Radioactive Waste Management and Decommissioning
NEA/RWM/R(2020)2
November 2021
www.oecd-nea.org

Characterisation Methodology
for Unconventional and Legacy
Waste

Nuclear Energy Agency

NEA/RWM/R(2020)2
Unclassified

English text only
22 November 2021

NUCLEAR ENERGY AGENCY

Radioactive Waste Management Committee

Characterisation Methodology for Unconventional and Legacy Waste

This document is available in PDF format only.

Contact: Rebecca Tadesse (rebecca.tadesse@oecd-nea.org), Head of Radioactive Waste
Management and Decommissioning Division

JT03485922
OFDE

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2 | NEA/RWM/R(2020)2

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CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 3

Foreword
The Expert Group on Characterisation Methodology of Unconventional and Legacy Waste(EGCUL) was established under the Nuclear Energy Agency (NEA) Radioactive Waste
Management Committee (RWMC) and held its first meeting in November 2018. The groupwas set up to develop a strategic approach to manage the complex characterisation processof waste, which consists of a sampling and detailed analysis plan based on statistical
approaches, calculation methods and evaluation process for the data obtained.
The group comprises experts who have experience in radioactive waste management
characterisation, in managing radiological contamination situations following accidents orin decommissioning legacy waste facilities.
The mandate of EGCUL ran from 1 August 2018 to 31 July 2020 and consisted of four
meetings at the NEA headquarters in Paris. The experts provided information on
characterisation techniques used as well as challenges and lessons learnt from theapproaches that have been taken. This information will assist those experts dealing withthe post clean-up activities at the Fukushima Daiichi site as well as others embarking or
continuing to work on the management of other accident and legacy waste.

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Acknowledgements
This report was prepared by the Expert Group on Characterisation Methodology of
Unconventional and Legacy Waste (EGCUL) set up under the programme of work of the
NEA Radioactive Waste Management Committee (RWMC). The EGCUL was chaired by
Dr Per Strand (Norway) supported by two Vice Chairs, Dr Kazuyuki Kato (Japan) andDr Nick Jefferies (United Kingdom). The NEA Secretariat wishes to express its sincere
gratitude to those mentioned above for co-ordinating the report, and to those providing thecase studies. Special thanks are also due to those who simulated discussion by presenting
information on their experiences.

Mr Jérôme Comte (France)
Mr Yoshikazu Koma (Japan)
Mr Sergei Floria (Russia)
Mr Viacheclav Navalikhin (Ukraine)
Mrs Jo Plyer (Consultant)
Mr Tomohiro Higashihara (NEA)
Mrs Madoka Koizumi (NEA)
Mr Ichiro Otsuka (NEA)
Mrs Rebecca Tadesse (NEA)
Mr Wei-Whua Loa (NEA)CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 5

Table of contents
List of abbreviations and acronyms ..................................................................................................... 8
Executive summary ............................................................................................................................. 11
Chapter 1. Introduction, objectives and scope.................................................................................. 16
1.1. Background ................................................................................................................................. 16
1.2. Objectives and scope of work ..................................................................................................... 18
1.3. Mode of operation of the group and structure of this report ....................................................... 18
References.......................................................................................................................................... 19
Chapter 2. Methodology features ....................................................................................................... 20
2.1. Differences between conventional and unconventional/legacy waste ........................................ 20
2.2. Toolboxes of common approaches ............................................................................................. 21
2.3. Data Quality Objectives process ................................................................................................. 23
2.4. Data Quality Assessment ............................................................................................................ 25
2.5. MARSSIM, MARSAME and MARLAP ................................................................................... 25
2.6. NEA Working Party on Decommissioning and Dismantling ..................................................... 27
2.7. Interaction with stakeholders ...................................................................................................... 35
References.......................................................................................................................................... 36
Chapter 3. Case Studies ...................................................................................................................... 38
3.1. France.......................................................................................................................................... 38
3.2. Japan ........................................................................................................................................... 51
3.3. Russia .......................................................................................................................................... 71
3.4. Ukraine........................................................................................................................................ 80
3.4.1. Characterisation of legacy waste produced at Rivne Nuclear Power Plant ............................. 80
3.4.2. Characterisation of radioactive waste and material from Chernobyl Nuclear Power Plant ..... 87
3.4.3. Characterisation of radioactive materials and fuel-containing material at the “Shelter”
Object ....................................................................................................................................... 94
3.4.4. Safety assessment of radioactive waste disposal site “Podlesnyi”......................................... 102
3.4.5. Radioactive waste temporary localisation sites ..................................................................... 107
3.5. United Kingdom ....................................................................................................................... 119
Chapter 4. Challenges and lessons learnt derived from the case studies ..................................... 131
References........................................................................................................................................ 140
Chapter 5. Characterisation methodology and adaptations and enhancements to address UL
waste challenges ................................................................................................................................. 142
References........................................................................................................................................ 153
Chapter 6. Conclusion and recommendations on good practice ................................................... 154
Annex A: Analytical methods for assessing radionuclides in the waste samples ......................... 156

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FiguresFigure 2.1. Data Quality Objectives methodology ................................................................................ 24
Figure 2.2. Data Quality Assessment (DQA) process ........................................................................... 25
Figure 2.3. Typical radiological characterisation objectives during the life cycle of a nuclear
facility.................................................................................................................................. 29
Figure 2.4. Characterisation approaches................................................................................................ 32
Figure 2.5. Data assessment and evaluation process ............................................................................. 33
Figure 2.6. Stakeholders and how they might influence the strategy for decommissioning ................. 36
Figure 3.1. Former effluent treatment facility configuration according to the processes ...................... 40
Figure 3.2. Theoretical and practical approach to waste characterisation ............................................. 41
Figure 3.3. Identification of the different types of tanks ....................................................................... 44
Figure 3.4. Sample plan ......................................................................................................................... 46
Figure 3.5. Comparison of alpha or beta gamma emitters – liquid/ solid ratio ..................................... 47
Figure 3.6. Comparing the measured 137Cs activity to a normal law..................................................... 47
Figure 3.7. 241Am/137Cs ratio in all samples .......................................................................................... 48
Figure 3.8. Radiological identification of the process by the 241Am/137Cs ratio .................................... 48
Figure 3.9. 241Am/137Cs ratio and associated clusters ............................................................................ 49
Figure: 3.10. Schematic representation of the Storage Management Plan ............................................ 56
Figure 3.11. Present and future storage conditions of rubble and secondary waste generated bywater treatment on-site at the Fukushima Daiichi Nuclear Power Plant .......................... 57
Figure 3.12. Concentration of radionuclides as ratios to that of 137Cs, for samples collectedin PCV............................................................................................................................... 59
Figure 3.13. Concentrations of 90Sr and 238Pu in the accumulated water collected from several
points................................................................................................................................. 60
Figure 3.14. Balance of nuclides between the solid substance and the filtrate from the accumulated
water.................................................................................................................................. 61
Figure 3.15. Behaviour of Np, which is presented as transport ratio of 237Np, compared with otheractinides ............................................................................................................................ 61
Figure 3.16. Flowsheet for the dissolution of sludge ............................................................................ 62
Figure 3.17. Estimation of the number of samples analysed for each waste streams over 10 years
at Laboratory-1 ................................................................................................................. 66
Figure 3.18. Grouping chemical elements by using transport ratio and an example of frequency
distribution ........................................................................................................................ 67
Figure 3.19. Decrease of uncertainty in parameters of log-normal distribution with increased
analytical data ................................................................................................................... 68
Figure 3.20. Example of bulk storage ................................................................................................... 71
Figure 3.21. General view of legacy bulk storage ................................................................................. 72
Figure 3.22. Work area .......................................................................................................................... 73
Figure 3.23. Preliminary characteristics of the FSUE FEO (formerly RosRAO) solid radioactive
waste final isolation sites .................................................................................................. 74
Figure 3.24. Bulk solid radioactive waste characterisation flowchart ................................................... 77
Figure 3.25. The Specially Equipped Near-Surface Solid Radioactive Waste Storage Facility ........... 81
Figure 3.26. Screw drill and SBC samples ............................................................................................ 83
Figure 3.27. Linear regression of the relationship between activity values of 90Sr and 137Cs ............... 85
Figure 3.28. Linear regression logarithmic relationship between activity values of 134Cs and 137Cs .... 85
Figure 3.29. Solid Radioactive Waste Treatment Plant and Solid Radioactive Waste Storage
Facility with the system for retrieval of radioactive waste (RAW) .................................. 88
Figure 3.30. Scheme of certification of difficult-to-measure radionuclides .......................................... 90
Figure 3.31. Schematic of the solid radioactive waste retrieval facility................................................ 93

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Figure 3.32. Various views of the destroyed unit 4 ............................................................................... 94
Figure 3.33. The Shelter Object during construction in November 1986 ............................................. 95
Figure 3.34. Schematic cross section of the destroyed unit 4................................................................ 96
Figure 3.35. The four different groups of LFCM .................................................................................. 98
Figure 3.36. “New Safety Confinement – Shelter Object” complex ..................................................... 99
Figure 3.37. Automated systems used to take (L)FCM samples ......................................................... 101
Figure 3.38. General view and layout of the RDWS Podlesnyi Module A ......................................... 102
Figure 3.39. Sampling and preparations for measurements ................................................................ 104
Figure 3.40. Diagram of radionuclide concentrations in the samples from Module A-1 .................... 105
Figure 3.41. Basemap RWTLS in Exclusion Zone ............................................................................. 108
Figure 3.42. Radioactive waste temporary localisation clamp/trench and Chernobyl Nuclear
Power Plant cooling pond ............................................................................................... 110
Figure 3.43. Exposure dose rate gamma survey (logging) for two boreholes at clamp 33 ................. 113
Figure 3.44. Example of a systematic process for characterising solid UL waste .............................. 124
Figure 5.1. A characterisation decision tree for accident waste .......................................................... 151
Figure 5.2. Characterisation roadmap ................................................................................................. 152
TablesTable 2.1. Key characterisation objectives carried out in each phase ................................................... 28
Table 3.1. Waste acceptance criteria ..................................................................................................... 42
Table 3.2. Management status of rubble ............................................................................................... 53
Table 3.3. Management status of felled trees ........................................................................................ 53
Table 3.4. Management status of used PPE .......................................................................................... 54
Table 3.5. Management status of used vessels generated from the water treatment plant .................... 54
Table 3.6. Management of waste sludge ............................................................................................... 54
Table 3.7. Management of concentrated waste liquid ........................................................................... 54
Table 3.8. Characteristics of the 6 target waste groups ......................................................................... 64
Table 3.9. Composition of the solid radioactive waste retrieved from the canyon ............................... 74
Table 3.10. The number of hours worked by each position in the team................................................ 75
Table 3.11. Equipment required for the solid radioactive waste retrieval ............................................. 75
Table 3.12. Expendable material used during and solid radioactive waste retrieval ............................. 76
Table 3.13. Classification of solid radioactive waste ............................................................................ 78
Table 3.14. Radionuclide vector for saline bitumen compound samples (as of 6 December 2018) ..... 86
Table 3.15. Waste acceptance criteria and measurement method for radioactive waste to bedisposed of at the specially equipped near-surface solid radioactive waste storage
facility ................................................................................................................................ 89
Table 3.16. Difficult-to-measure radionuclides and the key radionuclides that they are correlatedwith to derive a fingerprint ................................................................................................ 91
Table 3.17. (Larva-like) Fuel Contaminated Material distribution inside Shelter Object ..................... 97
Table 3.18. The relative mass content of separate elements in LFCM.................................................. 98
Table 3.19. Radionuclide analysis of the samples taken from Module A ........................................... 105
Table 3.20. Radioactive waste disposal site fingerprint ...................................................................... 106
Table 3.21. Characteristics of the main radioactive waste temporary localisation sites situatedin the Exclusion Zone ...................................................................................................... 109
Table A1. Summary of methods and instruments used to analyse samples taken from CEA
Cadarache......................................................................................................................... 157
Table A2. Method and analysis of samples taken from Rovno salt-bitumen compound .................... 159

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List of abbreviations and acronyms
ALARP

As low as reasonably practicable

BATBest Available Techniques

CDLM

Committee of Decommissioning of Nuclear Installations and Legacy
Management (NEA)CEA DEN

Nuclear Energy Division of Atomic Energy Commission (France)

CEMRW

Central Enterprise for the Management of Radioactive Waste

CFA

Conditions for Acceptance

CLADS

Collaborative Laboratories for Advanced Decommissioning Science (Japan)

CMS

Container monitoring system

CP

Characterisation Plan

CRPPHCommittee on Radiological Protection and Public Health (NEA)CR

Characterisation ReportCS

Control station

D&D

Decommissioning and dismantling

DAP

Data Acquisition Plan

DMS

Drum monitoring system

DQA

Data Quality Assessment

DQO

Data Quality Objectives

DTM

Difficult-to-measure (radionuclides)

ESS

Ecological safe system

EDR

Equivalent dose rate

EGCUL

Expert Group on Characterisation of Unconventional and Legacy Waste (NEA)

EGFWMD

Expert Group on Fukushima Waste Management and Decommissioning R&D
(NEA)

EGLM

Expert Group on Legacy Management (EGLM)

EPA

Environmental Protection Agency (United States)

FCM

Fuel-containing material

GCS

Gamma camera system

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 9

GDF

Geological disposal facility

GIS

Geographical Information Systems

GPG

Good Practice Guide (United Kingdom)

HAW

Higher Activity Waste

HLW

High-level waste

IAEA

International Atomic Energy Agency

ICP-AES

Inductively coupled plasma atomic emission spectroscopy

ICP-MS

Inductively coupled plasma mass spectrometry

ICRP

International Commission on Radiological Protection

ILW

Intermediate-level waste

ICSRM

Industrial Complex for Solid Radioactive Waste Management (Ukraine)

IRID

International Research Institute for Nuclear Decommissioning (Japan)

JAEA

Japan Atomic Energy Agency

LFCM

Lava-like fuel contaminated material

LfE

Learning from experience

LILW

Long-lived waste

LILW-LL

Low and intermediate-level long-lived waste

LLW

Low-level waste

LLWR

Low Level Waste Repository (United Kingdom)

LRW

Liquid radioactive waste

M&E

Material and equipment

MARLAP

Multi-Agency Radiological Laboratory Analytical Protocols (US EPA)MARSSIM

Multi-Agency Radiation Survey and Site Investigation Manual (US EPA)MARSAME

Multi-Agency Radiation Survey and Assessment of Materials and Equipment (US
EPA)

MDA

Minimum detectable activity

NASU

National Academy of Sciences of Ukraine

NDA

Nuclear Decommissioning Authority (United Kingdom)

NLS

Nuclear licensed sites

NMS

Neutron radiation monitoring system

NORM

Naturally occurring radioactive material

NRA

Japan Nuclear Regulation Authority

NSC

New Safe Confinement

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NSC-SONew Safety Confinement – Shelter Object (Ukraine)

ONR

Office for Nuclear Regulation (United Kingdom)

PIPS

Passivated implanted planar silicon detector (France)

PCV

Primary containment vessel

PPE

Personal protective equipment

QA

Quality Assurance

QAP

Quality Assurance Programme

QP

Quality Plan

R&D

Research and development

RMARF

Radioactive Material Analysis and Research Facility (Fukushima Daiichi Nuclear
Power Plant, Japan)

RWDS

Radioactive waste disposal site

RWMC

Radioactive Waste Management Committee (NEA)

RWTLSRadioactive Waste Temporary Localisation Sites (Ukraine)SO

Shelter Object (Ukraine)

SBC

Saline-bitumen compound (Russia)

SENSRWF

Specially Equipped Near-Surface Solid Radioactive Waste Storage Facility(Ukraine)

SFAIRP

So far as is reasonably practicable

SNRIU

State Nuclear Regulatory Inspectorate (Ukraine)

SSE CEMRWSpecialised State Enterprise Central Utility to Management Radioactive Waste(Ukraine)

SRW

Solid radioactive waste

SRWSF

Solid Radioactive Waste Storage Facility (Ukraine)

SRWTP

Solid Radioactive Waste Treatment Plant (Ukraine)

TGRCD

Task Group on Radiological Characterisation and Decommissioning (NEA)TEPCOTokyo Electric Power Company Holdings Inc. (Japan)

TIP

Traversing Incore Probe

UL

Unconventional and legacy waste

US EPA

United States Environmental Protection Agency

VNIPIET

All-Russian Scientific Research and Design Institute for Energy Technology

VSP

Visual Sampling Plan

WAC

Waste acceptance criteria

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Executive summary
Experience has shown that the development of a reliable and efficient characterisation andcategorisation methodology is a common challenge in the fields of post-accident
radioactive waste and legacy waste management. Therefore, a strategic approach should be
developed to manage the complex characterisation process of the waste, which iscomprised of a sampling and detailed analysis plan based on statistical approaches,
calculation methods and an evaluation process for the data obtained. State-of-the-art
knowledge and experience in such characterisation will be valuable for developing reliable
and efficient characterisation and categorisation methodology.
Consideration has been given to widely used waste stream characterisation procedures andmethodologies that have been developed primarily for application in conventional
decommissioning work. It has been recognised that old nuclear facilities may show
characteristics that might commonly be associated with legacies even though they may
have always been regulated to standards applicable at the time of their design and operation.
By explicitly recognising the challenges and lessons learnt, it has been possible to identify
opportunities for adaptations and enhancements to characterisation methods that could be
used to manage unconventional and legacy waste. The need for integration of waste
characterisation with other aspects of strategic planning for unconventional and legacy
waste management is highlighted. Conclusions are then drawn alongside further
recommendations for good practice and preparation of international guidance on theapplication of adaptations and enhancements. This information should support Japan’s
development of a new characterisation method.
The Expert Group on Characterisation Methodology of Unconventional and Legacy Waste(EGCUL) output is intended to support the management and characterisation of wastes
arising at post-accident and legacy sites. The focus of this report is on characterisation interms of safe management of radioactive waste by considering both the radiological andnon-radiological hazards. Several case studies have been prepared from France, Japan,
Russia, Ukraine and the United Kingdom supported by a range of material provided by
NEA member countries during the meeting held in December 2019 by the Committee on
Decommissioning of Nuclear Installations and Legacy Management.

Chapter 1. Introduction
Background
The NEA Radioactive Waste Management Committee (RWMC) established the Expert
Group on Characterisation Methodology of Unconventional and Legacy Waste (EGCUL)in 2018. Members of this group include experts who have gained experience incharacterisation of radioactive waste or in the characterisation of decommissioning
facilities. They provide technical options and ideas how to characterise large volume

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12 | NEA/RWM/R(2020)2
unconventional waste at the Fukushima Daiichi site. The EGCUL was established with theprimary aim of offering advice on the characterisation methodology for unconventional andlegacy waste at Fukushima Daiichi and of sharing experience with the international
community and other NEA members.
The EGCUL held meetings and site visits to Fukushima Daiichi Nuclear Power Plant site.
Members of the group visited:•

Fukushima Daiichi Nuclear Power Plant;•

Okuma Analysis and Research Center;•

Collaborative Laboratories for Advanced Decommissioning Science (CLADS)
research facility;•

Naraha Center for Remote Control Technology Development (NARREC).

The members of the EGCUL agreed the term “unknown” waste would refer to both
“unconventional” and “legacy” (UL) waste. The term “legacy” has many different genericand legal meanings both nationally and internationally with limited consistency in its
application (NEA, 2019). However, for the purpose of this document, the EGCUL decided
that the following descriptions would be used to define these terms:•

Legacy waste is radioactive waste that was generated by the historical activities at
a time when waste storage and treatment was not well planned. There typicallyexists little or limited information on the provenance1 or nature of the waste andthere may be considerable challenges to the retrieval of this waste.

•

Unconventional waste is radioactive waste that was generated in unexpected
situation, such as an accident, and has unknown properties.

•

Non-nuclear (e.g. from hospitals, industrial activities) and naturally occurring
radioactive material (NORM) was not included in the scope of the EGCUL report.

Objectives and aim
The objectives of the Expert Group on Characterisation Methodology of Unconventionaland Legacy Waste (EGCUL) are to:•Share state-of-the-art knowledge and experience in characterisation of a large
amount of unknown waste.

•

Discuss relevant issues and challenges on characterisation in member countries.

•

Provide international feedback on Japan’s newly developed characterisation
methodology. As the characterisation methodology is still being developed this will
not be addressed in this report. However, the international experience presented inthis report should support the ongoing development of Japan’s characterisation
methodology to be applied in the context of Fukushima Daiichi accident waste.

•

Explore possible links with other NEA groups concerned with legacy management
and radiological protection issues.

1

Provenance concerns understanding of where the material has come from and the processes that havegenerated it, as discussed further in Section 2.6.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 13•

Draft a report to be submitted to the RWMC on state-of-the-art methodology andbest practice on the characterisation of a large amount of “unknown” waste.

Scope
The present report covers important aspects relating to radiological characterisation andsurvey methodology of unconventional and legacy waste at nuclear installations withrespect to waste management and decommissioning. Brief outlines of the scope of Chapters
2-6 are presented below.

Chapter 2. Methodology features
Radiological characterisation plays an important role in the management of waste produced
from the decommissioning of nuclear power plants and associated facilities. Several
techniques have been developed for the characterisation of conventional solid radioactive
wastes. This chapter explains the United States Environmental Protection Agency (EPA)
Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) and the workof the NEA Working Party on Decommissioning and Dismantling (WPDD). The
MARSSIM provides information on planning, conducting, evaluating, and documenting
building surface and surface soil final status radiological surveys for demonstratingcompliance with dose or risk-based regulations or standards. The WPDD also identified
the same principles as MARSSIM in addition to the benefit of taking a “holistic approach”
to characterisation so that all aspects are considered. In general, the term radiologicalcharacterisation represents the determination of the nature, location and concentration ofradionuclides in a nuclear facility or waste material. The United States Environmental
Protection Agency (US EPA) and NEA characterisation methodologies described in thischapter are not designed for only conventional radioactive waste or only UL radioactive
waste. They describe principles and methods that can generally be applied to thecharacterisation of any solid radioactive waste. The scope of the characterisation
undertaken will depend mainly on the characterisation objectives, which follow from what
prompted wanting to characterise the waste, and on level of understanding of the wastes
that are to be characterised.

Chapter 3. Case studies
Members of the EGCUL presented experience of characterising unconventional/legacy
radioactive waste from five countries: France, Japan, Ukraine, Russia and the United
Kingdom. These include the following case studies:•

CEA Cadarache Centre, France;•

Fukushima Daiichi, Japan;•

network of final isolation sites for radioactive waste called Radon, Russia;•

characterisation of legacy waste produced at Rivne Nuclear Power Plant, Ukraine;•characterisation of radioactive waste and material from Chernobyl Nuclear Power
Plant, Ukraine;•characterisation of radioactive waste and material at the Shelter Object (SO),Ukraine;

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safety assessment of “Podlesnyi” radioactive waste storage facility including thecharacterisation of radioactive waste and material at module A, Ukraine;•

safety assessment of “radioactive waste temporary localization point” within
Chernobyl Exclusion Zone, Ukraine;•

Good Practice developed for the characterisation of solid radioactive waste, United
Kingdom.

The case studies comprise of the situation and objectives, the regulatory framework
including waste acceptance criteria (WAC), challenges and lessons learnt. They have been
produced to provide the perspectives of those involved in planning and delivering
radioactive waste characterisation on a variety of different sites which each have their own
challenges. The challenges and lessons learnt provide valuable insights which will bebeneficial to those developing characterisation strategies, objectives and plans as well as
highlighting the importance of having a flexible regulatory framework and supporting
infrastructure.

Chapter 4. Challenges and lessons learnt derived from the case studies
The similarities between the methods used in each of the case studies with those discussed
in Chapter 2 are summarised. The challenges and lessons learnt in each of the case studies
are summarised and discussed in more detail. The key themes discussed can be divided into
two main groups comprising of the radioactive waste management infrastructure and thestrategy which might be classified as the “enablers”, and the technical aspects.
Enablers include:•

radioactive waste management framework;•

waste classification and categorisation that includes UL waste;•

waste acceptance criteria for the treatment, storage and disposal of UL waste;•

defining end-states;•

defining priorities on complex sites and safety analysis.

Technical aspects include:•

characterisation project plan including sampling and analysis;•

development of nuclide vectors (fingerprints) for UL waste;•characterisation reporting and review;•

retrieval, conditioning and packaging of UL waste.

Chapter 5. Characterisation methodology and adaptations and enhancements toaddress UL waste challenges
This chapter focuses on the key differences between UL waste and conventional waste andhow the existing characterisation methodologies can be adapted or enhanced to cater forthis. The key themes that were explored are:•site prioritisation – using screening analysis to define characterisation objectives;•characterisation objectives – using a “life cycle” and “holistic” approach;
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 15•

historical information – what is likely to be available and how this can be used;•site reconnaissance and monitoring – techniques that can be applied;•

sampling plan and strategy – how the Data Quality Objectives (DQO) process canbe used and adapted, and the difficulty with taking representative samples;•

geostatistics – how this technique can be used for legacy and post-accident sites;•

developing radionuclide vectors (fingerprints) – things to consider for UL waste;•decision-making – the importance of defining “end-states”;•

regulatory
framework
includingwaste
acceptancecriteria
andcategorisation/classification of waste – the importance of the overall waste
management framework.

Figure 5.1. provides a summary and illustrates some of the adaptations and enhancementsthat can be made to the methodologies discussed in the case of characterisation of accident
waste based on the experience from Japan and the case of Fukushima Daiichi. At the sametime, the case studies show that a phased approach to characterisation may be needed andthat iteration of the process of characterisation may be needed at each decision point in thedecommissioning or remediation programme. Figure 5.2 illustrates the style of road map
and interactions that are likely to be relevant, based on the above discussion.

Chapter 6. Conclusion and recommendations
The radioactive characterisation methodologies designed for conventional waste can beapplied to UL waste with some adaptations and enhancements. It is important that theradioactive characterisation method is supported by a flexible legislative and regulatory
framework when there are significant uncertainties in the composition of the waste you aredealing with. These uncertainties need to be communicated in a way that all stakeholders
understand so that there is no ambiguity in the overall strategy and that a “life cycle”,“holistic” approach to characterisation can be developed. This will ensure that the available
resources are optimised and enable a comprehensive R&D programme to be developed tosupport radiological characterisation and ultimately decision-making.
The Expert Group has identified the following areas for further development:•setting a clear strategy underpinned by a regulatory framework to implement thepolicy;•having an integrated waste management strategy to optimise resources and time;•

having a clear and flexible waste classification system and WAC such that UL
waste could be included and that addresses radiological and other hazardous
content of the waste in a proportionate manner;•having a flexible and adaptable sampling methodology;•

understanding the end-state goal prior to undertaking characterisation activities;•

early and open dialogue with all stakeholders;•

understanding that characterisation is a continuous process;•having a clear review and validation process.

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Chapter 1. Introduction, objectives and scope
1.1. BackgroundRadioactive waste management requires planned and systematic actions to provide
confidence that the entire system, processes and final products will satisfy given
requirements for quality. To ensure the quality of the end product, it is necessary to know
and control the physical, chemical and radiochemical parameters within the entire waste
management life cycle, with special emphasis on waste conditioning, storage and disposal.
Testing and analyses to demonstrate the radioactive content and the quality of final waste
forms and waste packages are key components of this knowledge and control and areessential to an accurate characterisation of the waste.
The characterisation process is dependent on setting clear characterisation objectives andgathering the right information to underpin the decisions that need to be taken to manage
the waste. Conventional radioactive waste is understood to be operational or
decommissioning wastes produced as part of planned operations or decommissioning. The
provenance and material properties of the waste, and the process that has led to its
generation is understood. Generally, studies will have been undertaken prior to the waste
generation to identify the optimum solution for managing the waste, including its final
disposal. In relation to legacy and unconventional (UL) waste there are additional
challenges in setting these objectives because:•

The provenance and material characteristics of the waste for the most part is likely
to be less well defined.

•

There is likely to be limited information on how the waste was generated.

•

There are uncertain hazards and risks linked to the facilities that may have adegraded infrastructure or for which the infrastructure was never planned or in thecase of an accident the infrastructure has been destroyed.

•

Access to the waste for characterisation purposes is more likely to be difficult.

•

There is limited or no experience or information available from similar sites toinform the retrieval, treatment, transport, storage or packaging.

•

There are fewer equivalent wastes elsewhere, from which the properties of thewaste might be inferred. This means that it is more likely that techniques for theircharacterisation will not have been developed or are not commercially available.

•

The radionuclide fingerprint and radionuclide activity are likely to be more
uncertain. The same is generally true of non-radioactive hazardous contaminant in
the waste.

•

The radioactive waste is often heterogeneous and does not meet existing waste
acceptance criteria (WAC).

