditioning Large items needing size reduction for disposal
Total arising 632,900 ton
Hazardous substances TBA
2.5. Decommissioning VLLW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
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Table C2.5-1 Details of VLLW Decommissioning Wastes
Parameter Description
Waste Origin Contaminated and irradiated metal and concrete including
arisings from decontamination of some of the LLW. Arising
from various sources during decommissioning
Waste physical/chemical description Metal and concrete
Key parameters for conditioning Large items needing size reduction for disposal to licensed
disposal site
Nature of radioactive material Largely contaminated with some activation of metallic items
Main radionuclides/activity TBA
Total arising 25,983 ton
Hazardous substances TBA
2.6. Decommissioning LLW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
Table C2.6-1 Details of LLW Decommissioning Metal and Concrete Wastes
Parameter Description
Waste Origin Contaminated and irradiated metal and concrete. Arising from
various sources during decommissioning
Waste physical/chemical description Metallic and concrete
Key parameters for conditioning Large items needing size reduction for disposal to Low Level
Waste Repository
Nature of radioactive material TBA
Main radionuclides/activity TBA
Total arising 8,660 ton
Hazardous substances TBA
Table C2.6-2 Details of LLW Decommissioning Process Wastes
Parameter Description
LCW Demineraliser Bead Resin 80 m
3
arising over 20 years from operations ceasing
HCW Demineraliser Bead Resin 18 m
3
arising over 20 years from operations ceasing
HCW Evaporator Concentrated
Waste
48 m
3
arising over 20 years from operations ceasing
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2.7. Decommissioning ILW
The following information is extracted from (25) and is approximate at this stage. However, it provides
indicative quantities of wastes envisaged to arise in this category.
Table C2.7-1 Details of ILW Decommissioning Irradiated Metals
Parameter Description
Waste Origin Reactor core components and from areas subject to activation.
Waste physical/chemical description Activated metal components
Key parameters for conditioning Dry, metallic, non-combustible. Some large items needing size
reduction for packaging and disposal
Nature of radioactive material Activation products
Main radionuclides/activity (Bq/g) 1.30E+04~2.00E+10
Total arising 920 ton
Hazardous substances None
Table C2.7-2 Details of ILW Decommissioning Process Wastes
Parameter Description
CUW Filter Demineraliser Powder
resin/crud
17 m
3
arising over 5 years from operations ceasing
FPC Filter Demineraliser Powder
resin/crud
7.5 m
3
arising over 5 years from operations ceasing
LCW Filter Crud 12 m
3
arising over 20 years from operations ceasing
2.8. Spent Fuel
The following table summarises the general information currently available on the SF. (28) indicates the
detailed parameters which are required for GDA, and which will be needed in due course. This information
is currently on hold, pending the availability of an export licence. The table below also summarises specific
information which RWMD have requested to support the disposability assessment.
Table C2.8-1 Details of Spent Fuel Assemblies
Parameter Description
Waste Origin Uranium dioxide fuel pellets within Zircaloy cladding form
the fuel assemblies, which undergo fission and produce heat.
A proportion of this heat is ultimately converted to electricity.
The fission products and actinides produced during the fission
process are considered as waste. In addition to the fission
products and actinides, the structures of the fuel assemblies
are activated and at discharge from the reactor contain
activation products
Waste physical/chemical description The reactor core consists of fuel rods held in bundles by
spacer grids and top and bottom fittings. The fuel rods consist
of uranium dioxide pellets stacked in a cladding tube, plugged
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Parameter Description
and seal welded. Fuel element design is GE14. Materials of
construction are Zircaloy, stainless steel, Inconel, ceramic
UO
2
and Gd
2
O
3
Specific information required by
RWMD
1. How the SF is believed to evolve during interim storage
period. Initially this may be based upon current Japanese
practice, but may need to be modified in the future if UK
interim storage conditions are different;
2. SF data comprising a full radionuclide inventory of the
fuel and any non-fuel core components. To include initial
enrichment, burn up (and variability), any clad
impurities, burnable poisons. 1 year cooled data would
be helpful and provision of heat and A2 information if
available;
3. Details of Uranium type and specifically if any
reprocessed material was used as this can affect the
U236 profile;
4. Variability’s in burn up, for example a max core burn up
value may include variability’s for individual fuel
elements as a result of fuel shuffling;
5. Operational strategy for the core to provide
understanding of the burn up variations;
6. Information on the strategy to be adopted for failed fuel.
Key parameters for conditioning Dry, metallic, non-combustible, heat generating. Approx. 4.5
m long
Nature of radioactive material Fission products, activation products and actinides
Annual discharge from core 150 assemblies (approximate)
Total discharge from core 9600 assemblies (approximate)
Hazardous substances None
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Appendix D - Descriptions of Conditioning Options
1. Introduction
This Appendix provides detailed concepts for a number of options for management of the radioactive
wastes (VLLW, LLW and ILW) and the SF. The Environment Agency has requested that this should be
provided as a part of the GDA submission; hence the following is proposed to meet this requirement.