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Following the accident at the Fukushima Daiichi Nuclear Power Plant in March 2011,
various kinds of post-accident radioactive waste were generated. Specific management
methods/strategies are needed to manage this post-accident waste.
The NEA Radioactive Waste Management Committee (RWMC) established the Expert
Group on Fukushima Waste Management and Decommissioning R&D (EGFWMD) toevaluate the management of post-accident waste. As a first step, the expert group studied
lessons learnt from other nuclear accidents (e.g. Three Mile Island and Chernobyl)
including handling of any environmental contamination and the current status of related
waste management. The group then provided a strategic approach to the Japanese
government for effective management of radioactive waste from Fukushima Daiichi in2016 (NEA, 2016).
At the 50th Session of the RWMC, which took place in March 2017, the Japanese Nuclear
Damage Compensation and Decommissioning Facilitation (NDF) requested the RWMC
further assist in developing an integrated method for managing a large amount ofradioactive waste with unknown properties, focusing on radiological characterisation.
Many NEA countries (e.g. France, Germany, Norway, the Russia, the United Kingdom, theUnited States) have challenges in the management of legacy waste many of which are thesame as for UL waste. The development of a reliable and effective methodology formeasuring radiological, physical, chemical, radiochemical and other characteristics will bebeneficial for the management of UL wastes. The most effective way to do this is to refine
and adapt the current characterisation methodology used to manage conventional
radiological waste.
Several Japanese organisations have already started to develop a characterisation andcategorisation methodology for the Fukushima Daiichi waste, including the determinationof difficult-to-measure nuclides. There is also some experience gained from other countries
who have started to deal with legacy waste issues. In this regard, sharing information andexperiences in developing the methodology are beneficial for both international andJapanese organisations.
Alongside these activities, the Expert Group on Legacy Management, set up under theCommittee on Radiation Protection and Public Health, has considered the challenges innuclear and radiological legacy site management. A wide range of sites and facilities that
became radioactively contaminated because of past practices were discussed by the group.
Based on the experience from these case studies, guidance was prepared on thedevelopment of a practical and harmonised approach for the regulation of these sites.
However, significant challenges were recognised in applying such a framework because it
did not consider the difference in radioactive waste formation mechanism or nonradiological hazards (NEA, 2019).
The development of a reliable and efficient characterisation and categorisation
methodology is a common challenge in the fields of post-accident and legacy waste
management. Therefore, a strategic approach should be developed to manage thecomprehensive characterisation process of the waste including a sampling and analyticalplan. This should be supported by using statistical approaches, calculation methods andevaluation processes. State-of-the-art knowledge and experience in such characterisation
will be valuable to develop reliable and efficient characterisation and categorisation
methodology.

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1.2. Objectives and scope of work
The objectives of the EGCUL were:•

To share state-of-the-art knowledge and experience in characterisation on a large
amount of unknown waste.

•

To discuss relevant issues and challenges on characterisation in member countries.

•

Provide international feedback on Japan’s newly developed characterisation
methodology. As the characterisation methodology is still being developed this will
not be addressed in this report. However, the international experience presented inthis report should support the on-going development of Japan’s characterisation
methodology to be applied in the context of Fukushima Daiichi accident waste.

•

To explore possible links with the Committee on Decommissioning of Nuclear
Installations and Legacy Management (CDLM), and Committee on RadiologicalProtection and Public Health (CRPPH), and on the Expert Group on Legacy
Management (EGLM) on legacy management and radiological protection issues;and•To write a report to be submitted to the RWMC on state-of-the-art methodology
and best practice on the characterisation of a large amount of unknown waste.

The EGCUL output is intended to support the management of the complex wastes arisingfrom post-accident and legacy sites. The term “unknown waste” used in the EGCUL
mandate is considered to include both “legacy waste” or “unconventional waste” (terms in
the EGCUL title). The group decided not to apply any specific definition for these terms
but decided to adopt the following working descriptions 2:•

Legacy waste is radioactive waste that was generated by past activities and does
not comply with present criteria.

•

Unconventional waste is radioactive waste that was generated in an unexpected
situation and has unknown properties.

Radioactive waste generated from other industries other than nuclear and naturally
occurring radioactive material (NORM) waste were not considered in the scope of EGCUL.
The focus of this report is on characterisation for the safe management of the wastes, as far
as possible, taking a holistic view of the radiological and non-radiological hazards. Clearly,
management of safety has implications for logistical aspects of management.
The mandate of EGCUL commenced on 1 August 2018 and ran for 2 years.

1.3. Mode of operation of the group and structure of this report
Noting the above focus and scope, the EGCUL objectives have been met by the preparationand discussion of case studies from France, Japan, Russia, Ukraine and the United
Kingdom, supported by a range of other material from NEA member countries as well as
the wider activities of the NEA. Features of common approaches to conventional waste
characterisation are described in Chapter 2. Consideration has been given to mainstream
2.

This approach is consistent with work of EGLM and Challenges in Nuclear and Radiological Legacy Site
Management: Towards a Common Regulatory Framework (NEA, 2019), but the findings of the CDLM
sub-group on Terminology also needs to be considered.

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waste characterisation procedures and methodologies that have been developed primarily
for application in conventional decommissioning and contaminated land remediation work.
A variety of practical experiences are provided in the case studies in Chapter 3. These haveillustrated the need for the integration of waste characterisation with other aspects ofstrategic planning at these sites. The lessons learnt and challenges from each of the casestudies are explored further in relation to overall characterisation planning in Chapter 4. By
explicitly recognising challenges and lessons learnt, it has been possible in Chapter 5 to
identify opportunities for adaptations and enhancements to characterisation methods for the
management of UL solid radioactive waste. Conclusions and recommendations areprovided in Chapter 6 including areas of good practice that could be used in the preparationof international guidance on the application of the adaptations and enhancements.

References
IAEA (2007), Strategy and Methodology for Radioactive Waste Characterization, IAEA TECDOC 1537,
IAEA, Vienna, www.iaea.org/publications/7655/strategy-and-methodology-for-radioactive-wastecharacterization.
NEA (2019), Challenges in Nuclear and Radiological Legacy Site Management: Towards a Common
Regulatory Framework, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_40359
NEA (2016), Management of Radioactive Waste after a Nuclear Power Plant Accident, OECD Publishing,Paris, www.oecd-nea.org/jcms/pl_15008

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Chapter 2. Methodology features
This chapter provides a collection of various methodologies and good practice developed tocharacterise waste.

2.1. Differences between conventional and unconventional/legacy waste
The Expert Group on Unconventional and Legacy Waste (EGCUL) took the decision in its
first meeting not to develop precise definitions of the terms “legacy waste” or“unconventional waste”. These terms are not defined by international organisations (e.g. in
the IAEA safety glossary [IAEA, 2018] or by the NEA), largely because definitions ofthese terms at a national level have been found to be variable (NEA, 2019). However, theNEA has found it helpful to describe legacies in terms which correspond to thecircumstances in which the term is commonly applied (NEA, 2019). The same approach isadopted here. For use within EGCUL, the following understanding was used:•

Legacy waste is radioactive waste that was generated by past nuclear activities, andwhich does not comply with present legislation. Typically, these are wastes thatwere generated early in the history of a country’s nuclear programme, were notregulated in line with modern waste management and environmental protection
standards. For example, some legacy wastes were stored in vaults and silos withlittle information recorded on waste provenance or characteristics and with little
consideration of how the waste would eventually be recovered. Also, there was no
consideration of waste pre-treatment, treatment and conditioning for storage anddisposal. Therefore, the management of such wastes, to ensure safe storage in theircurrent facilities, to identify and implement optimised waste processing solutions
and to identify where and how the wastes are to be disposed, is challenging.

•

Unconventional waste is radioactive waste which was generated in an unexpected
situation resulting in a poor understanding of its properties. Such wastes are
generally produced because of incidents on a nuclear site. These incidents can cause
the spread of radionuclides and other hazardous materials and can result in thecontamination of the land surface and structures. Chronic, long-term, leaks ofradioactive liquids to ground – for example, from leaking pipes or structures – canalso generate significant volumes of below ground contaminated land.

Conventional and UL waste share some similarities as they come from the same range of
nuclear technology facilities and arise from the same basic processes linked to thefundamentals of the nuclear fuel cycle. However, significant differences may arise for thefollowing reasons:
1. The facilities, the sites on which they are located, and the adjacent areas affectedby historical operations are so old that relevant characterisation information wasnever required to be collected. Alternatively, information was collected in line withrequirements but has since been lost or was collected and has been retained but does
not satisfy modern day characterisation needs.

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2. The facilities (including older but also possibly newer ones) did not operateaccording to plan so that wastes have been generated that were not expected and,in some cases, cannot be fully characterised using the planned methods.
Higher activity UL wastes include for example, fuel cladding, fuel residues, sludges, flocs
and contaminated or activated metals. Lower activity UL wastes include for example,
contaminated land, soil, vegetation, structures and debris. The EGCUL took the decision
to focus on the characterisation of lower activity, solid UL wastes. The volumes of such
wastes may be substantial. For example, the Storage Management Plan for Fukushima
Daiichi estimates that approximately 700 000 m3 of lower activity waste from contaminatedrubble, soil, vegetation and used personal protective equipment (PPE) will be produced by
2030. For comparison, following the Chernobyl accident, it is estimated that between
400 000 m3 and 1 740 000 m3 of UL radioactive waste was generated (NEA, 2016). It isnoted that doses to workers involved in clean-up work are considered as planned exposures,
while doses to the public that occur in clean-up are considered as existing exposures andare subject to control through the application of Reference Levels, not dose limits andconstraints as identified in the IAEA International Basic Safety Standards (IAEA, 2014)
and the International Commission on Radiological Protection (ICRP) publication 103(ICRP, 2007).
“Conventional” radioactive waste is understood to be operational or decommissioning
wastes produced as part of planned operations or decommissioning. The provenance andmaterial properties of the waste, and the processes that led to its generation, will be wellunderstood. Generally, studies will have been undertaken prior to waste generation to
identify the optimum solution for managing the waste, including its final disposal.
The differences between conventional waste and UL wastes and the challenges that thesedifferences present were set out in chapter 1.The consequence of these differences is that,
before the start of waste characterisation, less is usually known about the characteristics of
UL wastes than of conventional wastes.

2.2. Toolboxes of common approaches
This section summarises some recognised international good practice for thecharacterisation of solid radioactive waste. These characterisation approaches are notdesigned for only conventional radioactive waste or for UL radioactive waste. They
describe principles and methods that can generally be applied to the characterisation of any
solid radioactive waste. The scope of the characterisation undertaken will depend mainlyon the characterisation objectives that follow from asking: “Why do we want to characterise
the waste?”
Different groups working within a radioactive waste management programme may havedifferent specific objectives, for example, historically, those concerned with waste storage
have not necessarily been concerned with waste disposal. The information required to meetthe characterisation objective can be quite different in these two areas. To answer thequestion from a top-down approach, the objectives driven perspective is that the waste mustby managed safely, securely and efficiently.

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Thus, characterisation is needed:•

To support decisions on treatment options, storage and final disposal. Suchdecisions should complement or inform the application of the national radioactive
waste strategy.

•

To allocate the waste to one of the national waste classification schemes waste suchas exempt/cleared, very low-level waste (VLLW), short-lived low-level waste(SL-LLW), long-lived low-level waste (LL-LLW), intermediate-level waste (ILW)or high-level waste (HLW).

•

For a site or facility in question, or for a site or facility which has yet to be identified,
all as part of a wider national radioactive waste management policy and strategy.
This needs to be completed with a view to communicating that this information to
all stakeholders.

The above scheme focusses on characterisation that supports management of radioactive
waste from a radiation and nuclear perspective. It also assumes that there already exists anational radioactive waste classification scheme and associated strategies for managementof waste, including disposal, in each waste class.
Experience shows that such arrangements are commonly in place for planned waste but not
fully the case for UL waste. In particular, the application of waste acceptance criteria for
conventional treatment, transport, storage and disposal may not be possible for UL waste,requiring variation in methods and revision of safety arrangements and their regulatory
supervision (NEA, 2016). This is illustrated in the French case study in Chapter 3 of thisreport. In addition, the waste may contain materials which are hazardous in a nonradiological context, e.g. chemically hazardous. It will generally be necessary tocharacterise the waste to understand those other hazards, and support decisions on
management that address the main hazard or risk (NEA, 2016, Chapter 4).
It is therefore necessary to consider what needs to be known about the waste in each ofthese areas and how that information needs to be organised to be met most efficiently.
Additional questions that follow comprise of:•What can be measured that provides direct information about what needs to be
known (preferred)?

•

What can be measured from which something relevant can be inferred (e.g.
fingerprints for difficult-to-measure radionuclides, data necessary to showcompliance with clearance levels)?

•

What other information could be relevant?

The most appropriate solution will also be influenced by the extent to which thecharacteristics of the waste are understood before characterisation begins. Also, following
an incident, access to the waste may be more problematic than for conventional radioactive
wastes. These last two points (limited understanding of waste characteristics and challenges
with accessing the waste) are in summary the major differentiators between conventional
waste and UL waste.
A staged or phased characterisation process is needed for any complex decommissioningor legacy project, with further decisions on next step being taken in the light of the mostrecent information, e.g. on the wastes (NEA 2019). The characterisation method shouldtake account of this. Chapter 3 of this report presents case studies of the characterisation of
UL waste and Chapter 4 presents the lessons learnt from the case studies.
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2.3. Data Quality Objectives process
The US EPA has developed the Data Quality Objectives (DQO) process as a systematicplanning tool to determine the type, quantity and quality of data needed to support
defendable decisions. The DQO process (US EPA, 2006) was developed to guide the costeffective and resource-effective acquisition of environmental data. It has since been
adapted for use in the nuclear industry in many countries for a wider range of tasks.
An important purpose of DQO is to design data acquisition plans that ensure thecharacterisation results support decisions at a required level of confidence or enableestimation of waste characteristics with the required level of uncertainty. These “dataquality objectives” can relate to issues such as (i) how well the characterisation data reflect
any spatial variability in the content of hazardous substances, (ii) the accuracy of theindividual measurements, and (iii) the overall uncertainty, or alternatively confidence, in
the results. DQO always takes account of the required confidence or uncertainty level.
Where high confidence or low uncertainty is needed (e.g. to prove a regulatory limit has
been met), DQO will help design a data acquisition plan with an appropriate distributionof sample/survey points that reflect the anticipated spatial variability in e.g. radionuclide
concentrations. Statistical approaches are used in DQO to design data acquisition plans
with the optimal number of sample/survey points. This generally leads to a requirement for
fewer samples/in situ measurements compared to conventional grid designs. This enables
characterisation to be faster, cheaper and more reliable.
The DQO planning process is used in the nuclear industry in three main ways:
1. During the early stages of decommissioning, to develop data acquisition plans forthe initial characterisation of materials, equipment, buildings or land suspected or
known to be contaminated.
2. During decommissioning or operations, to determine whether the concentrations ofcontaminants in waste materials fall above or below specific limits (e.g. as definedby the WAC for a waste disposal or treatment facility) so that wastes canappropriately managed.
3. Towards the end of a decommissioning, to develop data acquisition plans todetermine whether decommissioning, remediation and/or decontamination ofmaterials, equipment, buildings or land has achieved specified clean-up targets.
These applications are similar but, in the first case, the purpose is to determine a “bestestimate” of the concentrations of contaminants in the materials or wastes. In the other twocases the purpose is to test whether the concentrations of contaminants are above or below
a specific limit, to support a waste management decision.
The steps in the DQO process are shown in Figure 2.1. Each step defines criteria that will
be used to establish the final data collection design. The diagram shows the steps as
sequential, but the iterative nature of the DQO process allows one or more of the steps tobe revisited as additional data are collected.

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Figure 2.1. Data Quality Objectives methodology
Step 1. State the problem.
Define the problem that necessitates the study; identify the planning team, examine budget, schedule

Step 2. Identify the goal of the study.
State how the data will be used in meeting the objectives and solving the problem, identify study questions, define alternative
outcomes

Step 3. Identify information inputs.
Identify data and information needed to answer study questions

Step 4. Define the boundaries of the study.
Specify the target population and characteristics of interest, define spatial and temporal limits, scale of interference

Step 5. Develop the analytic approach.
Define the parameter of interest, specify the type of inference and develop the logic from drawing conclusions from findings

Decision making(hypothesis testing)

Estimation and otheranalytic approaches
Step 6. Specify performance or acceptance criteria.

Specify probability limits for false rejection and false
acceptance decision errors

Develop performance criteria for new data being collected or
acceptable criteria for existing data being considered for use

Step 7. Develop the plan for obtaining data.
Select the resource-effective sampling and analysis plan that meets the performance criteria

Source: US EPA, 2006.

Steps 1 to 5 are focused on identifying qualitative criteria such as the nature of the problem,the decisions that will need to be made, the types of data needed and the logic that will beapplied to draw conclusions from the data. Step 6 establishes acceptable quantitative
criteria on the quality and quantity of the data to be collected, relative to the ultimate useof the data. These criteria are known as performance or acceptance criteria, or DQOs. For
decision problems, the DQOs are typically expressed as tolerable limits on the probabilityor chance (risk) of the collected data leading the operator to make an erroneous decision.
For estimation problems, the DQOs are typically expressed in terms of acceptable
uncertainty (e.g. width of an uncertainty band or interval) associated with a point estimate
at a desired level of statistical confidence. Step 7 involves the design of a data collection
plan.
The benefit of applying DQO to waste characterisation is that it should ensure waste
management decisions are made consistently based on sufficient and appropriate data. The
number of samples that are needed to characterise wastes to an agreed level of confidence
is determined, which means that the effort and expense of sampling and analysis isminimised and justifiable. These characteristics are especially important for UL wastes.

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The outputs of the DQO process are used to develop a Quality Assurance (QA) Plan andfor performing Data Quality Assessment (DQA).

2.4. Data Quality Assessment
The US EPA has produced DQA guidance to determine if data obtained fromenvironmental data operations are of the right type, quality and quantity to support theirintended use. The process is illustrated in Figure 2.2. Although the guidance focuses onenvironmental decision-making, the tools presented for preliminary data review andverifying statistical assumptions are useful. They are, therefore, appropriate to theevaluation of characterisation data obtained from UL waste.
Figure 2.2. Data Quality Assessment (DQA) process
Step 1. Review the project’s objectives and sampling design.
Confirm the objectives are still valid. Review the sampling design and data collection against the project objectives, noting
consistency or any potential discrepancies

Step 2. Conduct a preliminary data review.
Review QA reports for the validation of data, construct basic statistics and generate graphs of the data. Use this to learn about
the structure of the data and to identify patterns, relationships or potential anomalies
Step 3. Select the statistical method.
Select appropriate procedures for summarising and analysing the data, based on the review of the performance and acceptancecriteria associated with he project’s objectives, the sampling design and the preliminary data review. Identify the key underlying
assumptions associated with the statistical test

Step 4. Verify the assumptions of the statistical test.
Evaluate whether he underlying assumptions hold, and whether departures are acceptable, given the actual data and otherinformation about the study

Step 5. Draw conclusions from the data.
Peform the calculations pertinent to the statistical test, and document the conclusions to be drawn as a result of thesecalculations. If the design is to be used again, evaluate the performance of the design

Source: US EPA, 2000.

The fundamental premise of DQA is that data quality, as a concept, is only meaningfulwhen it relates to the intended use of the data. It is necessary to know what the data are to
be used for before judging whether the dataset is adequate.

2.5. MARSSIM, MARSAME and MARLAPThe US EPA has produced two manuals: Multi-Agency Radiation Survey and Site
Investigation Manual (MARSSIM) and Multi-Agency Radiation Survey and Assessment ofMaterials and Equipment (MARSAME). These two manuals apply the DQO and DQA
methodologies to radioactively contaminated sites, materials and equipment. Each manualis concerned only with the characterisation of radiation-derived hazards, radionuclide
concentrations, dose rates and measurements from contamination surveys. Each manualpresents a method for planning, implementing, assessing and making decisions about
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regulatory compliance. Although designed to be applied under US EPA regulations, thegood practices in the methodology described in these documents can be adapted andsimplified to be applied under the regulatory frameworks of other countries.
MARSSIM is applicable to surveys of surface soil and building surfaces to determine, withan associated level of confidence, whether the site meets the release criteria for radioactive
contaminants. Focus is on the final status survey, which is undertaken after the completionof decommissioning, remediation and/or decontamination. MARSSIM does not cover othermedia (such as subsurface soil, surface or subsurface water and construction materials) orchemical contamination.
MARSAME is applicable to surveys of potentially or known radioactively contaminatedmaterial and equipment (M&E) to determine proper disposition of the M&E. M&E either
meets the clearance criteria for the site (i.e. can be removed from control under radioactive
substances regulation) or requires “interdiction”, which involves an increase in the level ofradioactive control or a decision not to accept control from another party. Examples of
M&E comprise of metals, concrete, tools, equipment, piping, conduit, furniture anddispersible bulk materials such as rubbish, rubble, roofing materials and sludge. Liquids,
gases and solids stored in containers (e.g. drums of liquid, pressurised gas cylinders andcontainerised soil) are also included in the scope of the manual.
A third document, the Multi-Agency Radiological Laboratory Analytical Protocols(MARLAP) manual, provides guidance for the planning, implementation and assessment
phases of projects that require laboratory analysis of radionuclides. It details the need for aconsistent approach to producing radioanalytical laboratory data that meet a project’s orprogramme’s data requirements.
MARSSIM and MARSAME manuals contain details of the planning, implementation,assessment and decision-making methodologies. In each case, the seven steps of the DQO
process are followed (see Figure 2.1.), which results in the following general approach.
•

Available information and data are used to rank or classify the land, material andequipment based on the likelihood that radioactive contaminants would result in
dose- or risk-based action levels being exceeded. The output is a sub-division of
the land, material and equipment into “survey areas”, each with an associatedranking or classification.

•

This ranking or classification is used to determine the level of survey effort
expended on each survey area. The highest level of effort is expended on survey
units with the greatest likelihood of exceeding the action levels. The numbers andlocations of samples or survey points in each survey area are determined. A key
input to deciding the number of measurements is the required level of confidence
in the decision to be made.

•A pass/fail statistical test for demonstrating compliance with the action level isselected. A data acquisition plan is then developed for each area, and the survey isundertaken. The results are then evaluated using the identified statistical test and adecision is made at the required level of confidence.

Both MARSSIM and MARSAME are used principally for making decisions on whether toclear land, materials and equipment from control under radioactive substances regulation.
There is no US EPA manual to describe planning, implementation, assessment anddecision-making or uncertainty estimation for a characterisation programme that has theobjective of selecting an appropriate waste management option; for example, determining

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whether radioactive waste meets the waste acceptance criteria (WAC) for a given treatment
or disposal facility. However, the principles of DQO and DQA, together with some of thesurvey methodologies described in MARSSIM and MARSAME, can be used to achievesystematic and justifiable planning, implementation, assessment and decision-making forcharacterising UL waste. Characterisation is likely to require the determination of theconcentrations of chemical contaminants and the physical and chemical properties of thewaste, in addition to determination of radiation-derived hazards.

Examples of relevant software to support design and interpretation ofcharacterisation activities
The Visual Sampling Plan (VSP) is a software tool developed by the Pacific Northwest
National Laboratory with support from, among others, the US EPA. It supports the designof data acquisition plans (sampling and/or surveying) that collect the right type, quality andquantity of data to support decision-making at the required level of confidence. Several
statistical sampling designs can be selected, including random, systematic, stratified andcombined judgement/probabilistic. Locations of samples are determined in VSP based on
the sampling design and the required number of samples. In addition to the sampling design,
VSP also provides data quality assessment and statistical analysis functions to support
evaluation of data and make decision recommendations.

2.6. NEA Working Party on Decommissioning and Dismantling
Within the NEA, the Working Party on Decommissioning and Dismantling (WPDD)
provides a focus for the analysis of decommissioning policy, strategy and regulation,
including the related issues of management of materials, release of buildings and sites from
regulatory control and associated cost estimation and funding. Beyond policy and strategy
considerations, the WPDD also reviews practical considerations for implementation suchas techniques for characterisation of materials, for decontamination and for dismantling.
The WPDD brings together senior experts in decommissioning from NEA and observercountries, with involvement also from other international organisations such as theEuropean Commission and the International Atomic Energy Agency (IAEA). It tracks
decommissioning developments worldwide and develops reports and position papers on
emerging issues. Its overarching aim is to contribute to the development of good practicethrough circulation of its reports and through dialogue between policy makers, practitioners,
regulators, researchers and international organisations.
One of the WPDD Task Groups is the Task Group on Radiological Characterisation andDecommissioning (TGRCD). Phase 1 of TGRCD addressed strategies for radiologicalcharacterisation in decommissioning of nuclear facilities and concluded in 2013 with the
publication of a Final Report (NEA, 2013). In the Final Report from Phase 1, “radiologicalcharacterisation” is concerned solely with determination of radionuclide concentrations andradioactivity (dose rate, measurements from contamination surveys, etc.). The deliverablefrom Phase 1 of the TGRCD also included a workshop held in Studsvik, Sweden in April
2012 (NEA, 2012). Phase 2 of the TGRCD addressed strategies for optimisation of nuclear
facility characterisation in a waste and materials end-state perspective. It commenced in2014 and concluded in 2017 with the publication of a final report (NEA, 2017). In the Final
Report from Phase 2, “radiological characterisation” is expanded to consider physical,
chemical and biological properties and characteristics of the waste. This is termed “holistic
characterisation” and is discussed further below (NEA, 2017).

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Both final reports identify good practice for radiological characterisation at different stagesof decommissioning and point to areas that could or should be developed further byinternational co-operation and co-ordination. A summary of the good practices from the
NEA report is summarised in the following sections of the report.

Characterisation objectives
It is crucial to understand the objectives of characterisation. Without understanding
characterisation objectives, it is not possible to determine the types, quantities and qualityof data/information required. The general objectives of characterisation are presented in
Section 2 of “Radiological Characterisation for Decommissioning of Nuclear Installations:Final Report of the Task Group on Radiological Characterisation and Decommissioning(RCD) of the Working Party on Decommissioning and Dismantling (WPDD)” (NEA,2013). Applying a structured process to define objectives is relevant to the characterisationof UL waste. However, recognising that not all characterisation is not undertaken to support
a decision but can also be for “estimation” as shown in Figure 2.3. Depending on thecharacterisation objective, quantification of uncertainty would not be required for all waste
characteristics.
The objectives of characterisation will change over the life cycle of the facility, assummarised in Table 2.1 and illustrated in Figure 2.3.
Table 2.1. Key characterisation objectives carried out in each phase

Phase

Work carried out

SitingBaseline surveys are undertaken to determine background radiation levels.

Construction

Construction materials are retained to support future activation calculations
and to define the natural radioactivity background.

Operation

Surveys are undertaken regularly for establishing dose rate and contaminationlevels, with additional surveys and measurements being required following
incidents involving plant contamination.

Transition

Detailed radiological surveys and measurements are required to support thedevelopment of the final decommissioning plan.

DismantlingRadiological characterisation goes hand in hand with all issues of safety anddose assessment, radioactive waste management, clearance of materials andbuildings, etc.

Licence termination(closure)

A final survey of the site and any remaining structures will be needed tosupport an application for release of the site from regulatory control.

The description of characterisation during the siting and construction stages is goodpractice and should be applied to new build facilities. However, this characterisation wasprobably never done for old facilities, so this information will be missing when they aredecommissioned whether there was an incident or legacy waste exist.

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Figure 2.3. Typical radiological characterisation objectives during the life cycle of a nuclear
facility

Source: NEA, 2013.

In principle, all life cycle stages also apply to the characterisation of UL waste. However,in addition to the probable lack of information during the siting and construction stages, it
must be recognised that the generation of UL waste is either unplanned or is planned with
only limited consideration of subsequent waste management. Consequently,
characterisation information from stages up to and including “transition” is likely to belimited or absent for UL waste. Most characterisation will occur during the “dismantling”
stage, and the scope of this characterisation should be suitable to meet the objectives fromearlier stages.
The publication Radiological Characterisation from a Waste and Materials End-state
Perspective: Practices and experience (NEA, 2017) proposes a “life cycle” approach tocharacterisation. In this approach, design of the characterisation programme at any stage inthe life cycle considers how the data collected might be used in subsequent life cycle stages.
This may well result in small changes to the characterisation design to maximise use of
information for a subsequent purpose. For example, when planning radiologicalcharacterisation for systems and installations, it is a good idea to devise the characterisation
plan in such a way that it may also provide information for the subsequent characterisationof building surfaces. However, when considering whether information can be used forseveral purposes, it is important to recognise that different characterisation objectives will
have different information requirements, for example different limits of detection ormeasurement uncertainty.

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It is beneficial to take a holistic approach to characterisation so that all aspects are
considered. In general, the term “radiological characterisation” represents thedetermination of the nature, location and concentration of radionuclides in a nuclear facility
or waste material. However, properties other than the pure radiological are important andconsequently, within both a high-level strategy and characterisation plans, characterisation
should be considered in the widest sense covering radiological, physical, chemical andbiological properties. For example, characterisation should include understanding the:•Volumes and characteristics of the waste or material in a nuclear facility. This could
include the volume and masses of contaminated and potentially contaminatedmaterials or waste, their physical form and their material characteristics(e.g. rheology of sludges and the presence of reinforcing bars in concrete structures).

•

Chemical characteristics of a nuclear facility and its associated solid wastes. Thesecan significantly impact on the decommissioning plan and the optimisation of waste
segregation and disposal plans (e.g. from worker safety and WAC standpoints).

•

Presence and the chemical form of reactive metals such as sodium, magnesium andaluminium. This can be important with respect to waste treatment, storage anddisposal.

•

Biological properties. For example, organic materials can generate organic-richsludges, biologically hazardous waste streams and bulk gases.

Structured approach to survey design
The report “Radiological Characterisation for Decommissioning of Nuclear Installations:Final Report of the Task Group on Radiological Characterisation and Decommissioning(RCD) of the Working Party on Decommissioning and Dismantling (WPDD)” (NEA, 2013)
suggests a structured sequence of seven steps to develop a survey design that meetscharacterisation objectives. The steps are consistent with those proposed by the US EPA asshown in Figure 2.1.
1) State the problem. Concisely describe the problem to be studied. Review priorstudies and existing information to gain a sufficient understanding to define theproblem.
2) Identify the decision. Identify what questions the radiological characterisation
campaign will attempt to resolve, and what actions may result.
3) Identify inputs to the decision. Identify the information that needs to be obtained
and the measurements that need to be taken to resolve the decision statement.
4) Define the boundaries of the characterisation campaign. Specify the time periods
and spatial area to which decisions will apply. Determine when and where data
should be collected.
5) Develop a decision rule. Define the parameter of interest and specify appropriate
limits of detection.
6) Specify limits on decision errors. Define the tolerable decision error rates based ona consideration of the consequences of making an incorrect decision.

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7) Optimise the survey design. Evaluate information from the previous steps, consider
alternative data collection designs and choose the most resource-effective designthat meets all objectives.
As noted above, not all characterisation is undertaken to support a decision. Depending on
the characterisation objective, quantification of uncertainty would not be required for all
waste characteristics.

Stepwise approach to characterisation
The report “Radiological Characterisation for Decommissioning of Nuclear Installations:Final Report of the Task Group on Radiological Characterisation and Decommissioning(RCD) of the Working Party on Decommissioning and Dismantling (WPDD)” (NEA, 2013)
recognises that careful planning of radiological characterisation is necessary and proposesthat a radiological characterisation programme or campaign normally comprises thefollowing steps.
1) An initiating step, where the targets of the campaign are defined and, if necessary,
where consent from the competent authority is gained.
2) A planning step, where historical information from the facility is evaluated andwhere the strategy and the plan for data acquisition is developed.
3) An implementation step, where data acquisition is carried out, if necessary aided by
calculation methods, e.g. for determination of activation.
4) A data assessment and evaluation step, in which the various results are interpreted
and reviewed, statistical evaluation of measurement results is carried out, etc., andwhere calculated results and measured data are compared.
5) A finalisation step, where the results are documented and (if necessary) reported tothe competent authority and are used for the various purposes and objectives forwhich they have been derived.
The report recommends that planning for characterisation take place early in the life cycleof the facility.

Optimisation of the approach to data collection
The first step regarding characterisation is to determine whether additional characterisation
data are required, or whether information obtained during the planning step (see above) is
sufficient and suitable to meet the characterisation objectives. If further information isrequired, the publication, Radiological Characterisation from a Waste and Materials Endstate Perspective: Practices and experience (NEA, 2017) recognises that there are several
approaches for radiological characterisation. The properties can be:•measured through collection of samples and subsequent ex situ laboratory analysis;•

measured through in situ measurement; for example, dose rate; gross alpha, beta
and gamma contamination levels; gamma spectrometry;•

inferred by calculation in the case of activation products;

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inferred by estimation, based on an understanding of relevant provenance andprecedence. “Provenance” concerns understanding of where the material has comefrom and the processes that have generated it. “Precedence” concerns information(data, documents, experience) from other comparable sites who have previously
undertaken relevant characterisation.

This is illustrated in Figure 2.4. Further information is provided in Sections 5 and 6 of thereport.
Only proven analytical and modelling techniques should be used. When selecting analyticaltechniques, consider the required limits of detection and accuracy, and any constraints on
measurement time. These will be apparent from the characterisation objectives. For
example, characterisation ahead of decommissioning does not have the same purpose asmeasurements during clearance procedures; this may lead to higher limits of detection
being acceptable and a wider range of analytical techniques being available.
Lastly, the balance between sampling/laboratory analysis and in situ measurement shouldbe evaluated for each characterisation programme. Materials and wastes with radionuclide
vectors that contain gamma-emitting radionuclides could be characterised in situ at large
scale using gamma spectrometry, although some samples would need to be collected andanalysed in a laboratory to determine the presence of non-gamma-emitting radionuclides
and to derive the radionuclide vectors (fingerprints). Conversely, materials containing purealpha emitters or beta emitters that emit only weak gamma radiation would require
sampling and subsequent laboratory analysis. Sampling and laboratory analysis are also
appropriate to characterise contaminants that have penetrated the matrix of materials like
concrete and to evaluate the accuracy of activation calculations.
Figure 2.4. Characterisation approaches

Source: NEA, 2017.

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The publication, Radiological Characterisation from a Waste and Materials End-state
Perspective: Practices and experience (NEA, 2017) describes the main components of thedata assessment and evaluation process, as illustrated in Figure 2.5.
Figure 2.5. Data assessment and evaluation process

Source: NEA, 2017.

The main components of data assessment and evaluation are summarised below.
•

Data verification and validation, which should be undertaken by experienced staff.
Data should have been collected according to a Quality Plan, which specifies themethods, equipment, calibration and checking procedures for data collection. Data
verification and validation is used to accept, reject or qualify data in a consistentmanner, recognising that data may have been collected from multiple sources andwith different levels of confidence. This includes audits to confirm compliance with
operating procedures; audits of documents, data and equipment; checks on datatranscription, and tracking/checking the performance of measuring and sampling
equipment. Validation involves checking that data have been collected in line withthe characterisation plan, determining the acceptability of data, and ensuring that
any identified corrective actions have been taken.

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Preliminary data assessment and review. This initial step involves identifying
whether data are suitable for subsequent evaluation. The output is a dataset forevaluation. The selection of data evaluation techniques should be used also be madeat this stage. The publication Radiological Characterisation from a Waste andMaterials End-state Perspective: Practices and experience (NEA, 2017) identifies
several commonly used data evaluation techniques: statistical techniques;
numerical modelling approaches, and geostatistical techniques. Visualisation byuse of Geographical Information Systems (GIS) can make data more accessible toother stakeholders, by summarising information, but care must be taken not to overinterpret such visualisations.

•

Verification of assumptions and uncertainty quantification. This step might require
further statistical analysis to determine whether assumptions regarding theunderlying distribution of the data are correct, and analysis of the uncertainty
associated with measurements.

•

Drawing conclusions from the data. This involves application of the chosen data
evaluation techniques to meet the characterisation objective, which may be adecision regarding waste management or an estimation of the quantities of waste indifferent categories.

Summary of key differences between characterisation of conventional and UL solidradioactive waste
The US EPA and NEA characterisation methodologies described in Section 2.2 are notdesigned for only conventional radioactive waste or only UL radioactive waste. They
describe principles and methods that can generally be applied to the characterisation of any
solid radioactive waste. The scope of the characterisation undertaken will depend mainlyon the characterisation objectives, which follow from asking why we want to characterise
the waste, and on level of understanding of the wastes that are to be characterised. The key
difference between characterisation of conventional and UL solid radioactive waste aresummarised below.
Differences in characterisation objectives:•

Arrangements are commonly in place for the management of conventional
radioactive waste, but this may not be the case for UL waste. For example, existing
waste acceptance criteria for conventional treatment, transport, storage and disposal
may not be appropriate for UL waste, requiring variation in methods and revisionof safety arrangements and their regulatory supervision. Characterisation would be
required to determine the optimum waste management solution for UL waste.

Differences in the level of understanding of the waste prior to characterisation:•

Prior to the start of waste characterisation, less is usually known about thecharacteristics of UL wastes than of conventional wastes. Possible reasons for thisinclude: less understanding of the provenance and material characteristics of UL
waste; the process that generated UL waste is less well-understood than forconventional waste; access to the waste for characterisation purposes is more likely
to be difficult in the case of UL waste, and; there are fewer or no equivalent wastes
elsewhere, from which the properties of the UL waste might be inferred.

•

Although a life cycle approach to characterisation is in principle applicable to bothconventional and UL radioactive wastes, life cycle characterisation information
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from stages up to and including “transition” is likely to be limited or absent for UL
waste. This is because generation of UL waste is either unplanned or is plannedwith only limited consideration of subsequent waste management.
•

The precedence and provenance information on UL waste may be lacking, therefore,
most emphasis for characterisation of UL waste will be on measurement.
Notwithstanding this, because more limited information is available prior tocharacterisation of UL waste, it may be appropriate to exercise additional caution
when planning data acquisition activities.

2.7. Interaction with stakeholders
Stakeholder engagement and working with contractors is an important area of consideration.
They are key enablers and should be considered when setting the strategy fordecommissioning (NEA, 2018). Stakeholder opinions and expectations must be consideredin the decision-making process to realise the best outcome and to minimise the risk that thecharacterisation work will not meet its objectives. Providing too little information may leadto unfounded expectations. Paucity of information, or poorly presented information, may
lead to distrust and/or loss of credibility. Therefore, the participation process must be coordinated and discussed with the stakeholders from the outset of the characterisation work.
When working with contractors there is a need to engage early so that the characterisation
planning is collaborative therefore helping to prevent any misinterpretation of information.
The same is true for regulators, particularly when they will need to take regulatory
decisions informed by the characterisation information. A good means to achieve this is todevelop (with all relevant stakeholders), execute and maintain a stakeholder engagement
plan from the start of the characterisation work.
For the successful engagement, it is important to identify the stakeholders and develop amanagement plan based on a clear communication strategy. The key stakeholders thatmight influence the strategy for decommissioning are shown in Figure 2.6. Managed andstructured communication with all stakeholders will ensure that people are informed of thestrategic decisions taken on the decommissioning project and provide them with theopportunity to comment and input on the planning phase (NEA, 2018). ISO/DIS11000
provides a guide for organisations covering the requirements of a strategic life cycle
framework to establish and improve collaborative relationships in and between
organisations.

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Figure 2.6. Stakeholders and how they might influence the strategy for decommissioning

Source: NEA, 2018.

References
IAEA (2018), IAEA Safety Glossary: Terminology Used in Nuclear Safety and Radiation Protection,2018 Edition, International Atomic Energy Agency, Vienna, www.iaea.org/publications/11098/iaeasafety-glossary-2018-edition.
IAEA (2014), Radiation Protection and Safety of Radiation Sources: International Basic Safety
Standards, General Requirements Part 3, International Atomic Energy Agency, Vienna,
www.iaea.org/publications/8930/radiation-protection-and-safety-of-radiation-sources-internationalbasic-safety-standards
ICRP (2007), The 2007 Recommendations of the International Commission on Radiological Protection,
ICRP Publication 103, Ann. ICRP 37 (2-4), Elsevier, https://journals.sagepub.com/toc/anib/37/2-4
NEA (2019), Challenges in Nuclear and Radiological Legacy Site Management: Towards a Common
Regulatory Framework, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_40359
NEA (2018), Preparing for Decommissioning During Operation and After Final Shutdown, OECD
Publishing, Paris, www.oecd-nea.org/jcms/pl_15058
NEA (2017), Radiological Characterisation from a Waste and Materials End-state Perspective:
Practices and experience, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_15056
NEA (2016), Management of Radioactive Waste after a Nuclear Power Plant Accident, OECD Publishing,Paris, www.oecd-nea.org/jcms/pl_15008
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 37

NEA (2013), “Radiological Characterisation for Decommissioning of Nuclear Installations: Final Reportof the Task Group on Radiological Characterisation and Decommissioning (RCD) of the Working
Party on Decommissioning and Dismantling (WPDD)”, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_19330
NEA (2012), Workshop on Radiological Characterisation for Decommissioning, Nyköping, Sweden,
www.oecd-nea.org/jcms/pl_49118/workshop-on-radiological-characterisation-for-decommissioning.
US EPA (2006), Guidance of Systematic Planning Using the Data Quality Objectives Process: EPA
QA/G-4,www.epa.gov/sites/production/files/documents/guidance_systematic_planning_dqo_process.pdf
US EPA (2001), Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),
www.epa.gov/radiation/download-marssim-manual-and-resources, accessed 3 February 2020.
US EPA (2009), Multi-Agency Radiation Survey and Assessment of Materials and Equipment(MARSAME), www.epa.gov/radiation/marsame-manual-and-resources, accessed 3 February 2020.
US EPA (2000), Guidance for Data Quality Assessment, www.epa.gov/quality/guidance-data-qualityassessment, accessed 3 February 2020
US EPA (2004), Manual and Supporting Documents (MARLAP), www.epa.gov/radiation/marlapmanual-and-supporting-documents, accessed 3 February 2020.

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Chapter 3. Case Studies
3.1. France
Dismantling programmes need to be executed for facilities having reached the end of theirservice life. This encompasses all activities carried out after final shutdown to reach apredefined end-state. Other than clean-up and dismantling operations, such programmes
also cover the retrieval and conditioning of legacy waste, including a range of crossfunctional activities across sites to support or coordinate projects. These clean-up anddismantling programmes are supported by research and development (R&D) activities toreduce the cost and duration of dismantling, the doses incurred, and the amount of waste
produced, while improving the working conditions on-site.
There are currently around 30 regulated nuclear facilities or individual facilities undergoingdismantling at the Nuclear Energy Division of Atomic Energy Commission (CEA DEN).
The broad range of facilities 1 to be dismantled means that the CEA is unable to use astandardised approach to characterisation. It is therefore important to clearly define the key
steps allowing the CEA to develop a decommissioning and dismantling (D&D) scenario
considering the management of waste produced by these operations. Over the years, the
CEA has gained significant experience both in project ownership and in the methodologies
and expertise required to accomplish such dismantling projects.
To illustrate the issues involved in characterising and managing legacy waste, the following
sections describe the approach developed by the CEA for the D&D of a former effluent
treatment facility. This approach can be divided into two parts. Firstly, a background studyof the facility (life cycle, incidents, etc.) was undertaken to provide information to define
the waste sampling plan. This included estimating the volumes and mass of contaminatedmaterials, its physical form, structure, geometry, materials. Secondly, a combination ofmodel calculations, direct measurements and destructive analysis were used to obtain aprecise characterisation of the facility.

Situation/objectives
This section describes the strategy chosen to characterise the effluent treatment facility at
the CEA Cadarache Centre, operated between 1962 and 2015. This facility received alphaand beta/gamma effluents with high radioactive levels or high salt concentrations. These
effluents were decontaminated by evaporation or by filtration after precipitation. In thisfacility, there are 130 tanks (from a few litres to 1 000 m3) connected to the1.

Test reactors, laboratories, fuel cycle workshops and plants, waste treatment and storage
facilities, etc. with different radioactive levels (low- to high-level waste) and radionuclide
spectra including long-lived radionuclides.

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decontamination processes, distributed throughout 8 buildings and 184 rooms (9 000 m²).
Preliminary investigations have shown that all the tanks contain radioactive waste residue.
This waste must be characterised by chemical and radiochemical analysis to assess thesource term, chemical composition of the residue (hazardous non-radioactive properties)
and to ensure that it is compatible with the D&D process (water content, density etc.).
It was not possible to investigate the contents of all the tanks due to the cost, time
constraints and inaccessibility of some of the tanks. The facility has gone through various
changes during its operational life. The characteristics of the effluents and the processes
used to treat them evolved over time with some parts of these facilities being modified
while others became redundant. The facility is currently shut down and must now becleaned and decommissioned. The facility equipment and processes can no longer be used
for either sampling operations or D&D. Operational information suggests there has been
little or no rinsing of the tanks so the majority of them are determined to be empty up to
the pump priming level.
An optimised sampling plan was devised based on the history/life cycle of the nuclear
facility. The preliminary phase of the investigation to underpin the sampling plan consistsof:•intrusive investigations including camera surveillance, residue thickness and dose
rate measurements, etc.;•

non-destructive measurements including dose rate measurements, sampling andgamma spectrometry measurements of the residue samples performed on-site;•

destructive measurements to determine activity levels and identification of specificradionuclides using a radiochemical laboratory.

Investigations have shown that all the tanks contain radioactive waste (residue). This waste,
a mixture of mud/sludge, was generated over more than 40 years of operation. It may havebeen a slow and continuous deposition process or a discontinuous and fast precipitation
process depending on the nature or the chemical composition of effluents to be treated. The
waste produced by the facility was intermediate-level waste (ILW) mainly on the alpha
spectrum with dose rates measured in mSv/h. The processes and the different routes areshown in Figure 3.1.

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Figure 3.1. Former effluent treatment facility configuration according to the processes

The main goals of waste characterisation are to determine the activity and specific activityof radionuclides (alpha-, beta- and gamma emitters) and hazardous substances (toxic
elements), and define the waste behaviour during processing, transportation, storage anddisposal. This information is used to define all the necessary D&D operations. This
includes optimising the waste processing strategy to determine the appropriate treatment,
conditioning and waste package requirements as well as the safety analysis of thedecommissioning operations. This will be achieved by carrying out an impact assessment
in normal or accident conditions and radiological dose calculations.
Characterisation objectives must be clearly identified before the characterisation
programme can be defined and executed (on-site measurements, sampling, laboratory
measurements and specific studies are required to understand the relevant physical andmaterial characteristics of the waste, etc.)
The development of a suitable waste characterisation programme is based on:•

analysis of the nuclear process;•background information on the nuclear specificities of the facility (references);•

preliminary measurements (dose rate, videos, on-site gamma spectrometry).

This information makes it possible to develop a waste sampling plan so that the samples
are as representative as possible and to identify the required analysis in compliance withthe waste acceptance criteria (WAC). The analysis considers the maximum acceptability
limit (LMA) for each radionuclide thereby defining the radionuclides with the highest
impact. Estimate the total activity, the list of nuclides (highlighting the key ones for
decision-making) and determine the list of radionuclides and scaling factors. Figure 3.2
provides a summary of the theoretical and practical approaches to characterisation.

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This analysis also needs to consider the schedule and cost factors (sampling operations,transportation requirements, laboratory analysis).
Figure 3.2. Theoretical and practical approach to waste characterisation

Regulatory framework and waste acceptance criteria
In France, the WAC fall under the responsibility of the National Agency for Radioactive
Waste Management (Andra). Andra defines the waste categories and related requirements
which are summarised in Table 3.1. For instance, they have defined two criteria: the total
activity of a waste package, and the LMA expressed in Bq/g of waste for each isotope. The
objective of the LMA criterion is to maintain an acceptable radiological impact with respect
to airborne transfers in the case of an accident during operation or post-closure phase.

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Table 3.1. Waste acceptance criteria
Category
Very low-level waste(VLLW)

Low-level waste (LLW)

Very short-lived waste
VSLW

Management throughradioactive decay

Short-lived waste
VLLW

Surface disposal (industrial facility for grouping,storage and disposal)

LLW-LL

LILW-SL

Surface disposal (Aubeand Manche disposal
facilities)

Intermediate-level waste(ILW)
High-level waste (HLW)

Long-lived waste

LLW-LLNot applicable
HLW

Near-surface disposal
under development

Deep geologicaldisposal
underdevelopment(Cigéo project)

Source: Andra, 2018.

Identifying the characterisation objectives
The first challenge for the team was to define the characterisation objectives based on theinformation that was available. This is important for both the operator who is defining the
D&D method which needs to be underpinned by the sampling, monitoring and analyticalprogramme as well as the regulator who needs to have confidence that the safety andenvironmental requirements are being met.
The team chose an optimised approach based on the history/ life cycle of the nuclear facility
to develop a waste sampling plan. The analysis plan was prepared which comprised of thelists of nuclides, uncertainty levels and detection level requirements.
The laboratory-based characterisation programme was divided into two phases. It wasnecessary to carefully analyse the results of the first phase to check the quality of the dataobtained by non-destructive gamma measurement before progressing to the second phase.
Once the first phase was deemed acceptable, the key radionuclides were identified so thatthe measurement details could then be defined for the second phase.

Identifying the characterisation method
There are many ways of characterising radioactive waste including:•calculation using neutronic activation codes or chemical codes;•

non-destructive measurements;•

destructive measurements following sampling.
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A combination of methods provides the most efficient and effective approach. In this case,the last two methods were used. The gamma spectrometry was used to determine easy-tomeasure gamma emitters for intermediate to high activity levels including 60Co, 137Cs and241
Am. These were then used to apply a scaling factor. This technique did need calibrationand geometry factor corrections (by computer codes) for quantitative purposes. This
technique, however, is not suitable for high dose rate, high volume or high mass waste.
All matrices can be characterised by laboratories for alpha, gamma and beta emitters usingchemical composition measurements. However, the analytical performance and the abilityof the laboratory to carry out a representative assessment of the sample depends on thesample composition i.e. hard-to-dissolve and/or heterogeneity of the samples a well as theactivity levels i.e. are gloveboxes needed, can present problems. The radiochemical
methods were evaluated for the measurements of long-lived and difficult-to-measure(DTM) radionuclides before the analytical methods were selected. Once all this information
has been gathered it was possible to develop a waste sampling plan, ensuring that the
samples were as representative as possible and to identify the required analysis incompliance with the WAC.

Identifying the tanks to be investigated
The facility operational process was reviewed, and the tanks were divided into different
categories depending on their location in the process. To optimise the number of tanks thatneeded to be investigated, reference tanks were chosen both downstream and upstream ofthe processes to get an idea of the composition of the residue/mud. Most of the results
demonstrated that the process was homogenous with only a small number of samples
identifying that part of the process was heterogeneous. This led us to identify sub-categories
that will require further investigation. This iterative approach enabled characterisation of
the whole facility.
The location of the tanks was not the only method used to determine the tank category.
Figure 3.3 provides a schematic representation of the different types of tanks that wereidentified. The following criteria were also used including:•technical feasibility of taking a sample from the tank and radiological dose
information;•

history of the tank and events which could call into question the representativenessof the tank within its category;•geometric representativeness of the tank compared to other tanks within thatcategory including the dose rate measured outside the tank. (similar geometries
coupled with similar dose rate measurements were required to have confidence toextrapolate sample results);•

costs and schedule.

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Figure 3.3. Identification of the different types of tanks

Sampling operations and non-destructive measurements on-site
The aim of the investigations of each tank was to ensure that sampling was exhaustive andrepresentative. The waste measurement, monitoring and sampling methodology included:•

Camera surveillance inspection.

•

Surface mapping of the residue (dose rate measurements).

•

Identification of the different areas to be investigated (physical state, visual aspect,dose rate, thickness variations).

•

Thickness measurement of the residue.

•

Three samples per area to identify possible variations or heterogeneity within anarea. The sampling tools were decontaminated between two samples. Each sample
location was precisely recorded. When the residue thickness was more than 20 cm
in the studied area, three other core samples were taken per 20 cm of residue.

The samples were collected using poles and specific tools adapted to the type of the residue(cups, syringes). All the samples were conditioned in appropriate storage bottles suitable
for shipping and direct gamma measurement (allowing for the proper modelling of themeasured residue). This prevented any deterioration in the sample. The sampling volumereflected the residue type in accordance with the analytical laboratory or transportation
constraints. A local storage area was also created in the facility to store duplicate samples.
The first phase of the analysis of the samples was conducted on-site by gamma
spectrometry which provided instant information on the activity level of key nuclides (60Co,
137
Cs, 241Am). These measurements and their physical composition allowed us to identify

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the samples, which were sent to the laboratories for radiochemical and chemical analysis.
In addition, crucial information on sample homogeneity was also obtained for each tank.

Optimisation of destructive measurements on sampled materials
As part of the first phase, the physical description including the volume, weight, pictureand dose rate were recorded. Once this information was documented the samples were
homogenised, prepared and analysed.
•

Liquid samples were filtered (at 2 or 0.45 µm) to determine the quantity ofsuspended solids (SS).

•A pH measurement was performed before and after filtration.

•

The bulk density of humid sludge samples was measured and a dry extract wasprovided. The analytical protocol was then adapted according to the sample watercontent, i.e. when the water content >50%, filtration was followed by separated
liquid/solid analysis. The solid part was dried to a constant weight. When the watercontent was < 50%, analysis was performed on the dried sample only.

•

Measurement of gamma emitters.

•

Measurement of alpha emitters.

The second phase of the characterisation programme was required to confirm the on-site
gamma non-destructive measurements as well as identifying more radionuclides with abetter detection limit and a lower uncertainty level. At least 24 gamma emitters and alphaemitters could not be directly measured through gamma spectrometry.
The second phase of the analysis was carried out on selected solid samples (according to
the results of first phase) and on all liquid samples. The following analyses were performed:•

U and Pu isotopic composition by mass spectrometry (this was carried out as themain radionuclides were alpha emitters, which are relevant with respect to waste
packaging and disposal).

•

Beta emitter for 90Sr, 63Ni, 3H and 14C.

•

Basic chemical composition.

•

Cl, F-, SO42- composition by ionic chromatography. The samples were strippedusing acid or basic water. This strategy was defined based on the WAC except in
the case of short-lived emitters and the total alpha activity (the alpha activity ofeach waste package after 300 years must be lower than 3 700 Bq/g).

The methodology is illustrated in the waste sampling plan in Figure 3.4.
It was important that the results provided a high level of confidence to proceed with thenext phase of D&D. For this reason, it was determined that 30 samples needed to be takenfrom 7 tanks for radiochemical and chemical analysis. Three laboratories were used toperform the analysis due to the high number of samples and because different
measurements and methods needed to be used.

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Analytical techniques
Several analytical techniques were used to determine the radionuclide content of the residue
from the tanks.
Figure 3.4. Sample plan

The results led to the identification of the following points:
1. The radiological source term was mainly located in the solid residue (with asignificantly higher specific activity than in liquid residue). The level of variation
reaches a maximum at around 2% for beta/gamma emitters and below 0.6% fortotal alpha emitters. This is illustrated in Figure 3.5.

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Figure 3.5. Comparison of alpha or beta gamma emitters – liquid/ solid ratio

2. The residue in each tank is homogeneous in relation to specific activity andradiological spectrum. The overall distribution in each sample analysed is generally
lower than the uncertainty obtained on the measurement. This can be explained as
the operator had organised a homogenisation campaign as illustrated in Figure 3.6.
In each tank, the 241Am or 137Cs activity levels measured on the samples have a
confidence interval lower than 50%. This low relative distribution indicates that thetank is homogeneous in terms of the radiological spectrum.
Figure 3.6. Comparing the measured 137Cs activity to a normal law

3. The radiological spectra are homogeneous within each former treatment process.
All the measured 241Am/137Cs ratios have been plotted as shown in Figure 3.7. The
results show two different distributions. Each distribution centred on the averagevalue with a standard deviation lower than 50% can be attributed to the treatment
process.

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Figure 3.7. 241Am/137Cs ratio in all samples

Figure 3.8. Radiological identification of the process by the 241Am/137Cs ratio

Evaporation treatment process

Filtration treatment process

Analysis of the radiological spectra obtained for the various investigated tanks allowed us
to identify three main groups of tanks. These groups are homogeneous and are consistent
with the facility processes. They present two distinct 241Am/137Cs ratios as shown in Figure
3.8. Consequently, it is possible to assign a tank to a cluster by studying its 241Am/137Cs
ratio and its location within the facility process as show in Figure 3.9.

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Figure 3.9. 241Am/137Cs ratio and associated clusters

1. The specific activity, in Bq/g, is different in each tank of a given process chain.
2. The measurement of two key nuclides (241Am and 137Cs) in a solid residue sampledfrom a given tank allows us to calculate the source term once the residue volume is
known.
3. The chemical composition of the residue will allow us to study options for wasteconditioning.

Estimating the source term
The mass of residue in each tank was calculated from the volume and the bulk density. This
last parameter proved difficult to measure because of the bias inherent in the sample. Whenthe sample is powdered or muddy, the material expansion must be considered. The values
deducted from laboratory measurements were 1.6 for dry residue, 1.3 for muddy residue
and 1 for liquid residue. The radiological source term was determined from the mass ofresidue in the studied area and from the specific activity determined by the spectrum
activity previously measured in each part of the facility. Considering the uncertainty in themass of the waste and the heterogeneous content of the tank measured, a preliminary
estimate of the source term gives an uncertainty of 50% (conservative approach).
All the tanks will be measured before conditioning and the real activity level must be
adjusted by direct gamma measurement.

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Conclusion and lessons learnt•

This section gives a brief description of the waste characterisation method used by
the CEA for its facility clean-up and dismantling programme. The method is basedon background knowledge of the site associated with radiochemical measurements.
Physical, chemical and radiochemical characterisation of the waste has beenpossible by using a combination of in situ and laboratory methods.

•

Gamma spectrometry was used to determine the easy-to-measure gamma emittersfor intermediate to high activity levels. This technique would not have been suitable
if the dose rate in the samples had been high or if they had been high volume/mass
waste.

•Developed within the framework of D&D projects the analytical strategy for theformer effluent treatment facility has allowed us to determine the source term of
the waste. Characterisation of seven carefully selected tanks (5% of the tanks in thefacility, i.e. 10% of the volume of residue in the facility) has provided information(source term, waste volumes and radiological spectra) that allowed us to define aclean-up scenario for all 127 tanks.

References
Andra (2018), Synthesis Report: National Inventory of Radioactive Materials and Waste,
https://international.andra.fr/sites/international/files/2019-03/Andra-Synthese-2018_EN_relu_HD.pdf

Bibliography
Martz, H.F., L.L. Hamm, W.H. Reed and P.Y. Pan (1993), “Combining mechanistic best-estimate analysis
and level 1 probabilistic risk assessment”, Reliability Engineering and System Safety, Volume 39,
Issue 1, pp. 89-108, DOI: 10.1016/0951-8320(93)90151-N
NEA (2015), Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility,
Thermal-hydraulics and Technologies, 2nd Edition, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_14972

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3.2. Japan
Fukushima Daiichi Nuclear Power Plant, owned by Tokyo Electric Power Company
Holdings, Inc. (TEPCO), is located on the Pacific coast in the middle of Fukushima
Prefecture, and has six boiling water reactors. When the disaster occurred on 11 March
2011, units 1 to 3 were in rated output operation, whereas units 4 to 6 were off duty for the
regular outage inspection. The arrival of the tsunami caused flooding of the nuclear power
plant electrical systems providing cooling and emergency backup power to units 1 to 5.
The resulting station blackout caused fatal damage to the reactor cores. The upper structuresof reactor buildings 1, 3 and 4 were severely damaged due to several hydrogen explosions.
Consequently, radioactivity contained in the damaged fuel was discharged to thesurrounding environment.
In the immediate aftermath of the accident, TEPCO focused its efforts on safety managing
the radiological hazard. In subsequent years TEPCO has been concentrating its efforts on
retrieving both the spent fuel in the fuel ponds and the damaged fuel from the nuclear power
plants as well as managing contaminated water and solid waste. The major
decommissioning challenge is the management of large volumes of waste with different
characteristics compared to those found under normal operations.