It should be noted that at this stage all available options are not listed and the options shown below are
current, commonly used techniques which are considered to be BAT and are included as viable options for
GDA. They are currently in use or planned to be used on specific legacy sites (e.g. Magnox Limited) and at
some operating sites (e.g. Sizewell B) and hence considered to be viable for inclusion.
For VLLW and LLW (as shown in Figure D1) the use of the LLWR is assumed as an example as there are
other facilities available, which have not been considered at this stage. For the ILW streams disposal to a
UK national GDF is assumed as, currently, this is the only proposed disposal route.
At the appropriate time, for specific waste streams all available options will be considered as part of a BAT
analysis. At this time there may also be other alternative technologies which are not available today and
have been developed to sufficient maturity for application.
The following diagrams and notes describe the processes for dealing with each of the waste stream and SF
categories which are discussed in the main document and where details are available in Appendix C.
The options comprise:
1 Very and Low Level Wastes
2 Intermediate Level Waste; Solidification of Sludge/Crud and IX Resins
3 Intermediate Level Waste; Encapsulation of Activated Metals
4 Intermediate Level Waste; Drying of Sludge/Crud and IX Resins
5 Intermediate Level Waste; Drying of Activated Metals
6 Spent Fuel management
Each of the above options is described for each of the process steps. As some of the steps are common to a
number of options the descriptions have been repeated for completeness so that the description of each
option will be complete.
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Waste
Accumulation
Slag/
secondary
waste
Sort
Decontamination/
Size reduce
Retrieve
Dry,
Compactable
(non-
incinerable)
Metallic
Combustible
Characterisation
and application of
the Waste
Hierarchy
VLLW
Dry, non-compactable &
non- incinerable, e.g. soils
& concrete. Solidified IX
resins, sludges and
concentrates in cement
matrix
Metal
Recycling
Consign to
VLLW Disposal
Facility
Combustible
Waste
Treatment
Super
compaction
VLLW
Disposal
Consign to
Recycling
Facility
Load into
suitable
transport
container
LLW Disposal
Load into
suitable
transport
container
Ash residues
Load into
suitable
transport
container
LLW Disposal
LLW Disposal
LLW Disposal
Load into
suitable
transport
container
Load into
suitable
transport
container
C
o
n
s
i
g
n
v
i
a
L
L
W
R
S
e
r
v
i
c
e
s
C
o
n
t
r
a
c
t
B
u
f
f
e
r
S
t
o
r
e
o
n
s
i
t
e
i
f
r
e
q
u
i
r
e
d
4
3
2
1
10
11
5
6
7
8
9
11
11
13
14
15
12
11
11
Figure D1 – Very and Low Level Wastes
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Notes for Figure D1 – Very and Low Level Wastes
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles.
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste item is dealt with in
an optimal manner according to its category and to enable the provision of adequate data for the
applicable waste route; for example to enable adequate justification for compliance with the applicable
Waste Acceptance Criteria (e.g.) of the LLWR.
3 The waste items are retrieved from their accumulation facilities using appropriate standard UK
practices. Solid waste items would be retrieved using tools and grabs in facilities which would mostly
be manually operated and able to ensure adequate protection to an operator from direct radiation dose
or contamination. Retrieval would be into a sorting facility or may be directly into a suitable container,
depending upon the waste item, for subsequent treatment.
4 Where waste items are mixed, a facility for sorting the waste will enable the streams to be segregated
and directed to the most appropriate process for subsequent treatment in accordance with the Waste
Hierarchy.
5 Dry, Compactable (non-incinerable) wastes will include heterogeneous items such items as those
which arise during maintenance operations within the nuclear island. The waste mainly consists of
contaminated RCA plant. Comprising metal plate, pipes, bulk, cables, lagging material, gas filters,
concrete and glass. HVAC filters (medium efficiency and HEPA type) may also be included in this
category.
6 Metallic wastes may arise from the above stream and would be segregated during the sorting process
7 Combustible wastes will comprise dry active waste generated through routine and maintenance
operations. The waste consists of CF and LCW spent hollow fibre filter membrane, polythene (sheet
and bag), paper, wood, cloth (wipes and gloves), rubber gloves, spent bead resin, spent active carbon.