Policies on solid waste management
The Decommissioning Office of the Nuclear Damage Compensation andDecommissioning Facilitation Corporation (NDF) is comprised of experts from Japan andaround the world who formulate strategies to deal with the mid- and long-term challenges
in decommissioning the nuclear power plant. The NDF produced the Technical Strategic
Plan 2017 which includes the characteristics of the solids waste based on the accident
condition, the decommissioning actions taken after the accident and the results of thecharacterisation (sampling and analysis) completed to date. Based on the characterisation
information collected so far, the plan proposes several solid waste management policies aswell as specific tasks that will be required to meet them. It also contains a statement for theprocessing/disposal of solid waste (NDF, 2017).
The Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO’s
Fukushima Daiichi Nuclear Power Station (The Inter-Ministerial Council for Contaminated
Water and Decommissioning Issues, 2017) was revised in September 2017 (InterMinisterial Council, 2017). The changes summarise the basic concept of solid waste
management integration with more detailed characterisation, processing requirements(including volume reduction) and disposal, based on the strategic proposals.
Several measures, summarised below, have been identified which detail the present efforts
as well as the research and development (R&D) requirements to meet with the radioactive
waste management policies (NDF 2017). These comprise of:
1. Containment and isolation: Solid waste is contained so that radioactive material is
not dispersed further into the surrounding environment. It is also important toprevent human exposure to harmful radiation.
2. Reduction of solid waste volume: The amount of solid waste generated bydecommissioning is reduced as much as possible to ease the burden of solid waste
management.

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3. Characterisation: Improvements need to be made on solid waste characterisation
including studies to determine the optimum number of samples required tocharacterise large volumes of waste with varying nuclide composition. This
information needs to be used in conjunction with e.g. nuclide migration models todetermine how the radioactive waste will be managed. In addition, a study will also
look at the analysis of non-radioactive substances to understand how they might
impact on the proposed processing and disposal techniques.
4. Storage: It is essential to understand the volumes and characteristics of the solid
waste that need to be managed to establish the specifications of the waste disposal
facilities and the technical requirements for waste packages. As thedecommissioning plan evolves and progress is made on characterising the waste
produced from these activities it will become clear how much solid waste needs tobe managed. The aim is for all solid radioactive waste to be stored within the
Fukushima Daiichi Nuclear Power Plant.
5. Establishment of appropriate processing methods to meet disposal requirements:
The system for defining the appropriate method of processing to enable radioactive
waste to be stabilised and immobilised needs to be established. This requires carefulconsideration as the disposal facility has yet to be determined.
6. Promotion of effective R&D: To define an efficient R&D programme, it is essentialthat all the technical fields such as waste characterisation, processing and disposal
are involved in decision-making. Issues and discussions on R&D should be shared
between parties and any conclusions need to consider the whole life cycle ofradioactive waste management.
7. Efficient implementation of R&D projects from the perspective of overall solid
waste management: To continue safe and efficient solid waste management, it isimportant to have the right infrastructure both in terms of human resources andfacilities.
8. Measures to reduce worker exposure to radiation: It is important that the health andsafety of workers is fully considered when defining radioactive waste managementtasks. This should involve the implementation of radiation exposure controls, safety
management and healthcare programmes based on the relevant laws/regulations.
The Roadmap specifies that the safety of radioactive waste management activities
including the proposed processing/disposal methods and technology should be determinedin the financial year 2021. The technical targets (which will continue to be underpinned by
future characterisation work) comprise of:•Establish a safe and rational disposal concept based on the characteristics andvolume of the solid waste generated at the nuclear power plant. Define the mostappropriate processing technology and develop the safety assessment method basedon the most likely disposal concept. Gather experience on how this has been done
from other countries.

•

Define the radiological analysis and evaluation methods for characterisation.

•

Develop the rationale behind selecting the preferred processing technology andequipment needed for the stabilisation and immobilisation of, for example,
secondary waste generated by the water treatment process.

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Set processing/disposal criteria for solid waste disposal using experience from othercountries.

•

Clarify issues and measures concerning the storage prior to conditioning the solid
waste.

The specification of the waste packages will be determined after the start of fuel debris
retrieval when more characterisation information will become available. The production of
the waste packages will begin once the disposal site requirements/criteria are understood.
It is expected that the processing facility will be installed on the site.

Storage of radioactive waste
As of 29 March 2019, approximately 460 000 m3 of waste was in temporary storage areas
designated according to surface radiation dose rate. Most of the solid waste consists of
rubble, uncut trees and personal protective equipment (PPE) etc. Among the secondary
waste generated by water treatment are about 4 300 used vessels. Improvements have been
made to the water tightness of the vessels in addition to shielding. The vessels are stored inused adsorption column storage facilities (TEPCO, 2019). Tables 3.2, 3.3 and 3.4 shows
the status of stored solid waste in March 2019. Tables 3.5, 3.6 and 3.7 shows the status ofwaste generated from the water treatment process as of April 2019.
Table 3.2. Management status of rubble
Surface radiation dose rate
(mSv/h)≤ 0.1

Storage method

Stored volume (m3) / Storage
capacity (m3)Outdoor storage

0.1~1

Outdoor sheet covered storage

1~30

Soil-covered temporary storage
facility, temporary storage tent,
outdoor containeer storage
Containers (in solid waste storge
building)
Total

195 300/252 700
77%
37 900/71 000
53%
18 500/31 700
58%

>30

15 100/46 500
33%
266 800/401 000
67%

Table 3.3. Management status of felled trees
Classification

Storage method

Root

Outdoor storage

Branch/leaves

Temporary storage pool
Total

Stored volume (m3) / Storage
capacity (m3)
96 800/134 00072%
37 300/41 600
90%
134 100/175 600
76%

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Table 3.4. Management status of used PPE
Storage method
Container

Stored volume (m3) / Storage capacity (m3)
56 000/68 300
82%

StoragemethodType of used vessels

Outdoor temporary storage areaof used vessels

Table 3.5. Management status of used vessels generated from the water treatment
plant

Caesium adsorption apparatus
Second caesium adsorption apparatus
High Integrity Containers from Multi-nuclide removal
equipment
High Integrity Containers from improved multi-nuclide removal
equipment
Used vessels from high-performance multi-nuclide removal
equipment
Used column from multi-nuclide removal equipment
Used vessels and filters from mobile-type strontium system

Storednumber
775
216
1 590
1 460

Storednumber/Capacity

4 332/6 372
68%

74
11
206

Table 3.6. Management of waste sludge
Storage place

Store volume (m3)

Sludge storage facility (indoors)

Stored volume (m3) / Storage
capacity (m3)597

597/700
85%

Table 3.7. Management of concentrated waste liquid
Storage method
Concentrated waste liquidstorage tanks (ouside)Store volume (m3)9 364Stored volume (m3) / Storage
capacity (m3)
9 330/10 300
91%

Storage Management Plan
In March 2016, TEPCO formulated the Solid Waste Storage Management Plan which
provides an estimate of the volume of waste expected to be produced over the next ten
years. It also defines the policy on the necessary radioactive waste infrastructure including
storage facilities and processing equipment. This plan is reviewed on an annual basis.
The Storage Management Plan (TEPCO, 2018) states that the volume of solid radioactive
waste, such as rubble should be reduced in volume where practicable. This will then betransported to a solid waste storage building around 2028. Another storage building will be
constructed for secondary waste generated by the water treatment plant along with large,
heavy waste such as the waste storage vessels. The proposed plan is shown in Figure 3.10.
These measures will consolidate waste storage areas and supersede many of the currently
scattered temporary storage areas shown in Figure 3.11.

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In the Storage Management Plan, the future generation of some waste has not been
calculated. Over the next ten years, as decommissioning progresses and the plan is revisedand updated it will be possible to provide estimates on the volume and types of radioactive
waste that are likely to be produced. These estimates will then be included in the annual
revision of the Storage Management Plan.

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Figure: 3.10. Schematic representation of the Storage Management Plan

Source: TEPCO, 2020.

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Figure 3.11. Present and future storage conditions of rubble and secondary waste generated
by water treatment on-site at the Fukushima Daiichi Nuclear Power Plant(a) Present storage condition of “rubble, etc.” and “secondary waste generated by water treatment”(b) Future storage condition of “rubbles, etc.” and “secondary waste generated by water treatment”
Source: TEPCO, 2018.

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Radiochemical analysis
To define radioactive waste management characterisation methodology various
contaminated materials have been sampled and analysed including “difficult-to-measure”(DTM) radionuclides for disposal. Some analysis has also been made on the chemicalcomposition and physical properties which will help to inform processing options.
However, the main area of focus remains on the radiochemical analysis.

RubbleSamples were collected around the reactors of units 1-4 and from inside them for units 1-3.
These were analysed to investigate the composition of radioactive nuclides to develop aprimitive fingerprint. In addition, samples were taken from inside the reactor buildings ofunits 1-4, the turbine building of unit 1 and the temporary storage area and weresuccessfully analysed. It was noted that the samples which had been in contact with theaccumulated water had a different radiochemical composition due to migration.
The local distribution of radionuclides on concrete samples were measured by imaging
plate method, so that radionuclides were retained on their surface. Two samples, whichwere collected from inside the primary containment vessel (PCV), were analysed. It was
found that there was contamination inside the Traversing Incore Probe (TIP) guiding pipe
(used to measure the axial and radial neutron flux distribution within the reactor core) andthe sediments that had been underneath the accumulated water. The ratios of radioactivity
to 137Cs (Bq/Bq) are shown in Figure 3.12 (IRID-JAEA, 2019a). The ratios for activation
products including 60Co were greater in the TIP sample compared with the sediment,
whereas actinides were greater in the sediments. This suggests that composition ofradionuclides is not uniform inside PCV.

Accumulated water
Previously, the contaminated water accumulated under the Centralised Radioactive Waste
Treatment Facility (Main Process Building, High Temperature Incinerator Building) was
analysed after being decontaminated using the chemical treatment system. The results ofthis revealed that radionuclide transfer to the basement water was different from that to theair. Further samples were taken and analysed from the upstream including the reactor
buildings and the turbine buildings. The ratio of 90Sr and 238Pu concentration to that of 137Cs
are shown in Figure 3.12 (IRID-JAEA, 2017). Most of the data points fall into a similarratio, as expected except for 238Pu, where the ratios in the water samples were greater than
two orders of magnitudes.

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Figure 3.12. Concentration of radionuclides as ratios to that of 137Cs, for samples collected in
PCV

Source: IRID-JAEA, 2017.

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Figure 3.13. Concentrations of 90Sr and 238Pu in the accumulated water collected fromseveral points

Source: IRID-JAEA, 2019b.

The samples taken of the accumulated water under the reactor building often contain solid
substances. Therefore, these samples were filtered and a separate analyse was performedon the filtrate and the solid. As shown in Figure 3.13 (IRID-JAEA, 2019b), alpha emitterswere concentrated in the solid sample, whereas the 137Cs was found in both the solid sample
and the filtrate.
Np is a long-lived alpha radionuclide and is potentially important in disposal safety due
to its relatively high mobility. Various samples of contaminated water were analysed for
237
Np, and the calculated transport ratio, which is the concentration ratio to 236U as thereference nuclide normalised with the ratio of radioactivity in the supposed source term, isshown in Figure 3.14 (IRID-JAEA, 2019b). The transport ratio for 238Pu, 241Am and 244Cm
237

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decreases as you move downstream of the reactor buildings suggesting that they are
selectively removed from the water. The 237Np transport ratio was approximately 10 orgreater compared to 238U as shown in Figure 3.15. This likely to be because Np is slightly
more soluble in the water forming stable NpO2+.
Figure 3.14. Balance of nuclides between the solid substance and the filtrate from theaccumulated water

Source: IRID-JAEA, 2019b.

Figure 3.15. Behaviour of Np, which is presented as transport ratio of 237Np, compared withother actinides

Source: IRID-JAEA, 2019b.
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Secondary waste generated from water decontamination
Secondary waste generated from the water treatment, is often difficult to analyse.
Previously, carbonate and iron hydroxide slurry as well as barium sulphate sludge to namea few were analysed to provide information on how to prepare the sample for analysis. To
quantify the DTM radionuclides in the solid samples, chemical flowsheets were developed
to define the most appropriate dissolution process. As an example, the flowsheet for sludge
analysis is shown in Figure 3.16. Barium sulphate was dissolved in a concentrated alkaline
solution to the separate 90Sr (IRID-JAEA, 2018).
Figure 3.16. Flowsheet for the dissolution of sludge

Source: IRID-JAEA, 2018.

The Multi-Radionuclide Removal System uses several different adsorbents. To estimate
the radioactivity contained in the adsorbents, water was analysed from the chemicaltreatment system to get an idea of the radiochemical composition.

Ash from incineration
Ash produced from the incineration of the PPE was analysed to determine how much 137Cs
and 90Sr was present. These radionuclides were selected as the probable main contaminants.
The composition was found to be like that of the rubble already analysed.

Database
The analytical results are reported in the secretariat meeting for “Team forCountermeasures for Decommissioning and Contaminated Water Treatment,” and iscollected in the database “FRAnDLi” (JAEA-CLADS, 2019). This centralised database

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contains information on all the waste and contaminated water samples taken including
radioactivity concentrations and other useful analytical data.

Regulatory framework and waste acceptance criteria
Fukushima Daiichi Nuclear Power Plant was designated as a “Specified Nuclear Facility”
by the Japan Nuclear Regulation Authority (NRA) to ensure that appropriate measures were
put in place to safely manage the facility. One of the main waste management challenges
is defining the appropriate waste processing and disposal requirement without any WAC.

Operation plan of the Radioactive Material Analysis and Research Facility
The Radioactive Material Analysis and Research Facility (RMARF) consists of an
Administration Building, Laboratory-1 and Laboratory-2. It is located within the
Fukushima Daiichi Nuclear Power Plant site and named 1F. It is expected that constructionof Laboratory-1 will be completed by the end of the 2020 financial year, and the JapaneseAtomic Energy Authority (JAEA) who is responsible for R&D will start operation fromthe 2021 financial year. Radioactive materials with a surface dose rate up to 1 Sv/h such as
rubble, incinerated ash and secondary waste from the water treatment facility will beanalysed at Laboratory-1. Laboratory-2 will be used to analyse highly radioactive waste
such as fuel debris. The detailed design of Laboratory-2 is being progressed.

Characteristics of 1F waste and waste streams
Radioactive waste generated following the accident are quite different compared to normal
operational and decommissioning waste. There are a wide range of contaminated areas, theexistence of uncharacterised high dose (sometimes novel) wastes and high volumes of low
dose radioactive wastes containing different concentrations of radionuclides and otherchemicals. This makes it extremely challenging to sample and analyse as well as to makeestimates on the amount of radioactive waste that will need to be processed, stored,
packaged and disposed of. Several R&D projects have been conducted to gather morecharacterisation information to determine how many waste streams there are and how theymight be best managed.
To develop an integrated waste management strategy of 1F waste, it is essential tounderstand the waste inventory. Currently, analytical models are being used to support theinventory estimation. One of the key objectives of waste characterisation is to increase theaccuracy of this data to improve the inventory estimation. An R&D project has been put inplace to establish the method to be applied for the determination of radioactivity using
e.g. scaling factors. This is expected to take 10 years. The first phase of the process has
included the classification of target wastes into six waste groups which can be analysed inLaboratory-1. The six waste groups are summarised in Table 3.8.

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Table 3.8. Characteristics of the 6 target waste groups
Waste stream

Specific example

Features of target
waste

Waste generated
from dismantlingbuildings

RPV, PCV andinside of building(metal andconcrete)
Rubble (metal andconcrete)

Most waste will be
generated after fueldebris retrieval.

Secondary waste
generated by water
treatment.

Spent adsorption
columns, slurry andsludge.

Flammablematerials
Cut down trees
Incineration ash

Flammablematerials
Cut down trees
Incineration ash

Small amount agenerated waste at
this time but it willcontinue toincrease. Varied
nuclidecompositions andvarious types ofadsorbent.
Large amount ofwaste generated due
to contamination bydispersed
radioactive
material.

Waste generated
from small-scale
fuel debris retrieval.

Tools

Contaminated soil

Soil

Metal and concrete
waste around/within
the building

Large amounts ofwaste contaminatedwith dispersedradionuclides.

Large amount ofwaste generated bycontamination ofdispersed
radioactive
materials(contaminated by
Cs)

Considerations on
formulating analysis
plan
Final volume ofwaste may be large.
Evaluate thesampling regions(width and depth).
The analysis will be
used to determine
what waste can bereused.
It is difficult tosample and analyse
due to the high dose
rate. New analyticaltechniques(dissolution andseparation) need to
be developed.
Hazardous highactivity waste
stream which will
be analysed inLaboratory-1.
Several samples
have already been
analysed.
Highlycontaminated, highdose rate
radioactive waste.
Generation ofhighly contaminatedsoil.

Number of existing
samples
N/A

58

12

133

N/A

13

Source: NDF, 2019.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 65

Current operation plan of Laboratory-1 and estimated analysis plan
It is essential to adequately characterise the radioactive waste generated from thedecommissioning process to consider the best rationale for storage and processing/disposal.
It is important to establish what analytical facilities will be required and ensure that thereis the right level of trained, technically capable resource to deliver it. By starting to develop
the organisational capability prior to the completion of Laboratory-1, this will ensure thatthe facilities are fit-for-purpose and that the necessary preparatory analytical work can
begin as soon as the facility is constructed. It is expected that although the laboratory willcommence start-up phase of operation during the 2021 financial year, full-scale operation
will not start until the 2023 financial year.
Start-up phase of Laboratory-1 (from 2021/22):•

Verification of sample results that have already been analysed by JAEA. The target
samples are metal and concrete waste generated around/within buildings,
flammable materials, cut down trees, incinerated ash and contaminated soil.

•

Check validity and credibility of radiological analysis data.

•

Train technical staff and prepare the operation manual.

Full-scale operation of Laboratory-1 (after the 2023 financial year):•

As a priority, waste streams generated from dismantling buildings, metal andconcrete around/within buildings, secondary waste from water treatment and fromsmall-scale fuel debris retrieval will be analysed. This is because the final volume
of waste and/or variety may be large.

•

High volumes of rubble with low radioactivity (a part of waste generated fromdismantling buildings, and metal and concrete waste generated around/withinbuildings) will potentially have high priority. This is because it is necessary tominimise the waste volume by recycling, re-use, clearance or pre-treatment.

•

The results of on-going R&D, such as partial automation of analytical process,simplification of nuclide separation process will continue to be progressed.

•

R&D will continue as required to establish new analytical techniques for novel
waste streams.

It is expected that 200 samples will be analysed per year from the 2023 financial year as
shown in Figure 3.17. The analysis plan for Laboratory-1 will be reviewed and revisedwhen required to meet with the strategic characterisation objectives.

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66 | NEA/RWM/R(2020)2
Figure 3.17. Estimation of the number of samples analysed for each waste streams over 10
years at Laboratory-1

Contaminated soil
Waste generated from small-scale fuel debris retrieval
Flammable materials, Cut down trees, Incineration ash

200

Secondary waste generated by water treatment

100

Metal and concrete waste generated around/within buildings

Waste generated from dismantling buildings
0
FY2021

FY2022

FY2023

FY2024FY2025

FY2026FY2027

FY2028FY2029FY2030Source: NDF, 2019.

Development of analytical method
It is expected that improvements will need to be made to some of the analytical techniques
for measuring long-lived and DTM radionuclides. This part of the work is both complex
and time-consuming. Several R&D projects are listed below:•

Inductively coupled plasma mass spectrometry (ICP-MS) could be used to measure
long-lived radionuclides instead of radiation measurement which requires chemical
separation and long counting times. Applying ICP-MS to the measurement of 79Se,
93
Zr, 93Mo, 99Tc will lead to a quicker analysis as well as a lower detection limit.
Triple-quadrupole ICP-MS would be advantageous to use in chemical separation
as it is possible to conduct the reaction with the gas within the device.

•

Automation of DTM radionuclides where possible utilising improved chemical
separation techniques and enhancing column chromatography. Improving thesetechniques is expected to reduce workers’ exposure.

•

A new device which can measure the distribution of radionuclides both on andbelow the surface of heterogeneous waste.

•

A new Quality Assurance system is being developed to ensure that data isrepresentative and validated so that it can be used in decision-making.

Estimation of inventory
A mathematical model has been developed to estimate radioactivity contained in the waste
including DTM radionuclides. The model is based on the estimated source term and how

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this might have been distributed over different parts of the receiving environment. To dothis the transport fraction was set, the behaviour of radioactive nuclides from past accidents
were used, supplemented with analytical data to replicate the observed behaviour in theenvironment. For the radionuclides that have not been detected by radiochemical analysis,
the transport fraction is set by considering their chemical nature with the concentration
being converted to transport ratio (IRID, 2019). As shown in Figure 3.18, elements weregrouped with the transport ratio: the frequency distribution showed a log-normal
distribution. As more analytical data becomes available the uncertainty of parameters of
the radionuclide distribution will decrease if the element grouping described is appropriate.
Figure 3.18. Grouping chemical elements by using transport ratio and an example offrequency distribution

Source: IRID, 2019.

Bayesian statistics were applied to evaluate the decrease in uncertainty of the parameters
shown in Figure 3.19. The uncertainty of log-normal distribution decreased with data
accumulation. As a result, the inventory has been expressed as probability density.

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68 | NEA/RWM/R(2020)2
Figure 3.19. Decrease of uncertainty in parameters of log-normal distribution with increased
analytical data

Source: IRID, 2019.

Conclusion and lessons learnt
After the accident, various contaminated materials and solid waste were analysed to obtainradionuclide and other chemical/physical composition data to underpin further
characterisation work. The results revealed that preliminary radionuclide composition
could be determined. However, it was found that at this early stage of characterisation, it
was important to analyse various types of waste and contaminated material to better
understand the radionuclide composition and distribution as well as the potential
categorisation. It was found due to limited knowledge about the composition of the waste
and contaminated material it was useful to apply Bayesian statistics.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 69

Challenges•

It is extremely difficult to characterise waste following an accident because themechanism of generation and the root of contamination is not fully understood andunlike most conventional operational waste, it is heterogeneous. It is recognisedthat any analysis performed on samples taken from a post-accident site should
therefore consider the stochastic property of the waste. A standardised stochastic
method should be developed that can be applied to post-accident waste. This should
also contain a method for planning the analysis based on a list of adopted reference
methods.

•

It is also important to consider the non-radiological chemical properties of the waste
as some of these might be hazardous and/or represent a problem when considering
treatment, processing, packaging and disposal options. A method of estimation is
also required based on the facility design and operational records backed up withsome analytical assessment which also requires further development.

•

It is difficult to define the volume of the radioactive waste inventory that needs tobe managed when there is no WAC available for processing, interim storage or
disposal. This is further complicated by the fact that decommissioning must
continue to be progressed.

•

It is difficult, if not at times impossible, to take representative samples ofradioactive waste when it is taken from an undetermined amounts and types ofwaste. Also, it is not possible to assume that any of the generation can be supposed
to be homogenous.

•

As there are difficulties in taking samples from high dose regions, only partial
sampling is supposed, thus homogeneous sampling and data collection is difficult.

•

All the challenges detailed make it difficult to formulate a sampling plan andanalytical plan. For example, the Data Quality Objective process is based on
frequentist statistics which then enables the formulation of a sampling plan. For thewaste of Fukushima Daiichi Nuclear Power Plant, a stochastic method of analysis
planning is being developed with the application of Bayesian statistics.

•

Currently, characterisation is aimed at providing information to support R&D workfor the development of standardised waste management technologies. In the future,
it is expected that the accumulated data and characterisation methodologies will beconsistently applied to support the process of authorisation for waste disposal.

References
Inter-Ministerial Council (2017), “Mid-and-Long-Term Roadmap towards the Decommissioning of
TEPCO’s Fukushima Daiichi Nuclear Power Station”, the Inter-Ministerial Council for
Contaminated Water and Decommissioning Issues", 26 September 2017,www.meti.go.jp/english/earthquake/nuclear/decommissioning/pdf/20170926_01a.pdf
IRID (2019), “Report for subsidy program titled Research and development on processing and disposalof solid waste”, International Research Institute for Nuclear Decommissioning, Japan, July 2019,https://irid.or.jp/_pdf/20180000_15.pdf?v=2「平成 28 年度補正予算廃炉・汚染水対策事業費補助金 固体廃棄物 の処理・処分に関する研究開発」平成 29・30 年度実施分成果報告、及び「平成 29 年度補正予算廃炉・汚染水対策事業費補助金
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70 | NEA/RWM/R(2020)2
固体廃棄物の処理・処分に関する研究開発 （先行的処理手法及び分析手法に関する研究開発）」（性状把握、保管・管理、処分、及び廃棄物ストリームに係る検討 平成 30 年度実施分成果報告

IRID-JAEA (2019a), “Analytical result of waste samples (rubble inside reactor buildings of units 1–3)”,
65th Secretariat meeting of the Team for Countermeasures against Decommissioning andContaminated Water Treatment, 25 April 2019 (in Japanese).
廃棄物試料の分析結果（１～３号機原子炉建屋内瓦礫）

IRID-JAEA (2019b), “Analytical result of waste samples (accumulated containing sludge under units
1–4, 237Np and 129I in contaminated water)”, 67th Secretariat meeting of the Team for Countermeasuresagainst Decommissioning and Contaminated Water Treatment, 27 June 2019 (in Japanese).
廃棄物試料の分析結果（１～４号機のスラッジを含有する滞留水

汚染水の 237Np 及び 129I）

IRID-JAEA (2017), “Analytical result of waste samples (accumulated water, water from treatment facility
and others)”, 44th Secretariat meeting of the Team for Countermeasures against Decommissioning andContaminated Water Treatment, 27 July 2017 (in Japanese).
廃棄物試料の分析結果（滞留水、水処理設備処理水等）

IRID-JAEA (2018), “Analytical result of waste samples (rubble, secondary waste from water treatment
system, contaminated water, treated water, soil)”, 56th Secretariat meeting of the Team forCountermeasures against Decommissioning and Contaminated Water Treatment, 26 July 2018 (inJapanese).
廃棄物試料の分析結果（瓦礫，水処理設備処理二次廃棄物，汚染水，処理水，土壌）

JAEA-CLADS (2019), Fukushima Daiichi Radwaste Analytical Data Library, FRAnDLi, https://frandlidb.jaea.go.jp/FRAnDLi/index.php?country=e, accessed 17 November 2019.
NDF (2019), “First 10-years Analysis Plan of Radioactive Material Analysis an Research Facility(Laboratory 1)”, Tokyo Electric Company Holdings, Incorporated, Japan Atomic Energy Agency,
International Research Institute for Nuclear Decommissioning, Nuclear Damage Compensation andDecommissioning Facilitation Corporation for The Third Meeting of the Expert Group on
Characterisation Methodology of Unconventional and Legacy Waste; EGCUL-3.
NDF (2017), “Technical Strategic Plan 2017 for Decommissioning of the Fukushima Daiichi Nuclear
Power Station of Tokyo Electric Power Company Holdings, Inc.: Overview”, Nuclear Damage
Compensation Decommissioning Facilitation Organisation, www.dd.ndf.go.jp/en/strategicplan/book/20171005_SP2017eOV.pdf
TEPCO (2019), “Status on management of sort of rubble, tree cut down, used protective clothing, and soon (as of 29 March 2019)”, 65th Secretariat meeting of the Team for Countermeasures against
Decommissioning and Contaminated Water Treatment, 25 April 2019 (in Japanese).
瓦礫類、伐採木、使用済保護衣等の管理状況（2019.3.29 時点）

TEPCO (2020), “Solid Waste Storage Management Plan for Fukushima Daiichi Nuclear Power Station ofTokyo Electric Power Company Holdings, Inc.”, July 2020 edition (issued in 30 July 2020).
東京電力ホールディングス(株)福島第一原子力発電所の固体廃棄物の保管管理計画（2020 年 7 月
版)

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 71

3.3. Russia
Russia is among one of the few countries that have all elements of the nuclear fuel cycle
starting with uranium mining and finishing with radioactive waste management. The soviet
atomic project, like many other countries, has left several legacy issues which now need tobe resolved. This includes developing new methods to process and store spent nuclear fueland manage radioactive waste as well as to rehabilitate contaminated areas. To resolvethese challenging issues, the government of Russia initiated several federal targeted
programmes in the field of provision of nuclear and radiation safety.
In the early stages of developing the nuclear industry, a great deal of highly varied work
was carried out simultaneously, from designing radioactive waste storage to defining urgent
arrangements for the temporary placement of radioactive waste. This simultaneousdevelopment meant that a systematic approach to radioactive management was impossible.
Also, at the time there was a lack of international experience and no defined good practice
to inform a radioactive waste management strategy. The focus of the industry was to ensure
the safety of their personnel dealing with radioactive waste and not the environmental
consequence of their decisions. This meant that temporary schemes for radioactive waste
management were put in place including storage using bulk design concepts as show in
Figure 3.20.
Figure 3.20. Example of bulk storage

Source: FSUE FEO.

The construction of near-surface repositories for low-level waste (LLW) and intermediatelevel waste (ILW) began in the mid-1960s. A number of advances in repository design havesince taken place, which have improved hermetic conditions and allowed the creation ofreliable monitoring systems. Historical near-surface repositories take the form of trenches
around 4-5 m deep, with the bottom of the trench covered with a hydro-insulating layer.
The walls are made of reinforced concrete blocks or monolithic 0.4 m thick reinforced
concrete. The top of the repository is covered with reinforced concrete slabs and a layer ofasphalt. The repository is made up of several sections. Radioactive waste was bulked in
1 m thick layers which were then plugged with cement solution prepared using liquid
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72 | NEA/RWM/R(2020)2
radioactive waste with low salinity. In the mid-1980s an additional level of around
3.5-4 metres high was built above some of the repositories, creating a two-storey
construction with a capacity of about 20 000 m3.
Up until the 1990s these repositories were considered disposal facilities and were intended
for the final disposal of LLW and ILW, with the surrounding rocks carrying out a basic
barrier role to ensure geo-ecological safety. From the end of the 1990s, following IAEA
recommendations, these near-surface type constructions were given a new status as“radioactive waste repositories with a limited period of storage.” The new designation was
based on the idea that adequate environmental protection from the hazards of radioactive
waste is dependent on the man-made barriers, i.e. durable matrices, type of packaging used,backfill between packages and the structural elements of the repository.
FSUE FEO (formerly RosRAO) was historically founded 65 years ago as a network of final
isolation sites for radioactive waste on the territory of the Soviet Union called Radon. Thoseparts of the site designated earlier than 2011 were defined as final disposal sites. The
government decided to reform the approach to handling radioactive waste followingchanges in approaches to the storage of radioactive waste, namely the transition from bulk
storage to container storage and the current trends in looking at deep geological waste
facilities. This resulted in the approval of federal law number 190 in 2011 which defined
the term “legacy” radioactive waste as that which was formed at the enterprise (site) prior
to the entry into force of this law, and that the state assumed obligations for further
management.
Some legacy waste was disposed of nearly 50 years ago without any consideration of thecompatibility of the different waste streams. For example, Figure 3.20 shows various bulk
waste temporary storage facilities. Tritium ampoules wrapped in paper and sealed inpolyethylene packs were discovered on top of ILW. This case demonstrates that it is
impossible, to make any assumption of what radioactive wastes has been disposed ofespecially when segregation was not common practice.

Situation/Objectives
Since 2011, following the changes to the federal law, 250 000 m3 of bulk radioactive waste
in Russia has been considered as retrievable. It is expected that this radioactive waste willthen be transferred to the national operator for further disposal.
Figure 3.21. General view of legacy bulk storage

Source: FSUE FEO.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 73

The focus of future work is to retrieve the radioactive waste from bulk storage, segregate
and then pack it into certified containers for transfer and disposal. To implement thisobjective, FSUE FEO arranged a pilot project for radioactive waste retrieval in 2014. This
involved several tasks including:•assessing retrievability and efficiency of solid radioactive waste (SRW) recovery;•

clarifying the composition and basic characteristics of SRW following segregation;•clarifying the processing methods once the SRW was retrieved;•

defining the technical and economic parameters.

Within the framework of the pilot project, a site for the retrieval and sorting of bulk SRW
was established directly on the floor slabs of canyons above the radioactive waste as shownin Figure 3.22. The retrieval and sorting equipment comprised of:•a hydraulic manipulator;•

a tray for sorting solid radioactive LLW using the local ventilation system;•a crane with a loading capacity of 5 tonnes.
Figure 3.22. Work area

Source: FSUE FEO.

In the process of SRW retrieval with the crane, it was necessary to evaluate the equivalentdose rate (EDR) value. If the SRW had an EDR less than 30 μSv / h, sorting andcharacterisation was carried out manually on a mobile pallet (with partial fragmentation of
the SRW). If the EDR was more than 30 μSv / h, the SRW was placed directly in thecontainer without sorting, followed by characterisation of the filled container as a wholeunit. To localise possible radioactive contamination and restrict access, the sites for SRW
retrieval and sorting were retrofitted with temporary fences based on lightweight building
structures with roll-up hangar lifting gates.
Prior to the work, a survey of the SRW was carried out. The following preliminary
characteristics of the SRW are shown in Figure 3.23.

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74 | NEA/RWM/R(2020)2
Figure 3.23. Preliminary characteristics of the FSUE FEO (formerly RosRAO) solidradioactive waste final isolation sites

3%
Construction waste

15%

17%

Metal radioactive
waste
Other

65%

Combustible waste

The specific activity of SRW was up to 3.7 x 106 kBq / kg (for beta and gamma). The LLW
in the canyon represented around 9% and ILW 10%.
As a result of the work performed, 120.7 m3 (68 040 kg) of SRW was retrieved (seesummary in Table 3.9).
Table 3.9. Composition of the solid radioactive waste retrieved from the canyon

SRW type

Volume(m³)

Mass(tonnes)

SRW
class

Averageactivity(kBq/kg)

Dominant
nuclideContainer
type

Container
quantity

Rubble

67.8

36.4

LLW

58

Cs, 60Co,
226
Ra
137
Cs, 60Co

Metal
(3m³)
Metal
(3m³)
Metal
(3m³)
Metal box
Concrete(1.8m³)

24

Metal

20.1

13.61

LLW

22

Metal

14.7

12.42

7

137

Combustible
Metal

14.6
3.5

2.81
2.8

Industrial
waste
LLW
LLW

11
4 140

137

137

Cs, 60Co
Cs, 60Co
Cs, 60Co

137

7
5
10
2

From the information in Table 3.9, it can be concluded that:•

The average specific activity of SRW was four times less than expected.

•

The estimated physical and chemical composition of the SRW corresponds to thesample that was previously retrieved.

•

By volume, the medium-level SRW (4 140 kBq/kg) was 3% (3 times less than the
expected) and the low-level category SRW (between 11-58 kBq/kg) was 97%.
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 75

As a result of radiation monitoring of retrieved SRW, the following maximum values wereidentified:
1. Measured specific activity of gamma emitters was beta 350 kBq / kg, alpha 3 kBq/ kg.
2. Contamination monitoring was alpha – 5 CPM and beta 10 000 CPM.
3. Dose rate: 2 mSv / h.
4. Individual dose of personnel for a month of work was 0.32 mSv.
5. Contamination of equipment, premises and personal protective equipment did notexceed permissible values.
The following technical and economic characteristics were determined. The overall
productivity was 1.5 m3/ day based on a two-shift operating mode. Each shift comprised of
two radioactive waste processors, a leading specialist in radioactive waste management anda dosimetrist. The average shift capacity of SRW retrieval was 6 m3/day. The optimal
number of personnel involved in one retrieval shift comprised of three radioactive waste
processors, a leading specialist in radioactive waste management and a dosimetrist.
Table 3.10 summarises the number of hours each position in the team worked, Table 3.11
and Table 3.12 provides a summary of the equipment required to complete the SRW
project.
Table 3.10. The number of hours worked by each position in the teamProject team
Chief specialist for radioactive waste
management
Leading specialist in radioactive waste
management
Dosimetrist
Radioactive waste processor
Deactivator
Radiophysicist Engineer
Locksmith

Hours
12
625
760
1 000
80
50
15

Table 3.11. Equipment required for the solid radioactive waste retrieval

Operation
Retrieval

Ventilation

Equipment
Palfinger manipulator craneForklift “Squid”
Angle grinder
Radiation monitoring devices
TOTAL
B-1 (10 kW)
B-2 (5.5 kW)
B-3 (10 kW)
B-4 (5.5 kW)
TOTAL

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

Hours
750
120
30
760
1 660
816
816
816
638
3 086

76 | NEA/RWM/R(2020)2
Table 3.12. Expendable material used during and solid radioactive waste retrieval

Expended materials
Metal container (3 m³)
Reinforced concrete container (1.36 m³)
Respirators
Cotton gloves
Nitrile coated gloves
Plastic sleeves
Plastic aprons
Plastic bags (30 l)
Overalls
Plastic show covers
NitrogenRags

Quantity used
362
276
70 pairs
100 pairs
3 pair
4200
10
10 pairs
Volume500 litres
Mass
30 kg

Bulk RW characterisation
The following tests and processes were arranged to verify that the SRW complied with the
WAC (summarised in Figure 3.24):•

Specific activity of gamma-emitting radionuclides.

•

Specific activity of strontium, plutonium, thorium, uranium (where there was anopportunity to take a representative sample).

•

Some of the SRW needed to be segregated.

•

Some SRW was pre-treated by drying and incineration.

•

The packages were visually inspected, then the EDR was measured and recorded
along with result from alpha/beta contamination monitoring.

•

The mass and dimensions of the waste were checked.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 77

Figure 3.24. Bulk solid radioactive waste characterisation flowchart

Source: FSUE “federal environmental operator” (FSUE FEO), 2020.
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78 | NEA/RWM/R(2020)2
The area used for gamma spectrometry monitoring of the retrieved SRW had too higher
background radiation level to categorise some of the waste as industrial. This would havegiven the project significant cost benefit as industrial waste accounted for 12% of the total
solid waste that was retrieved. The cost of the metal containers used for SRW disposal are
significant, so this kind of cost saving would have been extremely beneficial.

Radioactive waste framework and waste acceptance criteria
The classification of solid waste is divided into four categories: VLLW, LLW, ILW andHLW as shown in Table 3.13.
Table 3.13. Classification of solid radioactive waste

Waste category

Tritium

VLLW
LLW
ILW
HLW

<107
From 107 to 108
From 108 to 1011
>1011

Specific activity (kBq/kg)
Beta and Gamma
Alpha emittersemitters(excluding(excludingtransuranic
tritium)
elements)
3
< 10
< 102
3
4
From 10 to 10
From 102 to 103
From 104 to 107
From 103 to 106
>107
>106

Transuranicelements<101
From 101 to 102
From 102 to 105>105

Source: Adapted from Table B1, Moscow, 2017.

If the radiological characteristics of the waste are uncertain then the waste must beclassified to the highest category (Moscow, 2017). The following rule was applied for thebulk SRW with unknown radionuclide composition:- > 100 kBq/kg – for beta emitters- > 10 kBq/kg – for alpha emitters (except transuranic elements)- > 1.0 kBq/kg – for transuranic elements

Challenges•

The primary packaging (plastic bags) around the SRW had perished and as a result
it was difficult to extract the clamshell grapple which led to additional dust
generation. This consequently resulted in the excessive consumption of resourcesin the dust suppression system.

•

Due to the heterogeneity nature of the SRW it was difficult to characterise theradionuclides once it was placed in the metal containers.

Conclusion and lessons learnt•

A large amount of the retrieved metal waste was too big to fit into the standard
disposal containers, so it had to be reduced in size.

•

The area used to house the gamma spectrometer had too high a background level
to enable characterisation of waste as industrial. It would have resulted in asignificant project saving as 12% of the waste retrieved could have been categorised
as industrial.
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 79

References
FSUE “federal environmental operator” (FSUE FEO) (2020), “Bulk Solid Radioactive Waste
characterisation flowchart”, FSUE.
Moscow (2017), The Fifth National Report of the Russian Federation on Compliance with the Obligationsof the Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste
Management, State Atomic Energy Corporation Rosatom and the Federal Environmental, Industrial
and Nuclear Supervision Service with the involvement of IBRAE RAN and FSFI SEC NRS, Moscow,
www.rosatom.ru/upload/iblock/53d/53dc4676fb14b71f3e32ac651e64e3ff.pdf

Bibliography
Martz, H.F., L.L. Hamm, W.H. Reed and P.Y. Pan (1993), “Combining mechanistic best-estimate analysis
and level 1 probabilistic risk assessment”, Reliability Engineering and System Safety, Volume 39,
Issue 1, pp. 89-108, DOI: .10.1016/0951-8320(93)90151-N.
NEA (2015), Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility,
Thermal-hydraulics and Technologies, 2nd Edition, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_14972

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80 | NEA/RWM/R(2020)2

3.4. Ukraine
Ukraine has lot of experience of dealing with UL radioactive waste following the accident
at Chernobyl Nuclear Power Plant. This radioactive waste contains varying compositionsof radionuclides as well as other non-radioactive hazardous and non-hazardous waste. Mostof this radioactive waste was buried in trenches scattered across the exclusion zone
following the aftermath of this accident. There are no records documenting thecharacteristics of the waste that was buried in these trenches which creates problems for
retrieval, analysis of the radiological and non-radiological waste composition, processing,potential treatment, packaging and disposal (NEA, 2019). In addition, there is someexperience at Rivne (sometime referred to as Rovno) Nuclear Power Plant of characterisinglegacy waste from the liquid waste encapsulation plant. The following studies will be
reviewed:•characterisation of legacy waste produced at Rivne Nuclear Power Plant;•characterisation of radioactive waste and material from Chernobyl Nuclear Power
Plant;•

characterisation of radioactive waste and material at the Shelter Object (SO);•safety assessment of “Podlesnyi” radioactive waste storage facility including thecharacterisation of radioactive waste and material at module A;•safety assessment of “radioactive waste temporary localisation point” within
Chernobyl Exclusion Zone.

3.4.1. Characterisation of legacy waste produced at Rivne Nuclear Power Plant
Rivne Nuclear Power Plant (sometimes referred to as Rovno) is situated in the north westof Ukraine near the Styr river. In 1995, the saline bitumen compound (SBC) was
commissioned to condition liquid radioactive waste produced by the nuclear power plantup until 31 December 2002 when it was shut down in view of its fire hazard (Rovno Nuclear
Power Plant,Varash, 2016). This was the only nuclear power plant in Ukraine to use abitumen matrix to condition radioactive waste.
The initial product for immobilisation was liquid radioactive waste (LRW) vat residue frompower units 1 and 2 (WWER-440) and power unit 3 (WWER-1000). The SBC was formedin the process of LRW (vat residue) immobilisation into bitumen matrix at thebituminisation plant. The rotary film bituminisation plant had the capacity to produce
0.15 m3/h. Two Bitumen grades BND-60/90 and BND-90/130 were used during
immobilisation.
Records from the operator indicate that the radioactivity levels of the vat residue comingfrom the process changed from 5.5x106 Bq/kg to 7.0x107 Bq/kg. Also, the contribution toit of individual gamma-ray emitters (134Cs, 137Cs, 60Co) changed. The total amount ofprocessed (immobilised) waste was 147.8 m3 which amounts to 739 (200 litre) drums.
During the processing around 2.0 m3 of SBC was categorised as HLW. When theradioactive waste was immobilised, it was classified as ILW based on the criteria appliedat the time.
Drums of processed SBC radioactive waste were planned to be placed into 3 m3 concrete
containers followed by immobilisation with a cement matrix (the first localising barrier)and disposed of at the Specially Equipped Near-Surface Solid Radioactive Waste Storage
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 81

Facility (SENSRWSF) as shown in Figure 3.25. The SENSRWSF was part of production
complex “Vektor” of the Central Enterprise for the Management of Radioactive Waste(CEMRW).
Figure 3.25. The Specially Equipped Near-Surface Solid Radioactive Waste Storage Facility

Source: Kireev, 2019.

The SBC radioactive waste is now considered as legacy waste because there is a lack ofcharacterisation information related to the radionuclide composition, the filling degree of
the SBC containers and the waste mass. Also, there is uncertainty of the physical andchemical nature of the vat residue immobilised into the bitumen matrix as well as thepotential heterogeneity of the waste. The operator records indicate that there was directcontact between the nuclear fuel and the coolant due to the loss of integrity of the reactor
fuel elements of power units 1 and 2 so called “emergency component”.

Characterisation of the waste (according to regulatory, technical and contract
documentation)
The plan is that after characterisation and final SBC packaging the entire waste stream will
be disposed of at the SENSRWSF shown in Figure 3.24. To establish that the SBC wastecomplies with the WAC for disposal, a set of regulatory and technical documents wereapplied including:•recommendations for establishing waste acceptance criteria for disposal of
conditioned radioactive waste (SNRIU, 2004);•

guideline for characterisation, accounting and control of radioactive waste (INSC,
2017);•

short-existing, low and intermediate level nuclear power plant waste. Requirements
for the end conditioned product (NNEGC, 2013);•

nuclide vector to characterise SD Rovno Nuclear Power Plant saline-bitumen
compound (SE NNEGC “Energoatom”, 2019).

These documents are based on the regulation that requires the radioactive waste to be
immobilised in a cement matrix.
To ensure the safe disposal of the immobilised liquid radioactive waste, based on the resultsof the initial SBC characterisation the state enterprise national nuclear energy generatingcompany SE NNEGC “Energoatom”, decided to develop and agree new WAC. This will
include a final characterisation review to ensure that the SBC waste met the new physicaland chemical criteria defined in the WAC (NNEGC, 2013).

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Procedure for waste characterisation
The initial sampling and characterisation of the SBC radioactive waste led to thedevelopment of the following documents:•

methodology for sampling the salt and bitumen compound for characterisation ofalpha, beta, gamma-emitting nuclides content for nuclide vectors calculation;•

methodology for characterisation of salt and bitumen compound samples.

These documents were used to inform the subsequent sampling and characterisation
procedures.

Sampling of SBC
Sampling of Rovno Nuclear Power Plant SBC was carried out according to the document“Methodology for Rovno Nuclear Power Plant salt-bitumen compound sampling (Rovno
Nuclear Power Plant SBC) for characterisation of the alpha-, beta-, gamma-emitting
nuclides content for the purpose of radionuclide vectors calculating”, that determined theapproaches to the formation of representative containers batches and individual
representative samples. All stages of the sampling procedure and the technical aspects of
its implementation were based on published technical papers methods and statisticalanalysis (Kadenko and Pluiko, 2003) and bias and uncertainty of measurements (Pokhodun,
2006).
Sampling is an integral part of the characterisation process. The correct choice ofapproaches and sampling methods to a large extent determines the reliability and accuracyof the measurement results and whether it is possible to scale them over the entire SBC
waste stream.
The tasks performed during sampling comprised of:•Preparation of representative samples, the characterisation of which can be scaled
to the entire SBC waste stream.

•

Justification of the minimum required sample mass for characterisation according
to specified parameters.

•

Selection of the sampling procedures including the retrieval equipment andcontainer specification were established based on the properties of the bitumen
matrix. This included determining the EDR from the container.

The main steps to derive representative samples comprised of:
1. Dividing the SBC radioactive waste stream into homogeneous batches based on
two criteria: data and time of the final waste form production (time of container
filling) and information on the activity of containers obtained from the operations
register.
2. Three representative groups of drums were obtained: 1995-1996 waste productgroup (batch) with typical averaged activity values, 2001-2002 waste product groupwith typical averaged activity values and lastly a group of SBC with maximum
activity values in the stream, accumulated during the entire bituminiser working
period.

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3. The minimum required number of containers that needed to be characterised in thefirst two batches was calculated using statistics based on the accepted confidence
probability. This resulted in the selection of 15 drums. To select which containers
to characterise a random selection (a non-repeating random number generator) wasused.
4. To take into account the possible spatial heterogeneity of the radionuclides
distribution in the container, the sampling method proposed to take samples along
the full height of the container and to take two samples from containers of the third
batch (with maximum activity values). A screw drill was used as the sampling toolconsidering the physical and mechanical properties of the bitumen matrix as shownin Figure 3.26.
Figure 3.26. Screw drill and SBC samples

Source: SSE Ecocentre, 2018.

In total, 40 SBC waste samples were collected: 15 samples with average statistical valuesof 137Cs activity from each of two batches (formed on the basis of the time criterion) and10 samples with maximum values of 137Cs activity in the entire radioactive waste stream.
The minimum mass of SBC sample required was determined considering the requirementsof the applied techniques for alpha-, beta-, gamma-emitting radionuclides analysis, as wellas based on data on SBC activity (according to the operational register). The minimum
mass was determined to be 100 g.

Analysis of the waste samples
The alpha-, beta- and gamma-emitting radionuclides that were measured were defined in
the WAC of the disposal site (SNRIU, 2004). Each of the SBC waste samples were
homogenised (in masses of 2 to 15 g) then analysed using gamma spectrometry and divided
into aliquots to determine specific groups of radionuclides. The preparation of analyticalsamples for measurement was carried out using the following techniques including
decomposition of samples by wet ashing, dry ashing and acid decomposition. Thesemethods are detailed in Annex A.

Analytical techniques used to measure the specific activity of radionuclides in the
SBC samples
A variety of analytical techniques were used to derive the specific activity of theradionuclides required to be measured to meet the WAC. These methods are detailed in

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Annex A. The results of SBC waste characterisation were used for scaling factors
calculation and fingerprint (nuclide vector) determination.

Results of the characterisation of Rovno Nuclear Power Plant SBC samples
As a result of the characterisation of the samples by the content of alpha-, beta- and gammaemitting radionuclides, recommended for determining by RD 306.4.098-2004, numericalvalues of specific activity were obtained for the following isotopes: 3H, 14C, 63Ni, 60Co, 90Sr,
99
Tc, 134Cs, 137Cs, 235U, 238U, 238Pu, 239+240Pu, 241Am, 232Th. The activity of the remaining
isotopes (10Be, 59Ni, 93Zr, 94Nb, 129I, 135Cs, 236U, 241Pu, 242Pu, 243Am, 36Cl, 41Ca, 54Mn, 55Fe,
110m
Ag, 226Ra, 237Np) turned out to be lower than minimum detectable activity (MDA) of
the instruments used for its measurement.
The detection of 232Th isotope in the SBC composition is likely to be a naturally occurring
radionuclide present in the bitumen matrix. The specific activity of 232Th does not exceed
the maximum permissible specific activity approved for building materials and products by
National Standard (GOST 30108-94).
The obtained numerical values of specific activity, as well as MDA levels of radionuclides
that were not detected in the SBC composition, does not exceed the maximum permissible
specific activity values in the disposal WAC (Chernobyl, 2009).

Scaling factors calculation and fingerprint determination
The specific activity of 90Sr was derived using the linear regression model calculated by
the “weighted least squares method” as shown in Figure 3.27. The criteria of this model
selection were: comparatively high values of the correlation and determination coefficients(0.938 and 0.8798, respectively); low error values of the a, b parameters of the regression
equation (no more than 30%); the smallest error values of the calculated 90Sr specificactivity values in relation to the corresponding measured values.
The specific activity of 134Cs was derived using the linear regression of logarithm model
calculated using the “least squares method” ln [A (134Cs)] = 1.816·ln [A (137Cs)] ‒ 20.957.
The equation describing this model is identical to the power equation as shown in Figure
3.28. The criteria of this model selection were: the relatively high values of the correlationand determination coefficients (0.92507 and 0.85576, respectively); low error values of thea, b parameters of regression equation (no more than 10%); low error values of thecalculated 134Cs specific activity values in relation to the corresponding measured values.
The 137Cs isotope was chosen as the key radionuclide for the scaling factor calculation andfor determining the specific activity of 90Sr and 137Cs as they are all fission products. The
90
Sr isotope refers to difficult-to-measure radionuclides.
To determine the specific activity of the remaining radionuclides: 14C, 63N, 99Tc, 235U, 238U,
238
Pu, 239Pu, 240Pu, 241Am, a conservative approach was used. Since they are characterisedby a correlation of weak and medium strength in relation to the specific activity of key
radionuclide (137Cs is a key nuclide for fission products and 60Co is a key nuclide forcorrosion/activation products) this was deemed appropriate.
The activity of 232Th nuclide cannot correlate with the activity of any of the key
radionuclides since thorium is of naturally occurring in the bitumen matrix. The specificactivity of 232Th specific activity in the fingerprint was determined using a conservative
approach.

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A similar approach was used for 3H, since its activity values vary over a wide range (fromMDA to 2.66 x 1005 Bq/kg) and there is no relationship between 3H specific activity andthe activity values of the key radionuclides.
For radionuclides with activity values at the MDA level (10Be, 36Cl, 41Ca, 54Mn, 55Fe, 59Ni,
Zr, 94Nb, 110mAg, 129I, 135Cs, 226Ra, 236U, 237Np, 241Pu, 242Pu, 243Am) a conservative
approach was used which resulted in a threefold increase in the corresponding MDA values.
The radionuclide vector is summarised in Table 3.14.

93

Figure 3.27. Linear regression of the relationship between activity values of 90Sr and 137Cs

Source: SSE “Ecocentre”, 2019.

Figure 3.28. Linear regression logarithmic relationship between activity values of 134Cs and137
Cs

Source: SSE “Ecocentre”, 2019.
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86 | NEA/RWM/R(2020)2
Table 3.14. Radionuclide vector for saline bitumen compound samples (as of 6 December
2018)

No
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
29

Radionuclide
90
Sr
134
Cs
99
Tc
63
Ni
14
C
232
Th
238
U
238
Pu
239+240
Pu
241
Am
235
U
3
H
236
U
242
Pu
243
Am
36
Cl
41
Са
55
Fe
93
Z
129
І
241
Pu
226
Ra
54
Mn
94
Nb
110m
Ag
59
Ni
10
Be
135
Cs
237
Np

Scaling factor
A( Sr) = 0,062 A(137Cs) + 32511,770
A(134Cs) = 8E-10·A(137Cs)1,8158
5.55 x 1002 Bq/kg
3.57 x 1006 Bq/kg
1.68 x 1004 Bq/kg
2.04 x 10-01 Bq/kg
2.81 Bq/kg
1.41 x 10+01 Bq/kg
1.02 x 10+01 Bq/kg
2.17 x 10+01 Bq/kg
2.21 x 10-01 Bq/kg
7.98 x 10+05 Bq/kg
2.25 x 10-02 Bq/kg
2.25 x 10-02 Bq/kg
2.25 x 10-02 Bq/kg
3.00 x 10+01 Bq/kg
1.50 x 10+04 Bq/kg
7.50 x 10+03 Bq/kg
1.50 x 10+04 Bq/kg
3.00 x 10+01 Bq/kg
1.50 x 10+04 Bq/kg
4.50 x 10+03 Bq/kg
3.00 x 10+02 Bq/kg
3.00 x 10+02 Bq/kg
3.00 x 10+02 Bq/kg
1.50 x 10-02 Bq/kg
7.50 x 10-02 Bq/kg
3.00 x 10-02 Bq/kg
1.20 x 10+01 Bq/kg
90

Form of scaling factor
Linear equation
Power equation

Conservative approach

Source: SSE “Ecocentre”, 2019.

Challenges•

One of the most difficult tasks was to identify strategies for determining whichbatches to group together across the entire SBC waste stream. Also, determining
the required size of a representative batch, i.e. what is the minimum required
number of containers under study for proper SBC characterisation and scaling
factors calculation).

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Lessons learnt•

Significant time and resources were saved as it was possible to perform statisticaltests on the waste prior to commencing characterisation. This was possible because
detailed historical operational information was available.

•

Some of the radionuclides that needed to be analysed were outside the scope ofroutine laboratory measurements. These included DTM radionuclides such as 10Be,
41
Са, 55Fe, 59Ni, 135Cs. In this case, optimisation and testing of the laboratory
techniques was required. After completing this project, the Central Analytic
laboratory has gained considerable experience of radioactive waste
characterisation.

•

It is possible to infer the measurement of some difficult-to-measure radionuclides
by using others that are easy to measure. For example, the activity of 59Ni can bereliably determined from 63Ni.

•

To increase the reliability and reproducibility of the methods for determining
certain radionuclides the sampling protocol should contain information on sample
preparation. For example, a pyrolyzer is required to prepare samples for thedetermination of 3H and 14C.

•

When determining the fingerprint, there were uncertainties associated with the
choice of the scaling factor when applying the conservative approach. Also, there
were difficulties with determining the frequency of scaling factors verification.

3.4.2. Characterisation of radioactive waste and material from Chernobyl Nuclear
Power Plant
The industrial complex for solid radioactive waste management (ICSRM) was designed
and built based on the Chernobyl Nuclear Power Plant Decommissioning Plan developed
between 1996-1997. It was originally intended for the acceptance, processing and/or
disposal of solid radioactive wastes generated from decommissioning activities, including
radioactive waste being stored in the SRWSF constructed as part of the original design of
the Chernobyl Nuclear Power Plant. Later, the list of ICSRM tasks was expanded due tothe operational radioactive waste management of the Shelter Object (SO). In addition, thetemporary storage facility for low and intermediate-level long-lived waste (LILW-LL) andhigh-level waste (HLW) was included into list of facilities (lots) within the complex.

Current and future waste streams and their status
The main objective for radioactive waste management as defined by the strategy managers
is to characterise all the waste into the following waste categories for:•unrestricted or restricted release – according to the newly developed specificcriteria;•

temporary storage – HLW and LILW-LL;•

final disposal – low and intermediate level short-lived waste (LILW-SL).

The ICSRM and its characterisation and sorting capabilities are the main tools that theradioactive waste management infrastructure has available to achieve this objective.