8 VLLW items will arise from operations and maintenance and include such items as low activity piping,
motors and heat insulators from maintenance activities
9 Dry, non-compactable & non-incinerable wastes will include such items as soils & concrete and other
similar items which will be segregated from other wastes, as described above. This category will also
include the wet LLW streams which arise, such as ion exchange resins, sludges and concentrates
which will be rendered as dry and non-compactible, typically by mixing with a suitable solidification
medium (e.g. a cement powder formulation) in a 200 litre disposal drum. There are a number of
existing wastes from UK legacy sites which have previously been solidified in this manner and hence
a suitable formulation will be based upon these.
10 Where applicable to specific waste streams, decontamination and size reduce will be undertaken. This
step would typically be applicable where metals are prepared for subsequent melting and recycling or
where other items (e.g. concrete) may be decontaminated to allow for the bulk of the waste to be
recycled as aggregate or to be disposed of to a permitted land fill site rather than to the LLWR. Size
reduction may also be necessary solely to ensure that the waste items may be suitable for packaging in
a standard container such as a 200 litre drum.
11 When the waste is in a suitable form (e.g. via solidification of wet wastes, size reduction etc.) they will
be loaded into a suitable transport container. This may comprise the standard half height ISO container,
which is specified by and used at the LLWR. In this case, after transport to the LLWR, the container
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will be infiltrated with a cement grout and allowed to cure prior to emplacement within the LLWR
vault. LLWR also specify a number of other standard containers which may be used for transport of
wastes for subsequent treatment at the LLWR or at one of its suppliers where other processes are
applicable for treatment of the waste; for example super compaction or incineration.
12 The wastes, as accumulated into suitable transport containers, may be buffer stored on site for a period
of time. This may be required to comply with the Joint Waste Management Plan, agreed between the
site operator and LLWR which would include a schedule of waste deliveries to LLWR for onward
processing or direct disposal, depending upon the waste type.
13 It is assumed for the GDA submission that all VLLW and LLW will be consigned via a LLWR
Services Contract and will utilise the well-established facilities available for dealing with the wide
range of wastes specified. This will ensure that the Waste Hierarchy is adequately applied in directing
wastes for treatment via the most appropriate route to ensure that the quantities of waste consigned for
disposal in the LLWR vaults is minimised. Hence, the inputs from the buffer storage area will be dealt
with in a manner appropriate to the waste, which could be any one of the 5 outputs shown.
14 Slag secondary waste residues which arise from metal recycling will be treated according to their
characteristics. Their radionuclide content will be carefully determined to ensure that ‘concentration’
effect from the melting process has not rendered this waste as ILW. In the event that it has the residues
will be treated as ILW and its processing with follow the appropriate route. Typically, these LLW
residues may be packaged in a similar manner to wet wastes as identified above for ion exchange
resins, sludges and concentrates. In this case the residues may either be mixed with a prepared cement
grout in a mixing drum or by mixing the residues with water and adding a suitable cement blend
within a mixing drum, typically 200 litres. A process and formulation for solidification of ash and
dusts has previously been developed for UK legacy sites and hence these secondary wastes will be
dealt with using a similar process.
15 The ash residues arising from the incineration process can also be dealt with in a similar manner to the
above noted for the slag secondary residues. The comment for the slag residues, where the
‘concentration’ effect of incineration may have rendered the waste as ILW will also apply to this waste
stream.
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Figure D2 – Intermediate Level Waste; Solidification of Sludge/Crud and IX Resins
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Transfer to
buffer storage
as required
Pre treatment
Hydraulic
transfer
Solidify in
cement blend
formulation in
lost paddle
container
Cure, cap add
lid, inspect and
complete
package
Transfer to
shielded store
Monitored
storage period
Transfer to
GDF
1
2
3 4 5
6
7
8
9 10
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Notes for Figure D2 – Intermediate Level Waste; Solidification of Sludge/
Crud and IX Resins
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (25) as Sludge (Crud) and Powder Resin (which also contains particulate corrosion
product).
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specifications and guidance
as published by RWMD (13) to adequately inform the LoC process such that a LoC can be issued in
due course, subject to RWMD’s disposability assessment. The characterisation data may be obtained at
an appropriate point in the process where the waste stream is homogenised by a mixing/recirculation
technique to ensure that samples which are representative of the bulk of the stream are obtained.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. This technique will utilise equipment which is installed within the accumulation facility
and may comprise mixers to ensure that the wastes are homogeneous and suction devices which are
capable of extraction of the waste stream slurry at the solids concentration which would be determined
by the characterisation information.
4 Buffer storage of a batch of waste may be necessary to ensure that the subsequent process is
adequately controlled, or this facility may be required to allow for a waste solidification campaign to
be undertaken when a sufficient quantity of waste has been accumulated. This will then allow further
quantities of waste to be added to the accumulation facility to ensure that normal operations are not
held up. This buffer storage facility may also be the point at which the sampling is undertaken to
obtain the characterisation data noted in step 2 above.