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The management of unrestricted or restricted release and the final disposal of LILW-SL
are recognised as current radioactive waste management practices. However, the issues offurther managing HLW and LILW-LL stored in interim storage have been the matter of adeferred decision.
The Solid Radioactive Waste Storage Facility (SRWSF) shown in Figure 3.29 falls under
the radioactive waste management remit of the ICSRM. The SRWSF has several
compartments that were filled by means of loading without preliminary sorting andradiation monitoring between 1978-2003. Initial characterisation of this waste led to theconclusion that, except for asphalt-bitumen chipping from the SRWSF roof, the majority
is operational waste generated during the normal operation of RBMK-1000 power units.
However, as a means of mitigating the consequences of the accident at unit 4, radioactive
waste was taken to SRWSF which differs significantly from the waste generated during
normal operations. As a consequence of this the waste streams are categorised as UL waste
as they contain accident waste.
Figure 3.29. Solid Radioactive Waste Treatment Plant and Solid Radioactive Waste Storage
Facility with the system for retrieval of radioactive waste (RAW)

ICSRM

SRWSF

Source: Seyda et al., 2019.

Basic characteristics of SRWSF contain:•

The “Light” compartment (east) of the SRWSF contains additional PPE, aluminium
sheathing, pipelines, fittings, plastic compound, insulation, paper, wood, filters,
sand, workwear, swarf (non-pressed), electric motors and ion-exchange resin.

The “Light” compartment (west) of the SRWSF contains additional PPE,
aluminium sheathing, metal, reinforcement, plastic compound, insulation, building
waste, paper, wood, workwear, electric motors, ion-exchange resin and asphaltbitumen chipping from the roof.
The physical, chemical, radiological and other properties of the SRWSF waste will need to
be determined to ensure they meet with the WAC for the different stages of its treatment,
transportation, storage and final disposal. One of the main conditions for the acceptance of
SRW for storage is the conformity of the radionuclide content, regulated by the“Radioactive waste acceptance criteria for waste disposal in specially equipped nearsurface solid radioactive waste storage facility” (SENSRWSF) (Chernobyl, 2009). In
accordance with these criteria, based on the results of the radioactive waste characterisationa passport should be issued for each SRW batch, declaring the radionuclide composition,•

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specific activity of each nuclide and the total activity of the package content on the date ofwaste package transfer for disposal.
Prior to disposal of the package, it is necessary to determine the radionuclide composition
and activity of:••

any radionuclide with specific activity in the package more than 0.1 Bq/g for alphaemitters and 10 Bq/g for beta- and gamma emitters;
any radionuclide with a content that exceeds 0.1 g.

The characteristics of the waste shall meet the requirements of the WAC (Chernobyl, 2009)
established for radioactive waste packages intended for disposal. Table 3.15 provides a listof the radionuclides maximum allowable specific activity defined in the WAC and themeasurement methods.
Table 3.15. Waste acceptance criteria and measurement method for radioactive waste to bedisposed of at the specially equipped near-surface solid radioactive waste storage facility

Isotope
H

3

Be

10

C
60
Co
59Ni
63
Ni
90
Sr
93
Zr
14

Nb

94

Tc

99

129

I-

Cs
137
Cs
237
Np
235
U
236
U
238
U
238
Pu
239
Pu
240
Pu
241
Pu
242
Pu
241
Am
243
Am
135

Measurement method

MDA inBq/g
1

Liquid scintillation
Inductively coupled plasma mass spectrometry
0.5(ICP-MS)
Liquid scintillation
1
Gamma spectrometry
0.5
X-ray spectrometry
10
Liquid scintillation1
Beta radiometry or liquid scintillation
1
ICP-MS
0.1
Gamma spectrometry/ICPMS mass
0.5/7
spectrometry
ICP-MS
0.6
ICPMS mass spectrometry or X-ray
0.01
spectrometry
ICP-MS
0.1
Gamma spectrometry
0.5
Alpha spectrometry or ICP-MS
0.01/0.0001
ICP-MS
0.0001
ICP-MS
0.0001
ICP-MS
0.00001
Alpha spectrometry
0.01
ICP-MS
0.001
ICP-MS
0.001
Liquid scintillation
1
ICP-MS
0,0001
Alpha spectrometry /gamma spectrometry
0.01 / 0.5
Alpha spectrometry/ICP-MS
0.01/ 0.001

Source: ISP Nuclear Power Plant NASU Chernobyl, 2018.
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

Maximum permissible
content of Bq/g
3.77 x 10+15
2.50 x 10+06
1.30 x 10+07
5.90 x 10+07
1.90 x 10+08
2.88 x 10+05
3.85 x10+03
2.08 x 10+03
4.17 x 10+03
9.09 x 10+01
8.07 x 10+03
7.78 x 10+05
9.54
3.51
1.54 x 10+02
1.37
4.80 x 10+01
4.91 x 10+01
2.53 x 10+03
5.00 x 10+01
5.00 x 10+01
8.76 x 10+01
5.42 x 10+01

90 | NEA/RWM/R(2020)2

Sample preparation methodology
As there are several unknowns regarding the radioactive waste in the SRWSF
compartments the waste was firstly examined by measuring the total beta activity
equivalent exposure dose rate (EDR). This considered the spread in the correlations of theactivities of such key radionuclides as 60Co and 137Cs.
The optimal number of samples taken from the waste batch for radionuclide analysis
depends on the level of heterogeneity of the radionuclide composition in terms of volume
and the specific activity. The mass of the sample should be sufficient for all characterisation
requirements, considering the total number of determinants and the possibility of repeatedstudies. For samples that are relatively homogeneous by composition, it is sufficient to takefrom 5 to 6 dm3, and for substantially heterogeneous samples from 8 to 10 dm3.
It is important that radiochemical methods used to determine the specific activity ofradionuclides are reproducible. Depending on the measurement methods used MDA of thedetermined radionuclides can vary in a wide range of values – from 1x10-4 to 10 Bq/g.

Method used to determine the radionuclide vector/fingerprint
The radioactive waste is certified for disposal according to the radionuclide content
described in the WAC show in Table 3.14. However, the vast majority of these are DTM
alpha-emitting radionuclides. Figure 3.30 shows the scheme of certification used when
dealing with DTM radionuclides.
Figure 3.30. Scheme of certification of difficult-to-measure radionuclides

Source: ISP Nuclear Power Plant NASU Chernobyl, 2018.

It is possible to certify the DTM radionuclides by using a fingerprint (nuclide vector) andthe radionuclide reference method. To develop the fingerprint, the results of the

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radionuclide analysis of representative radioactive waste samples is correlated with the key
radionuclides. Table 3.16 details the DTM radionuclides and the key radionuclides thatthey are correlated with to derive a fingerprint.
Table 3.16. Difficult-to-measure radionuclides and the key radionuclides that they arecorrelated with to derive a fingerprintGroup
Key radionuclide
number Radionuclides
Main gamma line
energies in keV
1

60

Co

1 172, 1 332

2

106

Ru

512, 10, 1 050

3

154

Eu

123,1 274, 723,
1 005, 873, 996

4

137

Cs

662

5

241

Am

59

Halflife
5.3
years
373.6
days
8.6
years
30.2
years
432
years

Difficult-to-measure radionuclides
Radionuclide
Generalradionuclide
characteristics inthe group
55
Fe, 57Co, 58Co,
Construction steel
59
Ni, 63Ni
activation products
107
Platinum group
Pd,
106
103
metals(Ru+Rh), Ru
144
147
Ce, Pm,Lanthanides with
151
comparable
Sm, 154Eu, 155Eu
chemical properties
134
Isotopes of theCs, 135Cs
same chemicalelement
244
243
Cm, Cm,Actinides with the242
Cm, 242mAm,
same properties
243
Am

Source: ISP Nuclear Power Plant NASU Chernobyl, 2018.

The applied methods and techniques should provide the specific activity measurement ofcontrolled radionuclides in a package of at least 102 Bq kg-1 for alpha-emitting
radionuclides and 103 Bq kg-1 for beta-emitting radionuclides. The specific activity of theradionuclide does not exceed the maximum permissible value.
For the specific activity of radionuclides in radioactive waste package, the following
condition must be fulfilled:

X k

= ∑
i

Aik
Ai maxwhere: A ik is the specific activity of i-th radionuclide in the k-th package of radioactive
waste in Bq g-1 and the A I max is the maximum allowable specific activity of i-th
radionuclide in the package of radioactive waste in Bq g-1.

Radioactive waste characterisation and sorting methods
The following monitoring systems are used to characterise the solid radioactive waste fromthe SRWSF compartments:•

Neutron radiation monitoring system (NMS) is used to record thermal neutrons
using 3He counters and for EDR control.

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Drum monitoring system (DMS) is used to characterise 165 litre drums into LILWSL and LILW-LL.

•

LIM system (large-sized monitoring system) is intended to record the radiation
characteristics of long-length and non-compactible solid radioactive waste loaded
into the container for disposal. It is also used to monitor the dose rate from thecontainer during its loading and to register specific activity value for certain
radionuclides.

Based on the measurement results, the expected radiation characteristics of the loaded
container are calculated, with the following threshold values being considered:•

EDR limit on the container surface;•

the limit of total activity and specific mass activity of the container;•

the maximum allowable weighted sum of specific activity values for isotopes;•

the maximum permissible weight of the loaded container.

Several monitoring systems are used in the facility to support the sorting and retrieval work
including:•Gamma camera system (GCS) is intended to register radioactive hotspots in thewaste on the sorting tray. This provides the distribution of activity on the tray andprovides preliminary predictive classification of radioactive waste for subsequent
handling.

•

Solid radioactive waste retrieval facility monitoring system (LMS) is designed torecord the radiation characteristics of the waste retrieved from the compartments as
it is loaded into the solid radioactive waste retrieval facility internal container. This
ensures that the dose rate from the container is not exceeded in the process of its
loading.

•

System of radiation control of HLW drum loading (DFF) is designed to control the
EDR while filling the drum. It is also used to record the radiation characteristics of
HLW loaded in the primary package and ensure that the set dose rate is notexceeded. The EDR limit for the filled HLW drum should not exceed 80 mSv/h.

•

System for drum radiation monitoring prior to conditioning (DFC) is intended forthe initial characterisation of the drums that arrive after loading into the containeror that come directly from the incinerator (drums with ash).

•Container monitoring system (CMS) is designed to measure EDR and surface
contamination of the disposal containers intended for shipment.

All the operational decisions are made at the control station (CS). Figure 3.31 provides aschematic of the whole process solid radioactive waste retrieval facility.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 93

Figure 3.31. Schematic of the solid radioactive waste retrieval facility
Solid waste from SWRF
LMS (KC-1)

Reception area

LMS (CS -1)

MMS (CS-2)

S&f-camera
GKS (CS-5.1)

GKS (CS-5.1)

HLW
LILW-SL
Non-compactable/non-flammable

ILW-SL
Compactable

LLW-SL

LLW-SLFlammable

Solid waste fromexternal sources

LIM

LLW-SL
Compactable

NMS

LILW -SL

DFF(CS-5.1)
LILW-LL

DFC(CS-5.2)
НСА-КСО
Ashes indrums

Placement of
HLW into 200l
drum & KTZV0.2

DMS(CS-6)

CS-8
CS-9

Incinerator

Supercompactor

Packaging into
disposal containers

LILW-SL«cakes»Packaging into
disposal containers

CS-11.1

LRW buffer
storage tank
Cementation
facility

MMS (CS-2)

Liquid combustible wastefrom external sources

Placement of
LILW-LL into outerdrum (200 l)

CMS(CS-12)

MMS(CS-13)

Buffer storage ofrecycled waste incontainers for
disposal

To disposal(SENSRWSF)

Source: Kuprianchuk et al., 2018.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

Placement of 200l drums
с LILW-LL в KTZ 2.6

MMS(CS-11.1)

MMS(CS-13)

Temporary storage of HLW
and LILW-LL in LSWSF

94 | NEA/RWM/R(2020)2

Challenges•

There was a lack of information on the composition of UL waste stored in the
SRWSF compartments. A complicated sorting system had to be devised to divide
the initial waste stream into stable sub-streams. This method proved to be effectivein this situation.

Lessons learnt•

It was important to develop a multi-stream iterative characterisation process for thesolid radioactive waste from the SRWSF to consider the unknown properties of the
UL radioactive waste.

•

The characterisation methodology, including the monitoring and measuring
equipment needed to manage the radioactive waste had to be considered for all
waste categories that needed to be retrieved including LLW, LILW-long/shortlived and HLW. As a result of this radioactive waste is being processed to meetwith the WAC for the existing storage facility and the LILW-LL and HLW streams
are being stored until commissioning of the deep geological storage (currently adeferred decision).

•

The LILW-LL and HLW streams do not require more detailed characterisation now
but do require supervision to monitor any changes in the physical and chemical
properties during interim storage.

3.4.3. Characterisation of radioactive materials and fuel-containing material at the“Shelter” Object
Power unit 4 of the Chernobyl Nuclear Power Plant was destroyed after a beyond designbasis accident on 26 April 1986 as shown in Figure 3.32.
Figure 3.32. Various views of the destroyed unit 4
a)

d)

e)

b)

c)

а) Destroyed unit 4; b) reactor core fragments; c) highly active lavas of melted nuclear fuel and metal;d) radioactive water; e) radioactively contaminated building structures.
Source: Nosovskyi et al., 2016
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Immediately after the accident a temporary protective structure was constructed over thedestroyed power unit referred to as the Shelter Object (SO) shown in Figure 3.33. However,the SO was only a temporary structure designed for a period of 15 years, so the strategy
required that a New Safe Confinement (NSC) be designed commissioned and operated. In
addition, the SO was to be transformed into an ecologically safe system (ESS).
Figure 3.33. The Shelter Object during construction in November 1986

Source: Nosovskyi et al., 2016.

The overall evaluation of irradiated fuel remaining in the destroyed power unit, based on
the results of the emergency release assessment within the territory of the USSR and othercountries, gave a value of about 180 tonnes (95% from the reactor core full loading).
The initial characterisation task was focused on assessing the nuclear safety of theradioactive material in unit 4. This involved determining the location of the highest volumeof accumulated fuel-containing materials (FCMs) and the greatest probability of selfsustaining nuclear reaction. Another task was to ensure radiation safety of personnel, bothduring the acute phase of the accident and subsequently during the operation of the SO.
The results of activities carried out from 1986 to 1992 gave a basis to develop the safetyrelated documents necessary for the operation of the SO. It also made it possible to prepare
and resolve the initial conceptual issues regarding the future of unit 4 based on theassessment of quantities, physical and chemical properties of the fuel contaminatedmaterial and radioactive material located within and outside the unit (TACIS, 1998).
The aerial and water environment in the SO were characterised, as well as other objects
including the reactor core fragments, fuel-containing materials, spent fuel in the storage
pools of power unit 4 and the SO dust. A plan view of power unit 4 is shown in Figure 3.34.
It was determined that further data validation was needed on the current safety analysis
including issues of the characterisation of all the radioactive materials in the facility. Since
the end of the acute phase of the accident, throughout the entire period of the SO operationand during the construction and commissioning of the NSC characterisation has continued
to understand and control the dynamics of changes in the state of environment and materials
that affect the safety of the facility (TACIS, 1998).
The FCM and associated radioactive waste, the SO water and the dust were selected forfurther detailed study and characterisation as they were identified as the long-term safety
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issue. It was noted that all areas identified for further research and characterisation areclosely interrelated with one another. The degradation of FCM in conditions of highradiation dose rates, cyclic temperature fluctuations and condensation of atmosphericmoisture leads to destruction of lava-like fuel contaminated material (LFCM) and increased
leachability. The distribution of FCM and LFCM is summarised in Table 3.17. This resultsin the formation of liquid radioactive waste accumulating at the lower elevations of SO andradioactive dust which without additional suppression measures and under certain
circumstances can get outside the facility.
Figure 3.34. Schematic cross section of the destroyed unit 4

Source: Nosovskyi et al., 2016.

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Table 3.17. (Larva-like) Fuel Contaminated Material distribution inside Shelter Object

Premises (mark)FCM type and state

The Central Hall (35.50). Other
upper floor premises.

Rector core fragments (RCF)
are mainly buried undermaterials spilled at the active
stage of the accident. Under
RCF the FCM can be found.
Fuel dust
In area of the scheme “E”.
Fuel assemblies with the spentfuel.
Lava-like FCM (LFCM),
core fragments.

The south cooling pond.
(18.00-35.00).
The under reactor premises
305/2* (9.00) + 307/2+ the“OR” system + the reactor
vault.
Steam distribution corridor(6.00), including FCM in thevalves.
Pool bubbler, 2d floor (PB-2)*
Pool bubbler, 1st floor (PB-1)*
304/3*
303/3*, 301/5, 301/6*,
the “elephant’s foot” and others
Premises of rector unit liquidradioactive waste plant (VSRO)
and the turbine hall.
Fuel under cascade wall.
Fuel on the site under the layerof concrete and crushed stone.

Estimated fuel in FCM (onuranium basis)
Editorial note: In 1997-1998,
the total fuel dust amount wasestimated to be approximate 30
tonnes being able to rise 10-30.

-20

LFCM

75 (+25, -35)
Editorial note: In the papers of
1996-1997 the amount of fuelon this premise was proved tobe more than 60 t.
25±11

LFCM
LFCM
LFCM

8±11
1.5±0.7
11±5

Water with dissolved salts ofuranium.

3000 m3* of water < 3 kg U

RCF
RCF, fuel dust

Source: TACIS, 1998.
* The more recent data indicates that (Nosovskyi et al., 2016):(L)FCM, premises 305/2 is 470 m3 and between 50-110 t (on uranium basis)(L)FCM, Pool bubbler, 2d floor (PB-2) approximately 50 m3 and approximately 12 t (on uranium basis)(L)FCM, Pool bubbler, 1st floor (PB-1) 0.5-1.5 t (on uranium basis)(L)FCM, premises 304/3 is 50-70 m3 and 6÷2 t (on uranium basis)(L)FCM premises 303/3 is 2-5 m3(L)FCM, premises 301/6 is 20-60 m3(L)FCM dust approximately 10 t (on uranium basis)
Liquid radioactive waste – 300 m

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+0.3
0.6
-0.2

98 | NEA/RWM/R(2020)2

Types of lava-like FCM (LFCM)
Depending on the chemical composition and conditions of formation (melt temperature,
cooling rate, etc.) all LFCM are categorised into four different groups: polychromic LFCM,
brown LFCM, black LFCM and pumice LFCM as shown in Figure 3.35. The basic
materials that took part in formation of LFCM are nuclear fuel (UO2), zirconium,
construction materials of the reactor shell, serpentine filling, sand and concrete.
Average melting temperature was 1600-1700˚C and in some cases, it reached 2 700˚C.
Figure 3.35. The four different groups of LFCM

a)

b)

c)

d)

a) Polichromic; b) brown LFCM; c) black LFCM; d) pumice LFCM.
Source: SAUEZM, SSE Chernobyl Nuclear Power Plant, 2018.

The relative mass content of separate elements in LFCM of the SO are shown in Table 3.18.
Table 3.18. The relative mass content of separate elements in LFCM
LFCM
agglomerations
Averaged forblack LFCM
Averaged forbrown LFCM

Mg

Si

CA

Zr

U

2.4

30.1

5.1

3.2

4.4

3.7

30.5

5.0

4.4

7.6

Source: Nosovskyi et al., 2016.

On 29 November 2016, the Arch was successfully constructed over the reactor building
and NSC final commissioning was held on 10 July 2019. Since August 2019, the system
New Safety Confinement – Shelter Object (NSC-SO) has been in trial operation as shownin Figure 3.36. The Arch is 257 m wide, 108 m high and 150 m long and is designed to last
100 years. The work on the first dismantling of the SO unstable structures will be carriedout between 2020-2023.

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Figure 3.36. “New Safety Confinement – Shelter Object” complex

Source: National Report Ukraine, 2017.

NBC-SO characterisation knowledge and experience
At present, it is considered that the scope of the initial characterisation performed fornumerous parts of SO on FCM, considering the conditions, are deemed sufficient for thesafety substantiation. There are still some areas of the SO that are unsafe and currently stay
unexplored. The wastes from SO and whole site that is under the concrete layer are out of
scope of the current characterisation plan and as such have been deferred.
Detailed characterisation is required prior to dismantling the unstable building structures.
However, a more precise fingerprint (nuclide vector) for LFCM and associated radioactive
waste accumulations is currently considered irrelevant due to the deferred decision on
removal, conditioning, and disposal until the geological disposal facility (GDF) is
commissioned. In the meantime, it will still be important to continue to characterise thephysical and chemical properties of the FCM and associated radioactive waste to ensure
the safe operation of the NSC-SO. Also, this information will be required to inform theretrieval and further management until disposal.

Requirements for monitoring radioactive material and LFCM at complex NSC-SO
With the NSC commissioning and the NSC-SO integrated complex formation, the third
stage of the management begins to transition from a destroyed power unit, qualified as atemporary near-surface storage facility (temporary storage facility) of unorganised
radioactive wastes, to an environmentally safe system. During this time, and since 2000,
this is deemed as planned exposure. This can only be achieved with the complete retrievalof FCM and associated radioactive waste and their further disposal in the geological storage
facility.
Implementation of this stage shall consist of the following tasks:•

study of the FCM properties and behaviour for an extended period;•

approval of the final FCM retrieval strategy including the development of retrieval
technologies;•

construction of a GDF and associated radioactive waste management infrastructure.

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The monitoring programme for the evolution of (L)FCM at the NSC-SO was agreed by theState Nuclear Regulatory Inspectorate of Ukraine (SNRIU) in 2018 (SAUEZM SSE
Chernobyl Nuclear Power Plant, 2018). This programme covers all current methods of
FCM monitoring, as well as the processing and storage of monitoring results. To
characterise the current state and changes in the properties of FCM over time, nonoperational direct and indirect monitoring are used.
Non-operational direct monitoring is comprised of studying the FCM samples in alaboratory to:•determine the rate and degree of radionuclide leaching;•understand the control over the change in the state according to the criterion ofchange in dust generation;•measure the physicochemical and mechanical properties such as strength,
microhardness, viscosity, porosity and the characteristics of the pore space;•determine the content of crystalline phases in the glass matrix of LFCM;•

determine the rate of crystal nucleation and their growth rate in the glass phase of
LFCM;

•

understand the distribution of radionuclides at the micro level in all phases;•

understand ion mobility (including Cs, Sr, U, Pu, Am radionuclides) and the degreeof their complexation in aqueous solutions in pore channels and cracks on thesurface.

In addition, CCTV monitoring will provide further useful information to underpin some of
the laboratory work programme.
Studying the mechanical characteristics of FCM will improve the understanding of theongoing physicochemical processes and, due to this, improve the existing model of the
microstructure of FCM. This will also make it possible to give a more accurate short-termand long-term forecasts of their state.
Non-operational indirect monitoring of the FCM is comprised of studying:•

Dynamic changes in aerosol activity in the air around the SO. The density of theaerosols will be measured in Bq (m2/day).

•Radionuclide composition of radioactive aerosols and its fall-out. This will focus
on changes in the ratio of 241Am/154Eu, 137Cs/154Eu and 137Cs/241Am.

•

Radionuclide and phase composition of clusters located in the immediate vicinityof the FCM clusters. This will focus on the dynamics of alpha, beta activity ofaerosols in SO premise using the stationary radiation monitoring system.

The non-operational direct monitoring supplemented by the non-operation indirect
monitoring will be used to make a comprehensive assessment of the state of the FCM in
the SO. In addition to this information on the degradation of FCM can be obtained throughthe assessment of radionuclide and phase composition of the water located near theaccumulations of fuel-containing materials. Changes in the ratio of 241Am/239+240Pu,
90
Sr/239+240Pu, 238Pu/239+240Pu and the increase in uranium concentration in the radioactively
contaminated waters will be assessed.

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To minimise the operator radiation exposure robotic tools and remote-controlled units areused to take (L)FCM samples shown in Figure 3.37.

Figure 3.37. Automated systems used to take (L)FCM samplesa) Robotic sampling by TR-4 machine in room 301/5b) Remote sampler system in room
012/7

Source: a) Nosovskyi et al., 2016, b) SAUEZM SSE Chernobyl Nuclear Power Plant, 2018.

Challenges•

The characterisation of the (L)FCM has been particularly challenging as these arenovel wastes, contain large amounts of highly radioactive radionuclides and are
extremely difficult to sample due to the physical hazards in relation to the stabilityof the SO and the radioactive dust.

•

A significant part of SO radioactive waste is the matter of a deferred decision andtherefore must be realised by future generations. Due to this decision, the primary
goal of characterisation of radioactive materials becomes understanding thechanges that are occurring to the physical and chemical properties of the radioactive
waste.

•To resolve the deferred tasks, it will be necessary to start to redefine the basic SO
terms and definitions such as the new system NSC-CO, ESS, controlled state of
FCM and end-state of NSC-SO, etc. (Kilochytska and Michal, 2018). By updatingthese definitions, it will be possible to plan what future characterisation tasks are
required at when they need to be achieved.

Conclusion
Currently, there is no final FCM retrieval strategy, while the works on the geologicalstorage facility are at the stage of preparation for site selection. Therefore, the task ofstudying the FCM properties and behaviour becomes important to ensure the safety of the
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102 | NEA/RWM/R(2020)2
NSC-SO integrated complex. The execution of this task shall make it possible to assess theneed for preventative measures aimed at reducing the risk of a radioactive release and to
make reliable predictions of the FCM behaviour over a long period of time. This
information will also be necessary to inform the development and approval of the retrieval
strategy, packaging and onward disposal.

3.4.4. Safety assessment of radioactive waste disposal site “Podlesnyi”
The radioactive waste disposal site (RWDS) “Podlesnyi” is a surface facility designed by
the All-Russian Scientific Research and Design Institute for Energy Technology(VNIPIET) in 1986. For the rest of this case study it will be referred to as Podlesnyi. Thiswas constructed for the disposal of radioactive waste in response to the acute phase of theaccident. The design life of the facility was 20 years. The facility was designed for disposalof radioactive waste with gamma dose rates up to 50 R/h (500 mSv/h) for Module A-1 and
250 R/h (2.5 Sv/h) for Module B-1. Module A 1 (shown in Figure 3.38) is filled mainlywith loose and bulk construction radioactive waste, with a total waste volume estimated at
approximately 2 650 m3.

Figure 3.38. General view and layout of the RDWS Podlesnyi Module A

Source: Consortia BS-Westinghouse-PLEJADES, 2017a.

Podlesnyi contains contaminated materials and radioactive waste that are not compatible
with near-surface disposal, regardless of whether classified as HLW according to thecurrent Ukrainian classification or as LILW according to the IAEA classification. There
are no engineering barriers directly around the radioactive waste as it was not conditioned
or stabilised in a matrix and there is limited information on the barriers used in construction.
Further research is therefore required on the functionality and durability of the engineered
barriers as well as their effectiveness to retain the radioactive content. It would be useful tocarry out some modelling to identify and evaluate possible measures to improve the safety
at the facility.
To meet the requirements of the standards applicable to the disposal of HLW or long-lived
LILW, the waste must be removed, conditioned and transferred to a repository for disposal.
The retrieval start time shall be determined by the justified concept of radioactive waste
management and the availability of geological repository. It is estimated that the

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commencement of works on the removal/disposal of radioactive waste will be possible
within a period of about 100 years, which coincides with other the management of otheraccident waste across the nuclear power plant .
Therefore, Podlesnyi can be categorised as an emergency facility not suitable for continued
storage of legacy and UL waste with insufficient knowledge about the effectiveness of its
engineered and natural barriers (Consortia BS-Westinghouse-PLEJADES, 2017b).

Characterisation of Module A
In 2017, CEMRW together with the State Scientific Research Institution “Chornobyl
Center for Nuclear Safety, Radioactive Waste and Radioecology” under a contract with the
National Laboratory in Idaho performed non-destructive studies of Modules А-1 and В-2.
This included a gamma-ray surface survey, neutron flux measurements and magnetic
measurements, as well as samples of radioactive waste fragments from the surfaces thatwere taken and investigated. The works were carried out in accordance with the technicalsolution and the work programme on Module A survey agreed with the SNRIU. Based on
the results of non-invasive studies and the “Additional measures to revive the areas
subjected to radioactive contamination as a result of the Chernobyl disaster, on social
protection of the affected people, safe radioactive waste management” (The Decree of the
President of Ukraine, 2018), a research programme for Module А-1 was developed andimplemented.
The research programme for Module А-1 defined the objectives, tasks, scope and methodsof destructive tests to clarify what information was required to characterise the facility.
This is difficult to progress because there is insufficient information to carry out acomprehensive safety assessment even though the primary objective was to justify
decisions on the future management of the radioactive waste. Also, there is limited
information on the quantity of nuclear materials and radioactive waste present in the facility
which is required to justify the physical protection measures. Measures need to be
developed to ensure the long-term safety based on the analysis of information on the stateof radioactive waste and storage structures samples which have yet to be defined.
The research programme for Module А-1 and the corresponding technical solution obtained
approval from the SNRIU. The research was carried out by CEMRW in co-operation with
SSE Chernobyl Nuclear Power Plant, Institute for Safety Problems of Nuclear Power
Plants (ISP Nuclear Power Plant) of the National Academy of Sciences (NAS) of Ukraine
and SSE “Ecocentre”.
The work was distributed among all the participant organisations:•

SSE Chernobyl Nuclear Power Plant oversaw well drilling.

•

CEMRW co-ordinated works, gamma-ray logging and neutron logging in wells andthe spectrometric analysis of samples.

•

SSE “Ecocentre” conducted analysis of the samples taken from the wells including
mass spectrometry.

•

ISP Nuclear Power Plant of NAS of Ukraine carried out modelling using theMicroShield and MCNP program codes, measuring the neutron flow density inwells, and estimating the amount of nuclear fuel.