5 Transfer of the waste stream from the buffer tank to the waste packaging process will use an
appropriate UK standard technique for delivering a measured quantity of waste of defined properties
to the waste container. The control of the transfer equipment will be integrated with the downstream
equipment to ensure that waste delivery complies with the overall process to produce a predictable
waste package (defined as the combination of the disposal container and the wasteform produced
within it) in accordance with the requirements of the LoC. In this case the overall integrity of the
waste package is mostly provided by the properties of the cemented contents with a minor
contribution claimed for the container.
6 A pretreatment step may be necessary to adjust and/or optimise the properties of the waste stream to
adjust such properties as (for example) the solids concentration or the pH. This is to ensure
compatibility with the packaging requirements and solidification formulation which has been
demonstrated to RWMD via the LoC process to produce satisfactory, disposable, wasteform properties.
The pretreatment step will be conducted either in the buffer tank or disposal container depending on
what is required and the waste properties. For example, reducing the water content of the IX resin
stream can be undertaken via a submerged filter within the disposal container whereas to reduce water
content within a sludge stream may require some external equipment comprising such equipment as a
settling tank or filter system.
7 Solidification of the waste stream will be conducted within a UK standard container, which is
compliant with the appropriate RWMD specification. These containers will be manufactured from
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austenitic stainless steel (typical grade 316L or EN 1.4404 to BS EN 10088-2:2005) and incorporate
an integral mixing paddle which will remain with the wasteform once it has been generated within the
container. Typical container sizes, as used in the UK for many legacy site wastes, include 500 litre and
3 m
3
drums.
The cement blend and quantity added will be as defined for the specific waste stream. This will be
based upon formulations as used in Japan for these wastes or as for similar wastes, which have arisen
at legacy sites and have been solidified in the UK. The formulation will have been previously
endorsed by RWMD via the LoC process as producing a wasteform of adequate properties, in
compliance with the relevant specifications, as noted above. The long term evolution of the wasteform
will also have been demonstrated. Both of these aspects will have been demonstrated by suitable
supporting development trials with inactive simulants or based upon directly applicable supporting
data from (for example) Japanese or UK practice.
8 After production and curing of the wasteform within the disposal container the whole package is
completed by addition of an inactive grout cap to minimise the residual voidage within the package
and fitting of the container lid by a sealed, bolted closure. The package will then be inspected via
visual means and swabbing the external surface to ensure that any contamination is within the limits
required for onward transport and interim storage on site.
9 The packages will be transferred to an on-site ILW store using a suitable shielded route. Depending
upon the site layout this may be via a shielded transporter or a shielded route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and will include addressing of
the guidance published for ILW stores for UK legacy sites (22).
This document provides guidance for an integrated approach to the overall storage system by
considering package and store performance as a whole; identification of good practice, principles,
approaches and the use of ‘toolkits’ for specific analyses; e.g. package evolution and performance
models; requirements for package care and management, including monitoring, identifying ideal,
tolerable and failing performance parameters. Where monitoring indicates deterioration, such that
some packages may not be suitable for eventual disposal, suitable ‘reworking’ schemes will be
developed which may, in exceptional circumstances require re packaging of specific waste packages.
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Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Figure D3 – Intermediate Level Waste; Encapsulation of Activated Metals
Size reduce
Add grout to
infiltrate waste
Load into
disposal
container
Cure, cap add
lid, inspect and
complete
package
Transfer to
shielded store
Monitored
storage period
Transfer to
GDF
8
7
6
5 4
3
2
1 9
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Notes for Figure D3 – Intermediate Level Waste; Encapsulation of
Activated Metals
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (24) as Higher Activity Metals; Control Rods, Channel Boxes
7
& Others. The method of
accumulation in Japan is within a ‘wet bunker’ where the items will also undergo size reduction prior
to placing in containers and grouting. The means of accumulation for UK ABWR is to be determined
and may follow UK practice for gas reactor legacy sites where such items are accumulated in dry
shielded vault stores.
2 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. The method of retrieval will be determined by the requirements of the overall process and
will be significantly influenced by how the items are to be accumulated (i.e. in a wet or a dry facility)
and where it is appropriate to conduct subsequent operations, especially size reduction.
3 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specifications and guidance
as published by RWMD (13) to adequately inform the LoC process such that an LoC can be issued in
due course, subject to RWMD’s disposability assessment. The characterisation data may be obtained at
an appropriate point in the process. For example, a gamma assay station may be incorporated within
the facility where the items are loaded into a disposal container.
4 The waste items will be size reduced in a suitable facility which will ensure that the operations can be
conducted according to ALARP principles. The design of the facility will depend upon the
accumulation method chosen. Where accumulation is in a wet bunker the necessary shielding will be
provided by the depth of water. In this case underwater techniques will be used to size reduce. Where
the items have been accumulated in a dry storage environment the necessary shielding will be
provided by a suitable shielded cell or the size reduction may be undertaken within the dry storage
facility. The size reduction techniques will comprise cutting (in a manner to limit secondary waste; e.g.
shearing) and compaction. The latter technique will take into account the need for adequate grout
infiltration during the subsequent packaging operation.
5 The items will be loaded into a suitable UK standard container, which is compliant with the
appropriate RWMD specification. These containers will be manufactured from austenitic stainless
steel (typical grade 316L or EN 1.4404 to BS EN 10088-2:2005). Typical container sizes, as used in
the UK for many legacy site wastes, include 500 litre drums and 3m
3
boxes.
6 Encapsulation of the waste stream will be conducted within the container by addition of a suitable
cement grout to produce a predictable waste package (defined as the combination of the disposal
container and the wasteform produced within it) in accordance with the requirements of the LoC. In
this case the overall integrity of the waste package is mostly provided by the properties of the
cemented contents with a minor contribution claimed for the container.
The cement blend and quantity added will be as defined for the specific waste stream. This will be
based upon formulations as used in Japan for these wastes or as for similar wastes, which have arisen
7
Note it is assumed for GDA that these items remaining with the spent fuel assemblies and are managed/packaged
as part of this stream
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at legacy sites and have been encapsulated in the UK. The formulation will have been previously
endorsed by RWMD via the LoC process as producing a wasteform of adequate properties, in
compliance with the relevant specifications, as noted above. The long term evolution of the wasteform
will also have been demonstrated. Both of these aspects will have been demonstrated by suitable
supporting development trials with inactive simulants or based upon directly applicable supporting
data from (for example) Japanese or UK practice.
7 After production and curing of the wasteform within the disposal container the whole package is
completed by addition of an inactive grout cap to minimise the residual voidage within the package
and fitting of the container lid by a sealed, bolted closure. The package will then be inspected via
visual means and swabbing the external surface to ensure that any contamination is within the limits
required for onward transport and interim storage on site.
8 The packages will be transferred to an on-site ILW store using a suitable shielded route. Depending
upon the site layout this may be via a shielded transporter or a shielded route/tunnel which is part of
the site infrastructure.
9 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and will include addressing of
the guidance published for ILW stores for UK legacy sites (22).
This document provides guidance for an integrated approach to the overall storage system by
considering package and store performance as a whole; identification of good practice, principles,
approaches and the use of ‘toolkits’ for specific analyses; e.g. package evolution and performance
models; requirements for package care and management, including monitoring, identifying ideal,
tolerable and failing performance parameters. Where monitoring indicates deterioration, such that
some packages may not be suitable for eventual disposal, suitable ‘reworking’ schemes will be
developed which may, in exceptional circumstances require re packaging of specific waste packages.
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Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud and IX Resins
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Transfer to
buffer storage
as required
Hydraulic
transfer into
robust
shielded
container
Fit bolted lid,
test seals,
swab &
monitor
Transfer to
drying plant
Retest
closures
where
required, swab
& monitor
Transfer to
intermediate
storage facility
Monitored
storage period
Transfer to
GDF
1
8
7
6
5
4 3
2
9
10
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Notes for Figure D4 – Intermediate Level Waste; Drying of Sludge/Crud
and IX Resins
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. The wastes are
identified in (24) as Sludge (Crud), Powder resin (which also contains particulate corrosion product).
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specification as published
by RWMD (14) to adequately inform the LoC process such that a LoC can be issued in due course,
subject to RWMD’s disposability assessment. The characterisation data may be obtained at an
appropriate point in the process where the waste stream is homogenised by a mixing/recirculation
technique to ensure that samples which are representative of the bulk of the stream are obtained.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. This technique will utilise equipment which is installed within the accumulation facility
and may comprise mixers to ensure that the wastes are homogeneous and suction devices which are
capable of extraction of the waste stream slurry at the solids concentration which would be determined
by the characterisation information.
4 Buffer storage of a batch of waste may be necessary to ensure that the subsequent process is
adequately controlled, or this facility may be required to allow for a waste packaging campaign to be
undertaken when a sufficient quantity of waste has been accumulated. This will then allow further
quantities of waste to be added to the accumulation facility to ensure that normal operations are not
held up. This buffer storage facility may also be the point at which the sampling is undertaken to
obtain the characterisation data noted in step 2 above.
5 Transfer of the waste stream from the buffer tank to the waste packaging process will use an
appropriate UK standard technique for delivering a measured quantity of waste of defined properties
to the waste container. The control of the transfer equipment will be integrated with the downstream
equipment to ensure that waste delivery complies with the overall process to produce a predictable
waste package (defined as the combination of the disposal container and the wasteform produced
within it) in accordance with the requirements of the LoC. In this case the overall integrity of the
waste package is provided by the robust container and none would be claimed for the waste contents.