•

CEMRW and ISP Nuclear Power Plant of NAS of Ukraine carried out temperature
logging in the wells.

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104 | NEA/RWM/R(2020)2•

CEMRW, ISP Nuclear Power Plant of NAS of Ukraine and SSE “Ecocentre”
analysed and discussed all the research results to develop of preliminary
conclusions.

Characterisation of samples
Ten samples were taken from the borehole drilled into the body of solid radioactive waste
within Module А-1. All the samples were heterogeneous in both composition and particle
size characteristics. They contained sediment, fragments of concrete and other materials ofanthropogenic origin as shown in Figure 3.39.
Figure 3.39. Sampling and preparations for measurements

Source: Ozornov, 2019.

After homogenisation of the samples, test portions of 0.5-2.0 g were collected anddecomposed by ashing in a microwave ash muffle (Stage 1) and by acid leaching upon theapplication of heat (Stage 2). The radionuclides detected in the samples, the range of
specific activity and the method of determination are shown in Table 3.19.

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Table 3.19. Radionuclide analysis of the samples taken from Module A

Radionuclide
Co
Sr
94
Nb
137
Cs
154
Eu
155
Eu
234
U
235
U
236
U
238
U
238
Pu
239+240
Pu
241Am
60

90

Specific activity range in
kBq/kg2.93 x 10-01 - 4.42 x 10+01
3.87 x 10+03 - 3.50 x 10+05
2.93 x 10-01 - 2.27 x 10+01
6.12 x 10+03 - 3.01 x 10+05
9.06 x 10+00 - 9.77 x 10+02
1.06 - 1.55 x 10+02
2.60 x 10-01 - 2.70 x 10+01
1.34 x 10-02 - 2.34 x 10+00
2.06 x 10-02 - 3.26 x 10+00
8.83 x 10-02 - 1.33 x 10+01
4.20 x 10+01 - 6.50 x 10+03
9.40 x 10+01 - 1.85 x 10+04
1.42 x 10+02 - 1.53 x 10+04

Measuring method
Gamma-ray spectrometry
RC + LS beta-ray spectrometry
Gamma-ray spectrometry
Gamma-ray spectrometry
Gamma-ray spectrometry
Gamma-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry
RC + alpha-ray spectrometry

Source: Ozornov, 2019.

The pie-chart in Figure 3.40 shows the proportion of the specific activity of each measured
radionuclide as a percentage of the total specific activity. The sample was dominated by137
Cs and 90Sr. The other radionuclides included 238Pu (0.57%), 60Co, 94Nb, 154Eu, 155Eu,
234
U, 235U, 236U and 238U.
Figure 3.40. Diagram of radionuclide concentrations in the samples from Module A-1

1%
2%

1%

44%

Cs-137

Sr-90

52%

Am-241

Pu-239+240

Other

Source: Ozornov, SSE “Ecocentre”, 2019.

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Table 3.20 presents the relationship between the specific activity of reference radionuclide(137Cs or 60Co) and DTM radionuclides. All dependencies are linear and have a highdetermination factor, i.e. the given nuclide vector (fingerprint) can be scaled to all waste
within the repository.
Table 3.20. Radioactive waste disposal site fingerprint

Number
1
2
3
4
5
6
7
8
9
10
11

Radionuclide
90
Sr
154
Eu
155
Eu
94
Nb
234
U
235
U
236
U
238
U
238
Pu
290+240
Pu
241
Am

Scaling factor
A( Sr) = 1.1063A (137Cs) - 24 127
A(154Eu) = 3.20x10-03A (137Cs) – 67.03
A(155Eu) = 5.00x10-04A (137Cs) - 8.3151
A(94Nb) = 8.00x10-05A (137Cs) - 0.6103
A(234U) = 9.00x10-05A(137Cs) – 1.975
A(235U) = 8.00x10-06A (137Cs) -0.2024
A(236U) = 9.00x10-06A (137Cs) - 0.3101
A(238U) = 4.00x10-05A (137Cs) – 1.2189
A(238Pu) = 5.20x10-03A (137Cs) + 22.151
A(239+240Pu) = 1.33x10-02A (137Cs) + 28.845
A(241Am) = 5.03x10-02A (137Cs) – 1 061.5
90

Determination factor
0.9126
0.94280.9156
0.93940.93040.91940.80030.89310.85390.89830.9446

Source: Ozornov, 2019.

In addition to the survey of Module А-1 in 2018–2019, the following specified
characteristics of the facility were obtained:•

Volumes, radioactive waste mass, concrete filling volume and sand-gravel fillingvolume.

•Significant temperature increases (up to 10°С) compared to the surrounding
temperature was revealed in 6 wells based on the results of temperature logging.
This may be due to either radioactive decay or spontaneous fission of uranium.

•

Updated information on the total activity of radioactive waste in Module А-1 by
the Monte Carlo method using the MCNP program code for 2018 was 1.1x1015 Bq.

•

Information on the radionuclide composition of the radioactive waste was obtained.
10 samples were analysed for alpha-, beta- and gamma emitters, mass andspectrometric analyses and 130 samples were analysed by gamma, beta
spectrometric analysis. The results of this data indicated that the radioactive waste
in Module A was consistent with the average fuel composition of the SO radioactive
waste.

•

Based on the results of the modelling and calculations using radionuclide
composition data, the amount of nuclear fuel was estimated at 1.6 tonnes (inuranium).

To complete the Module А-1 study several tests need to be completed including direct
neutron measurements to confirm the presence and quantity of fuel in the compartment.

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Also, repeat measurements of the temperature in the wells are needed to gain a better
statistically verified understanding of the fluctuations through the radioactive waste. The
state of the compartment structure needs to be inspected and the analysis and verificationof the results needs to be discussed with all parties.
A strategy for the development of the long-term safety and transformation into an
ecologically safe system (ESS) was published on 10 May 2019 as instructed by the
SAUEZM (SSE CEMRW, 2019). The transformation into an ESS will be achieved inseveral stages, the first of which will be to obtain more reliable characterisation informationon the state of radioactive materials, including nuclear materials and the physical conditionof barriers and physical protection. It will be important to model the state and dynamics ofchange in the characteristics of the radioactive waste by using the results from all thesampling and analytical work, to select the appropriate management techniques. This
information can then be used to define recommendations and actions to improve the safety.

Challenges•

The management of Podlesnyi is complicated because the facility has no specificstatus as the solution has been deferred. This means that there is no deadline for
either commencing or completing the retrieval of radioactive waste from the facility
or defining future characterisation tasks.

•

The characterisation activity is now aimed at obtaining iterative information on thestate of the radioactive waste and nuclear fuel in Module A with subsequentmodelling of the ongoing processes. The characterisation in this case is not aimed
at compliance to either treatment, storage or disposal WAC, it is focused onradionuclides that currently constitute radiological and nuclear hazards.

•

There have been many different organisations, overseen by the regulator, involvedin the characterisation of radioactive waste. To ensure that the project was
successful, it was important that each organisation was set specific tasks and that
there was continued dialogue between all of them. Overall, this project represents
good stakeholder interaction within the Exclusion Zone.

3.4.5. Radioactive waste temporary localisation sites
After the accident at Chernobyl, between 1986-1987, during the process of
decontamination, more than 800 disposal sites for the “radioactive waste temporary
localisation” (RWTLS) were constructed. A total volume of approximately 1 million m3 ofradioactive waste was disposed of at the RWTLS. In total, nine RWTLS were constructed,
with conventional names: "Yaniv Station", "Naftobaza", "Pishchane (Sandy) Plato", "RedForest", "Stara Stroybaza", "Nova Stroybaza", "Prypiat", "Kopachi", "Chystohalivka".
Geographically, they are grouped into the following main RWTLS: “Yaniv Station”,“Prypiat”, “Pishchane (Sandy) Plato”, “Chernobyl Nuclear Power Plant Stage III”,“Chernobyl Nuclear Power Plant Cooling pond”, “Red Forest”, “Stroybaza” and“Neftebaza”. Some of the RWTLS in the exclusion zone are shown in Figure 3.41.

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Figure 3.41. Basemap RWTLS in Exclusion Zone

RTWLS “Sandy Plateau”
RTWLS “Naftobasa”
RTWLS “Pripyat”

5
RTWLS “Yaniv Station”
RWTLS “Old stroybasa”
RTWLS “Red wood”

6 RTWLS “New stroybasa”

Source: INSC, 2017.

The main characteristics for RWTLS are summarised in Table 3.21.

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Table 3.21. Characteristics of the main radioactive waste temporary localisation sites
situated in the Exclusion Zone
Disposal
site

Construction
time

Area m2

Nature of the
site

Yaniv
station1987-1988

1 280

2-3m deep
trenches withoutwaterproofing
and clamps

Prypiat1987-1988

Pishchane(sandy)plato
Chernobyl
NuclearPower Plant
III

1987-1988

Chernobyl
NuclearPower Plant
Cooling
pond

1986

Red forest

1987

Stroybaza

1987

Naftobaza

1987

1.5m deep
11 000
16 000700 – 25.9
trenches withoutwaterproofing
0.5-1m deep880
180 000
100 000
1.0×103 trenches without
37
waterproofing
The unfinished channel of the construction stage 3 is in the industrial site area. The
length is 1.5 km, the useful volume – 265 000 m3. For 1.3 km length the channel is
faced with monolithic concrete. Its end sections are made in earthen canal. Since 1986,
it has been used as a temporary storage facility for liquid radioactive waste – mainly
highly contaminated drainage water from the IV block section. The channel for liquidradioactive waste is not equipped.
0.2 –70 000Water
0.0074It was built on the territory of the Pripyat river floodplain by a hydraulic wash of a sandenclosing dam. The area is about 23 km2. The maximum volume of water in the pond isabout 150 million m3. Activity in the bottom sediments for some radionuclides(respectively Ci - TBq):Cs-137 (4 400 – 162.8), Sr-90 ( 650 – 24.1), Pu-239+240 (14 –0.52), Pu-238 (7 – 0.26), Am-241 (18-0.67).
Sludge, sand
1-3m deep
Wood, forest
4 000
250 000
500 000
1.3×104 trenches without
leaf litter481
waterproofing,clamps with acovering of sand
2-3m deep
Wood, forest
1 210
540 000
290 000
120-4.4
trenches without
leaf litter,waterproofing
building
construction,
scrap metal
1-3m deep
Soil, wood,
410
2 200
5 000
120-4.4
trenches without
rubbishwaterproofing,clamps withsand sprinkling

1986

Characteristics of disposed radioactive
waste
Volume in Activity CiWeight in t
m3
TBq
15 00030 000
1.0×103 37

700

Disposalmaterial

Building
structuresfrom railway
station,villages,suburbansettlements
and forest
Scrap metal,
householdwaste
Contaminatedsoil

Note: The parameters of the main RWTLS (except Chernobyl Stage III and Chernobyl Nuclear Power Plant
Cooling pond) are given according to the official data published in the collection "Problems of Chernobyl
Exclusion Zone", issue 2, 1995. Other sources of information were not considered due to uncertainty of theirstatus.
Source: Kireev, 2019.

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These disposal sites were built without project documentation or protective engineered
structures. Maps and location schemes of trenches and clumps of RWTLS, compiled by thecivil defence forces while conducting priority activities to mitigate the consequences ofsevere nuclear accident, were not preserved. The total number of trenches and clumps is
between 800-1 000, in which according to the assessment and surveys results, estimate that871 000 m3 of radioactive waste was disposed. Figure 3.42 shows two photographs, one of
the RWTLS clamp/trenches and one of the Chernobyl Nuclear Power Plant cooling ponds.
Figure 3.42. Radioactive waste temporary localisation clamp/trench and Chernobyl Nuclear
Power Plant cooling ponda)Radioactive waste temporary localisation clampb) Chernobyl Nuclear Power Plant cooling pond

Source: a) Kireev, 2019; b) SAUEZM Project “SATREPS,” 2019.

Several national laws were put in place supported by a regulatory framework to provide forthe survey/re-disposal of radioactive waste at RWTLS. In 1998, the "General Safety
Provisions for Disposal of Radioactive Waste" was adopted by the Law of Ukraine. Thisrequired that all operators of waste disposal facilities periodically (at least every 10 years)
reassess and update the storage safety justification (The Law of Ukraine, 1998). In
September 2008, "The National Targeted Environmental Program for Radioactive Waste
Management" was adopted (The Law of Ukraine, 2008). This piece of legislation requiredthat:•Radioactive waste would be identified and classified.

•

The data on the amount of waste in the Exclusion Zone and the zone of
unconditional (mandatory) resettlement would be refined to provide a moreaccurate radioactive waste inventory.

•

A state radioactive waste inventory would be reviewed every three years.

•

A state database system would be developed to track radioactive waste.

The budget programme "Support of ecologically safe condition in the Exclusion Zone andthe zone of unconditional (mandatory) resettlement" (КПКВК 2408110, 2020) madeprovisions for the handling, storage and disposal of radioactive waste from Chernobyl
Nuclear Power Plant as well as the operation and maintenance of the facilities constructed
for radioactive waste management. In 2018, the SSE CEMRW extended licence No.
EO000953 that was originally issued by SNRIU in 2011 for the operation and managementof radioactive waste disposal facilities and RWTLS. This also includes measures to ensure
current and long-term safety of the closed storage facilities by exploring technical solutions
for the re-disposal of RAW in agreement with SNRIU.
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In 2018, the decree by the President of Ukraine “On additional measures to revive the areas
subjected to radioactive contamination as a result of the Chornobyl disaster, on social
protection of the affected people, safe radioactive waste management” required that thesafety status of all radioactive waste disposal sites and decontamination sites/waste storage
facilities be assessed. It also stated that appropriate measures would need to be taken to
manage and minimise harm to people or the environment (The Decree of the President of
Ukraine, 2018).

Survey of the sites
A survey of the RWTLS was carried out considering the degree of potential hazardous
impact on the environment. Several factors were used to indicate a high hazard or potential
future impact including:•

EDR over 1.0 mSv/hour for personnel;•

surface processes that may result in the resuspension of radionuclides such as thepresence of vegetation (grass and trees) with the added risk of frequent fires;•changing level of groundwater;•

earthworks performed near the RWTLS locations.

These factors were used to determine which areas were to be surveyed first. A full survey
was conducted on the trenches and clamps of the RWTLS’s located at: "Peschanoe Plato"and "Naftobaza" as they are near the river Prypiat, and "Nova Stroybaza " and "Stara
Stroybaza” as they may impact on the safety of the SO transformation into an
environmentally safe system as well as the decommissioning of the Chernobyl Nuclear
Power Plant.
The inspection and decision-making to assist in the characterisation of waste in the trenches
and clamps was undertaken in a number of stages. Radiometric surveys, gamma and beta
surveys of the areas were conducted as well as surface soil sampling. A visual search fortrenches and clamps was undertaken using a set of anthropogenic and geobotanicalattributes with GPS receiver positioning.
All the survey measurement results were compiled and statistically analysed to form a mapof the site. This information was then used during the final stage of the preparation of thetechnical decision considering two options, either to retrieve the radioactive waste andtransfer it to a new storage site or apply additional measures to strengthen the existing
structure so that it meets safety requirement without re-disposal.
After regulator approval of the technical decision, CERWM prepared the work
implementation programme, including the necessary organisational, technical and safety
actions. The following field surveys were carried out using a mobile laboratory:•

in field determination of the radionuclide composition and the measurement of thespecific activity directly in the clamps and trenches using an external HPGe
detector;•

determination of the radionuclide composition and measurement of the specificactivity in objects located in hard-to-reach areas using a hand-held scintillation
spectrometer in calibrated geometry;•obtention of gamma-ray images using a gamma-ray monitor to detect abnormal
contamination;

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112 | NEA/RWM/R(2020)2•

determination of the geometric characteristics of the storage facilities, geologicalconditions and groundwater levels using ground penetrating radar;•sampling of radioactive materials for further analyses at an external laboratory;•

determination of macro element composition of samples including themeasurement of their density and humidity.

At the time of writing this report, approximately 60% of the objects sampled(approximately 800) have been investigated.
Based on the results of the analysis of the current, mid and long-term evolution of theRWTLS, areas have been identified where unauthorised human intrusion from construction
work may have a high radiological impact. Direct radiation exposure and resuspension ofdust containing hot particles and particles of irradiated fuel are the most evident
contribution from the point of view of the safety of the disposal facility. Therefore, theranking of trenches and clamps by its hazardous nature has been developed. To rank the
trenches and clamps the following levels of intervention were specified:•

EDR over 100 µSv/hour;•

gamma and beta-emitting radionuclide content over 10 kBq/kg;•

alpha-emitting radionuclides <100Bq/kg and for alpha-emmitting sources oftransuranics ≤ 6 kBq/kg;•

those containing the key radionuclides 137Cs, 90Sr, 239Pu, 239+240Pu that have beenhighlighted for re-disposal.

The monitoring boreholes, the majority of which were drilled before the accident, cannot
provide adequate information (or control) over discharges of contaminated groundwaterfrom the adjoining RWTLS flowing into the River Pripyat as shown in Figure 3.41. It istherefore important to consider this information when ranking the sites.

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Figure 3.43. Exposure dose rate gamma survey (logging) for two boreholes at clamp 33

Source: INSC, 2017.

Lessons learnt•

The activities performed to reassess the current safety of the RWTLS in the
exclusion zone are useful whether taking a decision now or postponing that decision
for future evaluation. On the one hand, if the results of characterisation indicate that
the RWTLS radioactive waste is suitable for re-disposal, then after validation of
the data it will be retrieved and taken to another safer storage site. If not, thenimprovement actions can be developed for strengthening the safety of RWTLS
storage while a future decision is taken.

Regulatory framework and waste acceptance criteria
The Supreme Council of Ukraine, Verkhovna Rada, approves the legislative framework onradioactive waste management. This is the executed through state policy public
administration bodies, the Ministry for Protection of the Environment and Natural
Resources, and the Ministry of Energy and Coal Industry (Minenergougol), and then
NAEK "Energoatom". The policies formed by the public administration bodies are thendelivered by the State Nuclear Regulatory Inspectorate of Ukraine (SNRIU) and theMinistry of Health protection.
The principles of state policy on radioactive waste management include the priority toprotect the life and health of personnel, population and the environment. The “polluter

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pays" and the producer has no right to dispose of it. This is implemented by meeting thenationwide targeted ecological programme for radioactive waste management.
This work is funded through an ecological tax paid by nuclear power plants and otherindustries, research facilities and hospitals producing radioactive waste.

The main stakeholders responsible for the management of UL wastes in the Exclusion Zone include:
•
•
•
•
•
•
•

State Agency of Ukraine for Exclusion Zone Management;
Centralised Enterprise for Radioactive Waste Management;
Institute for Safety Problems of Nuclear Power Plants of Ukrainian;State Specialised Enterprise “Chernobyl Nuclear Power Plant”;
State Specialised Enterprise “Ecocentre”;
State Scientific Research Entity “Chernobyl Centre for Nuclear Safety, Radioactive Waste andRadioecological problems”;
State Nuclear Regulatory Inspectorate of Ukraine.

Challenges•

There have been some significant changes in radioactive waste management
legislation and policy over the last couple of years that will have an impact on thedevelopment of future waste characterisation objectives and tasks.

•

Several Acts need to be amended to make improvements to radioactive waste
management. The Law of Ukraine needs to be amended around:‒ the use of nuclear power and radiation safety;‒ radioactive waste management;‒ decision-making on the siting, design and construction of nuclear facilities andnational important facilities intended for radioactive waste management.

•

A new waste classification system has been established with four classes ofradioactive waste. Very low-level waste (VLLW), low-level waste (LLW),
intermediate-level waste (ILW) and high-level waste (HLW).

•

There has been some change to the definition of some terms such as “Disposal ofradioactive waste” which means the placement of radioactive waste in storage,
without the intention of their retrieval, for the purpose of storage and isolation ofradioactive waste from the biosphere. Also, the “Radioactive waste acceptancecriteria for disposal” has been defined as the quantitative parameters(characteristics) of radioactive waste or qualitative requirements determined by theoperating organisation (operator) for the disposal of radioactive waste. In addition,
the state regulatory authority for nuclear and radiation safety is involved indetermining and agreeing whether radioactive waste can be accepted for burial.
Radioactive waste disposal facilities are specified according to the radiologicalcharacteristic of the waste being disposed of including surface disposal facility,

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near surface disposal facility, mid-depth disposal facility, geological disposal
facility.
•

The state is responsible for determining whether characterisation is sufficient tomake decisions on radioactive waste management. There is the need to establish anorm for specifying the requirements of independent checking of radioactive waste
characterisation result by a third party.

Conclusion
Any characterisation activity should be aimed at meeting the objectives that have been set.
Characterisation of UL waste, performed without clearly formulated objectives and tasks,
can be costly and inefficient, as well as lead to unexpected consequences. In general, theresults of the characterisation activity should serve as the basis for planning of future tasks.
This is an iterative process which should ultimately results in the characterisation/sortingof radioactive material to decide on whether it meets the relevant criteria i.e. to be left inits current state with or without additional engineering, or be retrieved for treatment,
storage and/or disposal. Where the disposal decision is currently deferred thencharacterisation may continue to be used to monitor and manage the physical and chemical
properties of the waste to maintain the safety assessment.
The management of large volumes of UL radioactive waste imposes a great responsibilityon the state to ensure the current safety of personnel and the public, as well as the safety of
future generations in terms of potential exposure. To ensure that the right decisions are
made, the state must be sure that these decisions are made using accurate and reliablecharacterisation information. Such confidence in characterisation results can only beachieved by state regulatory supervision at all stages from the development of standards
and issuance of licences through to independent checks of the characterisation results by
the third party designated government body or regulator. To minimise the regulatoryburden on the Licensee, a list of cases when such control is necessary, needs to beestablished. In the meantime, the role of the regulatory threat assessment process may serveas the appropriate method to make these decisions, as applied in Ukraine (Sneve et al,2016).

References
Antropov, V.M., O.V. Marinich, O.G. Tretyak, A.I. Khabryka (2020), “The current state of the radioactive
waste temporary storage sites at the Chornobyl exclusion zone”, State Specialized Enterprise «Central
Enterprise for the Management of Radioactive Waste», Chornobyl, Ukraine.
BS-Westinghouse-PLEJADES (2017a), INSC Project U4.01/10 (D) Investigation of radioactive waste
burial and temporary storage sites in the Chernobyl Exclusion Zone. Tasks D3, Addendum 12, RWDS
“Podlesnyi” site description, Consortia BS-Westinghouse-PLEJADES.
BS-Westinghouse-PLEJADES (2017b), INSC Project U4.01/10 (D) Investigation of radioactive waste
burial and temporary storage sites in the Chernobyl Exclusion Zone. Task D5, Draft of Technical
Report, Implementation of accepted approach to RWTLS safety assessment and ranking withrecommendations on their safety.
Cabinet of Ministries of Ukraine КПКВК 2408110 (2020), Passport of the budget program “Support ofecologically safe condition in the Exclusion Zone and the zone of unconditional (mandatory)
resettlement”

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

116 | NEA/RWM/R(2020)2«Підтримка екологічно безпечного стану у зонах відчуження і безумовного (обов'язкового)відселення».
INSC PROJECT U3.01/10 UK/TS/46, Subtask 2.1 (2017), “Support to SNRIU in the field of radioactive
waste management” (Component A), Instrument for Nuclear Safety Cooperation, European
Commission.
ISP Nuclear Power Plant NASU (2017), “Development of characterization methods for SRW and ICSRM
RAW packages”, R&D reporting and presentation materials by the Contract No 378/17 as of
05.10.2017, Institute for Safety Problems of Nuclear Power Plants of Ukrainian NAS, Chernobyl.
Kadenko, M. and V. Pluiko (2003), “Principle methods to assess statistical data and functions for theirdistribution”, Training manual, Kiev university, pp 45, 48, 94-98.
“Основи методів оцінки статистичних даних та функцій їх розподілу. Навчальний посібник. К.:Видавничо-поліграфічний центр "Київський університет", 2003.
Kilochytska, T. and V. Michal (2018) “IAEA prospective for post-accident situations. Chernobyl Shelter
after 30 years of recovery: status, challenges and future safety”. International Atomic Energy Agency.
Vienna 2018.
Kireev, S. “Central Enterprise for Radioactive Waste Management (SSE CERWM)”, presentation, SSE
CERWM 2019"Центральное предприятие по обращению с радиоактивными отходами“ (ГСП "ЦПОРО").
Презентация 2019.
Kuprianchuk, S., M. Pashinov, R. Godun (2018), “Issues of RAW characterization at Solid Radioactive
Waste Treatment Plant of Chernobyl Nuclear Power Plant”, Institute for Safety Problems of Nuclear
Power Plants of Ukrainian NAS (ISP Nuclear Power Plant NASU), Chernobyl.
NEA (2019), “Challenges in Nuclear and Radiological Legacy Site Management: Towards a Common
Regulatory Framework”, OECD Publishing, Paris. www.oecd-nea.org/jcms/pl_40359
NNEGC (2013), Short-existing, low and intermediate level nuclear power plant waste. Requirements forthe end conditioned product. Energoatom, No.1007, 1.037:2013, 03.12.2013
Krasnov V., A. Nosovskyi, V. Rudko, V. Scherbin (2016), The “Shelter” Object: 30 Years after the
Accident, Institute for Safety Problems of Nuclear Power Plants of National Academy of Sciences of
Ukraine (ISP Nuclear Power Plant NASU), ISBN 978-966-7875-2, pp 62-64, 80-114, 184-195.
Ozornov, A. (2019) “Characterization of the RAW samples from Module “A-1” of RWDF Podlesnyi R&D results presentation”, SSE Ecocentre.
“Характеризація проб РАВ модуля А-1 ПЗРВ «Підлісний»”, Державне агентство України зуправління зоною відчуження, ДСП «Екоцентр», 2019.
Pokhodun, A. (2006), Experimental investigation techniques. Bias and uncertainties ofmeasurements, Training manual, pp. 20-23, 80-82.
Экспериментальные методы исследований. Погрешности и неопределенности измерений /Учебное пособие. - СПб: СПбГУ ИТМО, 2006.
Rovno Nuclear Power Plant Varash (2016), “Analysis of the conditions of formation of saline bitumen
compound at SD Rovno Nuclear Power Plant (physical and chemical characteristics of VAT residue
arrived to bituminizer facility) in order to identify the number of sampling to conduct detailcharacterization of radionuclide content”, approved by Deputy Chief Engineer Ph.Kislitsin.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 117“Анализ условий образования битумного компаунда ОП РАЭС (физические и химические
характеристики кубового остатка, который подавался на установку битумирования) с целью
определения объема проб для проведения детальной характеризации радионуклидного
состава”.
SAUEZM (2018), Japan-Ukrainian Joint Project “SATREPS”, presentation, Chernobyl -Tokyo
SAUEZM (2018), “SO fuel containing materials (FCM) monitoring program (SO Nuclear and radiation
safety)”, SAUEZM, SSE Chernobyl Nuclear Power Plant, approved by decree of SNRIU, pp 48-70.
“Програма моніторингу паливовмісних матеріалів (ПВМ) ОУ (ядерна та радіаційна безпека
ОУ)”. Державне агентство України з управління зоною відчуження. Державне спеціалізоване
підприємство «Чорнобильська АЄС»SE NNEGC Energoatom (2019), “Nuclide vector to characterise SD Rovno Nuclear Power Plant salinebitumen compound”, approved by NNEGC Energoatom First Vice-President, Technical Director
O.Shavlakov, approved by decree of SNRIU in its letter No.15-41/2877-3201.
Радіонуклідний вектор для характеризації солебітумного компаунду ВП «РАЕС». Робочі
матеріалиSENSRWSF (2009), “Radioactive waste acceptance criteria for waste disposal in specially equipped nearsurface solid radioactive waste storage facility (SENSRWSF)”, Revision 5. – Chernobyl.
“Критерії приймання радіоактивних відходів на захоронення в спеціально обладнаному
приповерхневому сховищі твердих радіоактивних відходів (СОПСТРВ)”, Перегляд 5.
Державне агентство України з управління зоною відчуження, “Центральне підприємство зповодження з радіоактивними відходами». Чорнобиль.
Seyda, V., V. Kuchinskiy, D.Stelmakh. (2019), “Existing RAW management system at SSE Chernobyl
Nuclear Power Plant (working materials)”, SSE Chernobyl Nuclear Power Plant.
Siegien, K., M. Sneve, I. Kostenko, Y. Yesypenko, O. Pecherytsia (2016), “Ukrainian Regulatory Threat
Assessment: identifying priorities for improving supervision of nuclear and radiation safety andsecurity”. StrålevernRapport 2016:10. Østerås: Statens strålevern.
SNRIU (2018), "General Safety Provisions for Disposal of Radioactive Waste", Clause 17 Chapter ІІ,
approved by the order of the State Nuclear Regulatory of Ukraine dated 13 August 2018, No.331.
SNRIU (2004), “Recommendations for establishing waste acceptance criteria for disposal of conditioned
radioactive waste in near-surface disposal (RD 306.4.098-2004)”, approved by decree of SNRIU,
No.160, as of 25 October 2004.
Рекомендації щодо встановлення критеріїв приймання кондиційованих радіоактивних відходів
назахороненняу приповерхневих сховищах
(РД
306.4.098-2004),затверджено
наказом Держатомрегулювання від 25.10.2004 № 160.
SSE CEMRW (2019), “Action strategy on ensuring the operation safety of RWDF Podlesnyi”, SAUEZM,SSE CEMRW Central Enterprise for management of radioactive waste. Chernobyl.
“Стратегія дій щодо забезпечення безпечної експлуатації ПЗРВ "Підлісний", Державне
агентство України з управління зоною відчуження, “Центральне підприємство з поводження зрадіоактивними відходами». Чорнобиль.
TACIS (1998), “The Shelter’s Current Safety Analysis and Situation Development Forecasts (updated
version)”, pp.18-30.