The robust container would be one of a number of designs currently being considered within the UK
by some legacy sites (e.g. Magnox Limited) and at the Sizewell B PWR. They are manufactured from
ductile cast iron, a material which is readily formed into the required shape by casting/machining and
whose properties are enhanced by the metal composition and controlled cooling to increase the
material’s ductility. The currently available designs can be supplied by Gesellschaft für
Nuklear-Service mbH (GNS - http://www.gns.de/language=en/4877) and Croft
(http://www.croftltd.com/).
6 Depending upon the layout of the retrieval equipment and the rest of the process plant the container
may then be sealed at this point by addition of a sealed lid secured by bolts. If the containers location
was in a contamination controlled area the container’s external surfaces would also be swabbed and
monitored prior to transfer to later stages in the process to ensure that contamination levels are within
allowable limits.
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7 The waste within the container will be dried by removal of bulk water and as much interstitial
moisture as required in order to achieve the required dryness level. This process will require
penetrations to be made into the container for the application of a vacuum, and some external heating
will also be applied to achieve this dryness. The process plant will include appropriate monitoring
equipment to measure the residual moisture content and hence determine an end point for the drying
process.
8 After the required dryness level is achieved the process equipment will be disconnected and the
container resealed by insertion of appropriate closures, which will be tested for leak tightness. The
container’s external surfaces would again be swabbed and monitored to ensure that contamination
levels are within allowable limits.
9 The packages will be transferred to an on-site ILW interim storage facility using a suitable route.
Depending upon the site layout this may be via a suitable transporter or a route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and may include addressing of
the guidance published for ILW stores for UK legacy sites (22). However, the applicability of this
guidance will be reviewed as it was developed by the NDA for shielded stores which are intended for
the long term interim storage of unshielded ILW packages, mostly comprising cemented wasteform
within thin walled stainless steel containers.
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Figure D5 – Intermediate Level Waste; Drying of Activated Metals
Waste
Accumulation
Characterisation
and application of
the Waste
Hierarchy
Retrieve
Size reduce
and sort as
necessary
Transfer into
robust
shielded
container
Fit bolted lid,
test seals,
swab &
monitor
Transfer to
drying plant
Retest
closures
where
required, swab
& monitor
Transfer to
intermediate
storage facility
Monitored
storage period
Transfer to
GDF
1
8
7
6
5
4
3
2
9
10
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Notes for Figure D5 – Intermediate Level Waste; Drying of Activated
Metals
1 The wastes will be accumulated in suitable controlled facilities in accordance with current UK practice
to ensure adequate containment and control of the materials and that subsequent retrieval for onward
processing and disposal can be undertaken in accordance with ALARP principles. Higher Activity
Metals; Control Rods, Channel Boxes
8
& Others. The method of accumulation in Japan is within a
‘wet bunker’ where the items will also undergo size reduction prior to placing in containers and
grouting. The means of accumulation for UK ABWR is to be determined and may follow UK practice
for gas reactor legacy sites where such items are accumulated in dry shielded vault stores.
2 All wastes will be managed in accordance with the Waste Hierarchy to ensure that waste generation is
minimised. The characterisation data will be sufficient to ensure that each waste stream is dealt with in
an optimal manner according to its type and to enable the provision of adequate data for the applicable
waste route. This will enable adequate justification of compliance with the specification as published
by RWMD (14) to adequately inform the LoC process such that an LoC can be issued in due course,
subject to RWMD’s disposability assessment. The characterisation data may be obtained at an
appropriate point in the process. For example, a gamma assay station may be incorporated within the
facility where the items are loaded into a disposal container.
3 The waste streams are retrieved from their accumulation facilities using an appropriate standard UK
technique. The method of retrieval will be determined by the requirements of the overall process and
will be significantly influenced by how the items are to be accumulated (i.e. in a wet or a dry facility)
and where it is appropriate to conduct subsequent operations, especially size reduction.
4 The waste items will be size reduced in a suitable facility which will ensure that the operations can be
conducted according to ALARP principles. The design of the facility will depend upon the
accumulation method chosen. Where accumulation is in a wet bunker the necessary shielding will be
provided by the depth of water. In this case underwater techniques will be used to size reduce. Where
the items have been accumulated in a dry storage environment the necessary shielding will be
provided by a suitable shielded cell or the size reduction may be undertaken within the dry storage
facility. The size reduction techniques will comprise cutting (in a manner to limit secondary waste; e.g.
shearing) and compaction.
5 Transfer of the waste stream to the waste packaging process will use an appropriate UK standard
technique for delivering a measured quantity of waste of defined properties to the waste container.
At this stage the waste may be characterised by gamma assay or similar. The control of the transfer
equipment will be integrated with the downstream equipment to ensure that waste delivery complies
with the overall process to produce a predictable waste package (defined as the combination of the
disposal container and the wasteform produced within it) in accordance with the requirements of the
LoC. In this case the overall integrity of the waste package is provided by the robust container and
none would be claimed for the waste contents.
The robust container would be one of a number of designs currently being considered within the UK
by some legacy sites (e.g. Magnox Limited) and at the Sizewell B PWR. They are manufactured from
ductile cast iron, a material which is readily formed into the required shape by casting/machining and
whose properties are enhanced by the metal composition and controlled cooling to increase the
material’s ductility. The currently available designs can be supplied by Gesellschaft für
8
Note it is assumed for GDA that these items remaining with the spent fuel assemblies and are managed/packaged
as part of this stream
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Nuklear-Service mbH (GNS - http://www.gns.de/language=en/4877) and Croft
(http://www.croftltd.com/).
6 Depending upon the layout of the retrieval equipment and the rest of the process plant the container
may then be sealed at this point by addition of a sealed lid secured by bolts. If the containers location
was in a contamination controlled area the container’s external surfaces would also be swabbed and
monitored prior to transfer to later stages in the process to ensure that contamination levels are within
allowable limits.
7 The waste within the container will be dried by removal of any residual bulk water and as much
interstitial/adherent moisture as required in order to achieve the required dryness level. This process
will require penetrations to be made into the container for the application of a vacuum, and some
external heating will also be applied to achieve this dryness. The process plant will include appropriate
monitoring equipment to measure the residual moisture content and hence determine an end point for
the drying process.
8 After the required dryness level is achieved the process equipment will be disconnected and the
container resealed by insertion of appropriate closures, which will be tested for leak tightness. The
container’s external surfaces would again be swabbed and monitored to ensure that contamination
levels are within allowable limits.
9 The packages will be transferred to an on-site ILW interim storage facility using a suitable route.
Depending upon the site layout this may be via a suitable transporter or a route/tunnel which is part of
the site infrastructure.
10 The site will include a store which is suitable for the interim storage (for a period up to 100 years) of
the ILW packages. The store will be built to standard UK requirements and may include addressing of
the guidance published for ILW stores for UK legacy sites (22). However, the applicability of this
guidance will be reviewed as it was developed by the NDA for shielded stores which are intended for
the long term interim storage of unshielded ILW packages, mostly comprising cemented wasteform
within thin walled stainless steel containers.
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Accumulate in
Fuel Pool of
Reactor Building
1
Retrieve
Load into Dry
Cask
Load into
Multi Purpose
Container (MPC)
Load into
Sealed Channel
Transport by
Wet Cask
Load into
KBS-3 Container
On-site
Dry Cask Storage
On-site
MPC Storage
Modular Vault
Dry Storage
On-site
Fuel Pool
Storage
On-site Dry
Container
Storage
R
e
-
p
a
c
k
a
g
e
f
o
r
D
i
s
p
o
s
a
l
(
i
f
n
e
c
e
s
s
a
r
y
)
T
r
a
n
s
f
e
r
t
o
G
D
F
OPTIONS
2
3
5
7
9
11
4
6
8
10
12
13
14
Figure D6 – Spent Fuel Management Options
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Notes for Figure D6 – Spent Fuel Management Options
1 After discharge from the reactor the spent fuel assemblies (each comprising the fuel element bundle
and its surrounding channel box) will be accumulated in the spent fuel pool for a period of decay
storage to allow them to cool. This period is usually several years (believed to be approx. 5 years for
typical BWR spent fuel) and the spent fuel pool water will be conditioned to maintain the correct
temperature, cleanliness and chemistry conditions.
2 The method used for retrieval of the spent fuel/channel box will depend upon the subsequent storage
option chosen. The usual method for removal of the spent fuel assemblies is to load them vertically
into a flask which is submerged into the spent fuel pool, followed by lidding and removal of the cask.
Operations to decontaminate the flask exterior and condition the internal water are also undertaken
prior to transport of the flask to its intended location. This is usually a reprocessing facility but for the
UK ABWR spent fuel reprocessing will not be assumed and hence the retrieval method is likely to be
different. The envisaged retrieval method, as part of each of the noted options will be described in the
relevant sections below.
3 Loading the spent fuel assemblies into a dry cask will require a number of steps which will include the
following:
a. Load the spent fuel assemblies, under water, into an inner container, which is compatible with
the chosen dry cask. This container may be an MPC, as described in 5 below.
b. Drain the water from the container and seal with a lid closure. This would be completed in a
facility which provides appropriate shielding, containment for contamination control and
remotely operated equipment.
c. Dry the internal environment and backfill with an inert gas.
d. Transfer the containers into a dry cask.
4 The dry casks are stored in a suitable facility which is likely to be similar to an industrial type
warehouse which will provide weather and security protection. The casks are likely to be transported
into and out of the facility using specialist heavy equipment, probably similar to that offered by Holtec,
as shown at http://rampac.energy.gov/docs/education/Q8.pdf. It is unlikely that the store will require
any conditioning of the internal environment as the casks can be inspected externally and any
necessary maintenance undertaken.
5 Use of a Multi-Purpose Container (MPC) similar to the type used by Holtec (as above) inside the dry
cask may allow the spent fuel elements within to be disposed of directly to a GDF without further re
packaging. This would be subject to granting of a LoC, which will include a disposability assessment,
from RWMD and other decisions related to logistics and commercial/business requirements. For
example, a container suitable for GDF disposal is likely to be manufactured from relatively expensive
materials and to more stringent requirements than required for a period of on-site interim storage
inside the cask. This could entail greater costs at this stage than if packaging for disposal was delayed
until a GDF was available. The time period from the start of spent fuel interim storage to the
availability of a GDF may also allow for the development of, as yet unknown, more cost effective
techniques for producing a disposable package. Or the requirements for disposability at the GDF may
have been modified to allow for a more cost effective method of spent fuel packaging to be used.
These potential benefits would need to be weighed against the requirements for repackaging of the
spent fuel, when the GDF was available, assuming that an MPC was not initially used.
6 Storage of the MPC would be within a suitable dry cask in facilities as described in 4 above.
7 The spent fuel assemblies may be transferred into sealed channels, either singly or in groups after
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being removed from the spent fuel pool and dried. The sealed channels would be part of a Modular
Vault Dry Store (MVDS) facility, similar to that described in the following link:
http://www.wmsym.org/archives/1988/V2/80.pdf. The operations to remove, dry and transport the
assemblies to the MVDS would require specialist equipment, probably remotely operated within a
shielded containment facility and utilising a shielded transport vehicle to transfer the assemblies from
the reactor building to the, probably separate, MVDS building.
8 Interim storage within a MVDS system can continue until a GDF is available as the cooling is by
passive means, which increases its reliability. The link in 7 above describes one such system, which is
based upon the proven method for dry storage of spent Magnox fuel at the Wylfa power station.
9 Use of a wet cask for removal of the spent fuel assemblies, as described in 2 above, will facilitate the
transport of the assemblies to an on-site fuel pool storage facility, which would be separate from the
reactor’s operational spent fuel pool.
10 Fuel pool storage would continue for an extended period until a GDF was available. This would
include separate facilities for conditioning, cooling and cleaning the water to maintain the correct
quality.
11 The KBS-3 container is a developed container which may be suitable for disposal to a GDF, subject to
granting of a LoC by RWMD. Hence, the spent fuel assemblies may be loaded directly into these
containers using a procedure similar to as described in 3 above. However, as these containers are
likely to be more expensive than those used for interim storage only, their use will need to be carefully
considered taking account of factors such as discussed in 5 above. Use of these containers may also
eliminate the opportunity of utilising any improved methods or take account of any changes in the
Regulatory and disposability regimes during the interim storage period.
12 The KBS-3 containers would be stored during the interim period within a suitable dry cask in facilities
as described in 4 above.
13 Where appropriate, the spent fuel assemblies will need to be re-packaged for transport to and disposal
at a GDF. As described in 5/6 &11/12 above, where the assemblies are initially stored in an MPC or
the KBS-3 container these are likely to be acceptable for disposal, subject to granting of a LoC by
RWMD. Hence, re-packaging should not be required for these options.
For options 3/4 and 7/8, where the spent fuel assemblies have been stored for the interim period in
dry facilities, re-packaging by transfer into a suitable disposal container will require a suitable facility
which incorporates shielding and containment for contamination control and remotely operated
equipment.
For option 9/10 above transfer into a suitable disposal container will require similar facilities as above
plus draining and drying facilities to ensure that the spent fuel assemblies contain minimal (or
preferably zero) residual moisture prior to loading into a suitable disposal container.
These latter options will require the granting of appropriate LoCs by RWMD prior to re-packaging for
disposal.
14 Transfer of the packaged spent fuel assemblies to a GDF will be subject to granting of the appropriate
LoCs, as described above. Depending upon the packaging adopted for disposal a suitable transport
method will need to be adopted. For example, the Holtec system referenced above includes various
transport casks. However, within the UK the use of these or any other transport cask will need to be
assessed by the appropriate authority as well as the noted requirement for LoCs. The Office for
Nuclear Regulation (ONR) currently covers Nuclear Transport aspects within their remit under its
Radioactive Materials Transport (RMT) function.
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