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118 | NEA/RWM/R(2020)2
The Decree of the President of Ukraine (2018) “Additional measures to revive the areas subjected toradioactive contamination as a result of the Chernobyl disaster, on social protection of the affected
people, safe radioactive waste management”, No.196/2018, Item 4.
The Law of Ukraine (2008) “National Targeted Environmental Program for Radioactive Waste
Management”, No.516-VI as of 17 September 2008.

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3.5. United Kingdom
The United Kingdom’s nuclear programme commenced in the 1940s and is now some
80 years old. Over that time, a wide range of radioactive wastes has been produced fromcivil nuclear reactors, weapons programmes, research sites and fuel reprocessing sites.
Although there is no official definition of “legacy waste” in the United Kingdom, it canbest be described as the radioactive waste produced during the infancy of the UK civil
nuclear industry’s development, at a time when, unfortunately, waste storage and treatment
was not well planned. Therefore, there is limited available information on either the waste
provenance or of waste characteristics. Also, little consideration was given to retrieval fromstorage facilities and to subsequent waste processing.
The United Kingdom has a legacy of unconditioned higher activity waste (HAW).
A significant fraction is stored in ageing facilities, such as vaults, silos and ponds that do
not meet modern standards. The Nuclear Decommissioning Authority (NDA) is
responsible for prioritising the retrieval, transfer, conditioning and passive storage of these
HAW. Some UK nuclear sites also contain areas of radioactively contaminated land,
generated from leaks, spills and historical operations, generally in the early part of the site’shistory. These wastes can be considered as unconventional (UL) wastes and are
overwhelmingly of low activity but can be of large volume. The volume of radioactively
contaminated land beneath the Sellafield site is estimated to be several million cubic
metres; this site accounts for most of the radioactively contaminated land on UK nuclear
sites.
It is evident that there are various type of UL wastes in the United Kingdom that havedifferent characteristics, such as activity levels and volume, and will require a range of
waste management solutions. As the information required from waste characterisation
depends on the characterisation objective, it follows that there will be no single optimum
characterisation approach for UL waste. Therefore, rather than present a specific example,
this case study is based on good practice developed for the characterisation of solidradioactive waste. This identifies the key issues, lessons learnt and “pitfalls” that are
relevant to the characterisation of large volumes of lower activity UL waste.

Situation/Objectives
The characterisation of radioactive waste produced on nuclear sites in the United Kingdom
is the responsibility of the nuclear site operator; these are the companies such as Sellafield
Ltd, Magnox and DSRL, who operate the legacy sites on behalf of NDA. There are no
regulatory requirements on the characterisation of radioactive waste. However, suitable andsufficient characterisation is required by regulation to achieve specific objectives, to:•Enable the development of optimised waste management solutions. This will also
ensure that waste packages can be safely transported, stored and disposed.

•

Appropriately sentence wastes and meet the requirements of organisations thattreat, store and dispose of radioactive waste through compliance with WAC.

•

Support compliance with regulatory requirements.

•

Help maintain stakeholder confidence that radioactive waste is being managed anddisposed of appropriately.

•

Facilitate appropriate planning.

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Recent inspections by regulators have concluded that waste characterisation practice in the
United Kingdom is steadily improving but some areas for enhancement were identified.
Consequently, the NDA commissioned a Good Practice Guide (GPG) on thecharacterisation of solid radioactive wastes. This GPG details the principles, processes andpractices that should be followed when characterising solid radioactive waste. The
document has now completed the review process and is being “trialled” by site operators
to evaluate its “usability”. The GPG is not currently publicly available, however, this casestudy summarises the key aspects. Separately, nuclear industry guidance in the United
Kingdom is already publicly available on clearance and radiological sentencing (Nuclear
Industry Safety Directors Forum, 2017). This Good Practice Guide details the principle,
processes and practices that should be used when determining whether an article or material
may be released from any further regulatory controls. This is determined based onradiological protection considerations.

Government policy and regulatory framework
United Kingdom government policy (DEFRA et al., 2017) does not recognise UL wastes
as distinct categories of radioactive waste. The management of radioactive waste on nuclear
licensed sites (NLS) in the UK is regulated by the Office for Nuclear Regulation (ONR)
and the environment agencies for the relevant devolved administration. The environment
agencies (via permits or authorisations) regulate the generation, management and disposalof radioactive waste. These requirements need to consider all stages of the life cycle of the
NLS, from design, construction, commissioning, operation and decommissioning.
Over the last three years these permits/authorisation have evolved to consider the
management of both radiological and non-radiological waste as well as describing thestandards that operators must meet to surrender their permits (Environment Agency, 2018).
The permits/authorisations now require that the operator maintain a waste management
plan and a site-wide environmental safety case, which together demonstrate throughout thelife cycle of the regulated facility, how the:•production and disposal of radioactive waste is managed to protect the environment
and to optimise the protection of people;•

disposability of radioactive waste that will require disposal on or from the premises
is assured;•public and the environment are protected from the non-radiological hazards ofdisposals of radioactive waste;•premises will be brought to a condition at which it can be released from regulation
under this permit.

The operator must apply optimisation when developing its waste management planbalancing the safety of people and the environment against the cost, potential future usesof the site or the impact of transporting waste and materials (Environment Agency et al.,2018).
Environmental permits/authorisations include a range of other conditions, such as requiring
Best Available Techniques (BAT) to be applied to the generation, management or disposalof radioactive waste. The nuclear licensing system enforced by the ONR requires licensees
to have arrangements in place to minimise, so far as is reasonably practicable (SFAIRP),

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the rate of production of radioactive waste and the total quantity of radioactive waste thatis accumulated on the NSL.
There are several high-level waste management principles that waste producers must
consider in decision-making:
•

BAT;

•

optimisation;•

integrated planning;•the “proximity” principle;•

the SFAIRP principle.

In addition, both the ONR and the environmental agencies have developed specificregulatory principles applicable to waste characterisation. For example, the environmental
regulators require waste producers to use BAT / BPM to characterise and segregate theirradioactive waste based on physical and chemical properties and treat and dispose ofradioactive waste. Regulatory guidance states that waste should be characterised within aquality framework using BAT and that a systematic approach to waste characterisation
should be adopted to acquire data that are sufficient to support waste management decisions
throughout the waste life cycle.

Waste acceptance criteria (WAC) and disposability assessment
All waste storage, disposal and treatment facilities in the United Kingdom use WAC to set
out the requirements of the site operator for the acceptance of waste to be treated or
disposed. The purposes of WAC are to:•ensure compliance with Environmental Permits, Environmental Safety Cases andother environment regulatory requirements set out in the Guidance on theRequirements for Authorisation (the GRA) for disposal at near surface facilities
and geological disposal facilities (Environment Agency et al., 2009);•

assure operational and long-term safety requirements for all components including
handling, transport, storage and disposal;•

act as a benchmark for communication between waste producers and facility
operators and will form part of a BAT/BPM assessment to identify optimal waste
management options;•allow for standardisation of operations (waste package handling, on-site transport,etc.);•

allow for optimisation of the facility and enable the disposed inventory to be
maximised with confidence in overall safety performance;•

assure waste tracking;•assist waste producers in selecting appropriate waste package and processing
technologies.

A key point of relevance to the management of large volumes of lower activity UL waste
is that some such wastes have already been disposed in the United Kingdom, either to the
Low Level Waste Repository (LLWR) or to conventional waste disposal sites that have
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been licensed for the disposal of VLLW. A case study can be found in the NEA publication,
Challenges in Nuclear and Radiological Legacy Management: Towards a Common
Regulatory Framework (NEA, 2019).
The United Kingdom has not yet developed a geological disposal facility (GDF) for the
disposal of higher activity solid radioactive waste. This is because the details of a future
GDF are unknown, it is not yet appropriate or possible to develop WAC for higher activity
wastes (HAW). Radioactive Waste Management (RWM), the organisation responsible for
implementing government policy for the disposal of HAW, has therefore developed the
Disposability Assessment Process to facilitate the assessment of HAW that are destined for
disposal at a future GDF. The disposability assessment process allows waste producers toprocess (i.e. pre-treat, treat and condition) higher activity radioactive waste in a mannersuitable for disposal in a future GDF. One aspect of this process is to ensure that theradionuclide content and composition of waste consignments received for future disposal
at a future GDF are sufficiently well characterised to comply with the condition of a futureauthorisation from the environmental regulator. To establish the package properties, RWM
undertakes a technical evaluation of the “Nature & Quantities” of the waste based on thecharacterisation information provided by the waste producer. This assessment uses thecharacterisation information provided by the waste producer and provides information thatis used in subsequent steps in the GDF Disposability Assessment process. It has thefollowing objectives:•

Critically analyse the quality of the characterisation information available.

•

Provide, or confirm, a description of the origins, composition and quantity of thewaste proposed to be packaged.

•

Define the expected range of waste package compositions resulting from the
proposed packaging process, covering physical and chemical composition,
including packaging materials and radionuclide composition.

•

Establish the relative significance of the quantities of radionuclides expected to bepresent by comparing the inventory determined through the “Nature andQuantities” assessment with the total inventory for disposal to the GDF. The latteris contained in the Derived Inventory and is used for GDF post-closure safetyassessment (RWM, 2014).

Within this framework, guidance on the characterisation of solid radioactive waste has been
produced.

Challenges
Waste characteristics are required for the following reasons:•

The radioactive properties (such as radionuclide activity concentrations, total
activity inventories and external dose rates) and the concentrations of hazardous
non-radioactive properties should be characterised. This information is required toenable correct classification and sentencing of waste and to enable safe treatment,
packaging, transport, storage and disposal options to be developed.
Characterisation of waste for hazardous substances is required because constraints
will apply on the concentrations and inventory, or even the presence, of somehazardous substances for some waste management options.

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Relevant physical, chemical and biological characteristics that determine how thewaste will behave when it is processed, transported, stored or disposed. For
example, an understanding of the physical size, material components and flow
characteristics/viscosity of raw waste is needed to enable design of a waste retrieval
process. Chemical and biological processes can result in significant changes to thewaste, encapsulant and container, which can impact the subsequent performance of
the waste package.

•

The principal challenge for characterising UL wastes is that less is known about
their provenance and composition than for other radioactive wastes, which means
additional characterisation is likely to be needed. There are also fewer opportunities
to infer the characteristics of UL wastes from review of existing information,
understanding of provenance or through comparison with similar waste elsewhere.

A systematic approach to planning waste characterisation
A systematic process, such as that illustrated in Figure 3.42, is recommended forcharacterising solid UL waste for radioactivity, hazardous non-radioactive substances andrelevant chemical properties, physical and biological properties. The approach aligns withthe widely used Data Quality Objectives (DQO) process developed by the United States
Environmental Protection Agency (US EPA, 2006). The final step in Figure 3.44, decisionmaking, is not part of the characterisation process but is included here to emphasise thatcharacterisation should always be undertaken for a purpose.
Applying a systematic process will help to ensure that when a waste characterisation
campaign has been completed, it will have provided sufficient data to make the required
decision with an appropriate level of confidence. Following completion of each of theplanning, implementation, assessment and decision-making phases of the process, it isbeneficial to undertake a “Learning from Experience” (LfE) session. LfE from previous
waste characterisation projects should be reviewed before embarking on a new
characterisation project. It is also important to ensure that suitably qualified andexperienced staff are involved in all the planning, implementation, assessment anddecision-making steps.

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Figure 3.44. Example of a systematic process for characterising solid UL waste

Source: Adapted from Figure 4.1 NDA, 2019.

Step 1: Identify the characterisation objective
The temporal and spatial boundaries of the waste to be characterised should be determined.
Heterogeneity is likely to increase due to the presence of more materials or of wastes
derived from more than one sources or process. It is worth considering segregating waste
during pre-treatment, which is likely to simplify subsequent characterisation and thesubsequent waste management. It is important to recognise that a balance needs to be struckin the cost of segregating and characterising many homogeneous waste streams against asmaller number of less homogeneous waste streams.
It is necessary to identify the characterisation objective, as this will determine what needs
to be measured. For example, is characterisation being undertaken to identify the optimum
waste processing solution or to optimise a chosen solution or to sentence waste? The
objective may be to make a yes/no decision or to build understanding of some aspect ofwaste behaviour.
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Step 2: Determine the information required to meet the characterisation objective
A simple process such as the following should be used. Developing questions that may be
useful to help identify what characterisation information will be required.
•

Are all radionuclides within the waste to be measured in every sample, or will aradionuclide fingerprint approach, coupled with analysis of an easy-to-measureradionuclide (or a small number of easy-to-measure radionuclides), be usedinstead?

•What is needs to be known to assess the environmental and human healthconsequences of alternative waste management strategies?

•

What needs to be known to ensure safe transport of waste?

•

What are the requirements of the WAC/Conditions for Acceptance (CFA) for theplanned treatment and storage solutions?

•

What do needs to be known to demonstrate the disposability of the conditioned
waste?

For any characterisation objective, one of the drivers for characterisation will be to ensurethat subsequent management of the waste is optimised; for example, that BAT is selected
and that doses are kept as low as reasonably practicable (ALARP). The answers to thesequestions should start to develop the list of parameters and processes for which informationis required. These should cover all of contaminant concentrations and the physical,
chemical and biological characteristics of the waste. A possible pitfall is to neglect toidentify information requirements for non-radioactive contaminants and on waste
characteristics.
The next consideration is to determine how each piece of information will be used toachieve the characterisation objective. Is it for decision-making or to build understandingof how the waste will behave during subsequent waste processing, storage and disposal?
Once this is understood it is possible to decide and justify whether the value for theparameter or characteristic should be determined by a qualitative approach (such as expert
judgement) or a quantitative approach (such as probabilistic analysis). For either the expert
judgement or probabilistic approach, it is good practice to decide whether theparameter/characteristic is to be represented by a best-estimate value or a cautious value(or whether both measures are required). A decision must then be taken and documented
on what level of confidence (or conversely “uncertainty”) is acceptable for the
representative (or “best-estimate”) value of the parameter/ characteristic.
The output will specify the information required, how each parameter is to be represented(“best-estimate” or “cautious estimate”), the acceptable level of uncertainty of eachparameter and the required accuracy and limit of detection of the characterisation
technique. This will enable appropriate approaches to characterisation and analyticaltechniques to be selected (Step 5). A “life cycle approach” to waste characterisation isrecommended, as this approach will ensure that all the information that may eventually berequired on a UL waste is considered at the outset. A “life cycle approach” does not meana “just in time” approach or always acquiring only the data needed to achieve the immediatecharacterisation objective. The objective of the approach is to optimise the number ofcharacterisation phases, the timings of these phases and the information acquired in eachphase.

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Steps 3 and 4: Review and evaluate existing information, and decide whether further
characterisation information is required
The review should establish the quality of the data and its suitability for describing thecharacteristics of the waste. Any limitations in the subsequent use of the information should
be identified. Generally, it is appropriate to carry out the review of existing information
against a set of requirements informed by the characterisation objective, required levels ofconfidence and potential waste management options. It is recommended that a systematic
review should consider aspects such as:•

Is information available on all the required parameters and characteristics?

•

Is the information of the appropriate quality? For example, are detection limits andmeasurement accuracies acceptable?

•

Is the information suitable for determining representative values (or “best-estimate”
values) of required parameters and characteristics?

•

If an understanding of uncertainty is required, is there sufficient information toreach a decision at the required level of confidence?

The basis on which information was collected should be understood. Was it intended to be
representative of the waste or to reflect an aspect such as a “contaminant hot spot”? If thebasis is not provided in the documents reviewed, then it is recommended that such data
should not be used to derive representative values for the waste. A lack of knowledge on
the processes that generated some large volume UL waste and the lack of similar wastes onother sites limits the extent to which waste characteristics can be inferred. This is a key
difference between review of conventional and UL wastes, and means the state ofknowledge of the latter at the start of new data acquisition activities is generally lower than
the former.
Following the review, a decision should be made on whether further characterisation
information is required to meet the characterisation objectives, or whether the existing
information / inference is adequate.

Step 5: Develop the approach for characterising the waste
New information to characterise UL waste can be obtained in three ways:•

by calculation (nucleonics calculations of radionuclide activity concentrations);•

by non-destructive assay;•

by sampling and analysis.

It is important to recognise that a combination of the above approaches may be appropriate,
and that characterisation is likely to be an iterative process. From cost, efficiency, healthand safety and environmental perspectives, characterisation by calculation should beconsidered first, followed by characterisation by non-destructive assay and then finallycharacterisation by sampling and analysis. Notwithstanding this, the waste producer should
decide and justify which approach or combination of approaches to acquire new
information is most appropriate for the waste under consideration. Generally, large volume
lower activity UL waste is likely to be characterised using sampling and analysis or
combinations of sampling and analysis and non-destructive assay.

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The following subsections provide guidance on aspects of characterising large volumes oflower activity UL waste.
Developing radionuclide fingerprints
Radionuclide fingerprints can apply at different levels. At the highest level, a fingerprint
can describe the overall proportions of radionuclides in an operational process. There is
likely to be a large amount of information available to generate such a fingerprint. At alower level, the fingerprint can describe the average radionuclide composition of a waste
stream, part of waste stream or an individual component. Generally, there are only limited
data to generate waste-related fingerprints, and this will influence the approach to
producing the fingerprint.
Radionuclide fingerprints are produced for most radioactive waste. Once the fingerprint
has been established, waste is generally characterised during subsequent waste
management operations based on the analysis of a small number of “easy-to-measure”
gamma-emitting radionuclides: for example, Co-60 and Cs-137. The concentrations ofother radionuclides in the waste are derived by scaling the fingerprints according to themeasured activities of the easy-to-measure radionuclides. Furthermore, detailed
characterisation will be required if the waste management process results in segregation or
separation of radionuclides within the waste.
New radionuclide fingerprints will be required for large volumes of lower activity UL
waste because of the lack of comparable wastes elsewhere. Several radionuclide
fingerprints, for different waste materials (e.g. soils, concretes, vegetation) and different
areas of the site, are likely to be needed. The following should be considered whendeveloping radionuclide fingerprints:•

Identify the planned use or uses of the fingerprint. Do not use the fingerprint forother uses without considering whether it is appropriate for those other uses.

•

Understand the acceptable variability in the fingerprint, which will depend on theuses to which it is put. If the observed variability exceeds this level, segregate thewaste or sub-divide the waste volume and produce individual fingerprints for subsets of the waste.

•

Decide whether it is appropriate to determine the fingerprint based on “bestestimate” proportions of radionuclides? Or on cautious estimates of the proportionsof those radionuclides that have the greatest effect on waste management decisions?
Best estimates are generally appropriate for large volumes of lower activity UL
waste.

•

Are there implications arising from radioactive decay or ingrowth? Should
corrections be made to account for time?

Once the radionuclide fingerprints have been established, large volumes of lower activity
UL waste during subsequent waste management operations (e.g. demolition, dismantling,
remediation of contaminated land) will generally be characterised by non-destructive assay
based on the analysis of a small number of “easy-to-measure” gamma-emitting
radionuclides: for example, Co-60 and Cs-137. The concentrations of other radionuclides
in the waste are derived by scaling the fingerprints according to the measured activities ofthe easy-to-measure radionuclides. Furthermore, detailed characterisation will be required
if the waste management process results in segregation or separation of radionuclides
within the waste.
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Lastly, the people who apply the radionuclide fingerprint will not necessarily be the people
who developed it. Hence, justification of the fingerprint and information on its variability
needs to be attached to the fingerprint to ensure it is correctly applied.
Characterising heterogeneity
A homogeneous waste is one in which constituents or properties are constant throughout
the waste. In contrast, a heterogeneous waste is one in which constituents or properties varyfrom place to place. Almost all waste streams exhibit some level of heterogeneity, which
can affect characteristics such as radionuclide activities, concentrations of hazardous
chemical substances, material composition and physical, chemical or biological properties.
Two types of heterogeneous materials can be considered. In the first case, the value (suchas the concentration of a contaminant) is randomly distributed throughout the material. In
statistical terms, this is described as a “spatially uniform” distribution: the statisticalproperties of the material are the same throughout the volume of material. In the second
case, the statistical property varies from place to place within the volume of material. In
statistical terms, this is described as a “spatially variable” distribution. Waste generally, but
not always, tends towards spatial uniformity once they have been retrieved, treated (ifappropriate) and packaged. The greatest spatial variability is generally associated with
waste that is characterised in situ.
As waste heterogeneity becomes greater, it becomes more challenging to design and justify
a characterisation approach that aims to achieve an unbiased estimate of waste
characteristics. In the case of a waste exhibiting spatial variability, it may be necessary toconsider geostatistical approaches to data analysis. Geostatistics will also provide guidanceon designing sampling and surveying plans.
Uncertainties in waste characteristics become greater as waste heterogeneity becomes
greater. At some level of uncertainty, it may be appropriate to segregate the waste toproduce two wastes that are more homogeneous, and which can be characterised with ahigher level of confidence. Inappropriate estimates of waste characteristics may be derived
if heterogeneity is not considered in an appropriate manner. This can affect both bestestimate and cautious estimates.
For large volumes of lower activity UL waste, gamma surveys (for example ofcontaminated land or contaminated facilities) are used to determine the distribution ofradionuclides at and close to the material surface. Commonly, radionuclide distribution is
spatially variable, and hotspots are present. In these circumstances, gamma measurements
can be used to inform any targeted sampling undertaken and/or early decisions on
segregation of the waste during recovery. Gamma measurements are also widely used todetermine the activity of waste produced by demolition or excavation activities; this isachieved through placing the waste into bags (typically 1m3 in volume) and scanning eachbag to calculate (with the use of the radionuclide fingerprint) the average radionuclide
activities and uncertainties in the average. From these measurements, the heterogeneity of
the waste produced can be determined and estimates of the total activity of wasteconsignments produced. In the planning process, a decision will have been made on activity
concentrations of waste that can be left in situ (i.e. not removed for storage or disposal
elsewhere). Gammas surveys, coupled with sampling and analysis, are used to characterise
the site or facility after the lower activity UL waste has been removed, and to demonstrate(validate) that the end point of clean-up has been achieved.

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Step 6: Produce the documentation for waste characterisation
This step recommends the production of a series of documents.
•

Quality Assurance Programme (QAP), which describes the management
arrangements and procedures to be applied to ensure that the characterisation
campaign meets the characterisation objectives.

•

Characterisation Plan (CP), or equivalent document, which presents and justifies
the proposed approach to waste characterisation. The CP is a key QA document,
and it should provide the rationale for the chosen characterisation approach. For
example, are measurements intended to estimate representative properties of thewaste or to characterise an aspect, such as a contaminant hotspot?

•

Data Acquisition Plan (DAP), or equivalent document, which specifies the dataacquisition activities. These might include: the specification of nucleonics
calculations, the numbers, types and locations of new samples and the analyticalschedule, including the required limits of detection and measurement accuracy. The
DAP is a lower level document than the CP and should be linked to it.

•

Characterisation Report (CR), or equivalent document, which presents andinterprets the results of the waste characterisation campaign.

It is good practice to produce a Quality Plan (QP) to demonstrate that all activities havebeen undertaken in accordance with the requirements of the QAP. At the end of eachactivity, the relevant line in the QP is signed off to confirm the activity has been completed
in line with the controlling document; details of any documents produced from the activity
are added. At the end of the waste characterisation process, all activities are signed off. At
this stage, the responsible manager will approve the completed QP to confirm that all work
has been successfully completed.
It is important to record and manage the new information collected so that it can be usedlater for secondary purposes. To enable this, it is essential to record the basis for collecting
the information.

References
Department for Environment, Food and Rural Affairs (2007), “Policy for the Long Term Management of
Solid Radioactive Waste in the United Kingdom”, London,
https://assets.publishing.service.gov.uk/government/uploads/system/uploads/attachment_data/file/25
4393/Low_level_waste_policy.pdf
Environment Agency (2020), “Guidance: RSR Permits for nuclear licensed sites: how to comply”,
www.gov.uk/government/publications/rsr-permits-for-nuclear-licensed-sites-how-to-comply/rsrpermits-for-nuclear-licensed-sites-how-to-comply
Environment Agency et al. (2018), “Management of radioactive waste from the decommissioning of
nuclear sites: guidance on the requirements for release from radioactive substances regulation”,
www.gov.uk/government/publications/decommissioning-of-nuclear-sites-and-release-fromregulation.
Environment Agency, Scottish Environment Protection Agency, Northern Ireland Environment Agency(2009), “Near-Surface Disposal Facilities on Land for Solid Radioactive Wastes: Guidance on
Requirements for Authorisation”, https://assets.publishing.service.gov.uk/government/uploads/
system/uploads/attachment_data/file/296507/geho0209bpjl-e-e.pdf
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130 | NEA/RWM/R(2020)2
NDA (2019), “UK Good Practice Guide on Solid Radioactive Waste Characterisation”, Nuclear
Decommissioning Authority Report, NDA015834.
NDA (2014), “Geological disposal: An overview of the generic Disposal System Safety Case”, Nuclear
Decommissioning Authority, https://rwm.nda.gov.uk/publication/geological-disposal-an-overviewof-the-generic-disposal-system-safety-case-december-2010/?download
SDF (2017), “The UK Industry Guide To: Clearance and Radiological Sentencing: Principles, Process
and Practices”, produced by the Clearance and Exemptions Working Group and published on behalfof the Nuclear Industry Safety Directors Forum,
www.nuclearinst.com/write/MediaUploads/SDF%20documents/CEWG/Clearance_and_Exemption_GPG_2.01.pdf
US EPA (2006), “Guidance of Systematic Planning Using the Data Quality Objectives Process”,
EPA QA/G-4, US Environmental Protection Agency, www.epa.gov/sites/production/files/documents/
guidance_systematic_planning_dqo_process.pdf.

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Chapter 4. Challenges and lessons learnt derived from the case studies
The case studies from France, Russia, Ukraine and the United Kingdom have demonstrated
that the characterisation of UL wastes presents numerous challenges both from theperspective of sampling and analysis through to making decisions on retrieval, treatment,
storage and disposal. This chapter summarises the similarities between the methods usedin each of the case studies with those discussed in Chapter 2 and highlights some of the key
challenges and lessons learnt.

Toolbox of common characterisation approaches applied in each of the case studies
Each of the case studies was reviewed to determine which part of the methodologies
summarised in Chapter 2 were used. In the case studies from France and the United
Kingdom there is evidence that several aspects of the characterisation toolboxes describedin Chapter 2 have been used. In the case studies from Japan, Russia and Ukraine there has
been some application of the methodologies. There are many reasons why thesemethodologies have not been applied in these case studies including:•

Aspects of the methodologies described in Chapter 2 are not appropriate or willneed to be adapted/enhanced. This is because some of the case studies deal withcomplex issues associated with the characteristics of the waste (radiological andhazardous chemical properties) that are not covered in the methodologies. Also, thelack of representativeness due to heterogeneous nature of the waste as well as thedifficulty in sampling from high dose regions presents some novel challenges.

•

The legislative and regulatory framework is still in development or it is not fully
designed to deal with UL wastes.

•

There are limited resources, lack of programme management and scientifictechnical expertise in dealing with UL waste. This includes the analytical facilities
and techniques needed to determine the radionuclide and non-radionuclide contentsof the waste. Several case studies described the work that is currently being
progressed to address these challenges.

•

The necessary enabling infrastructure for treatment, storage and disposal are notcurrently available. This includes no waste acceptance criteria, treatment
technologies or storage/disposal routes.

These barriers are not new, having been echoed in the survey conducted by the IAEA
in 2012 (EUR, 2016).
The potential enhancements and adaptations of the methodologies described in Chapter 2
are explored further in Chapter 5. Some of this information has been based on the
preliminary findings from the R&D work being carried out internationally and specificallyat the Fukushima site.

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Radioactive waste management framework
The original policies and regulations that were developed for radioactive waste
management were mainly derived to protect the public and workers from the dangers ofionising radiation, and latterly, also to protect environment. However, in most countries,
the overall radioactive waste management framework has evolved to include for example,
new waste classification schemes, definitions for interim/end-states, waste package
specifications and criteria for disposal site selection. These areas continue to be developed
to provide a framework supported by the right technical and analytical infrastructure toenable effective and efficient characterisation and ultimately management of radioactive
waste. The case studies demonstrate that countries are still in various stages of developing
the framework for UL waste resulting in some deferring decisions until this has been
resolved. A decision to delay may result in risks increasing over time so it is essential thatthe regulatory framework is developed so that the best management options can be selected.

Waste classification and categorisation
In the case studies, countries were shown to be in various stages of developing or
implementing a waste classification scheme. Common practice shows that schemes for
waste classification and/or categorisation reflect the nationally relevant factors, particularly
for the waste that is expected to be produced from normal operations. The classification ofradioactive waste can be developed to consider different requirements such as operationalor long-term safety, on the processes of waste formation or the storage/disposal facility
specification (IAEA, 2009a). In this case, the development of the scheme should consider
all aspects of the waste management strategy, including communication with stakeholders,and not just the final disposal route.
The case studies highlight that some UL wastes do not naturally fall into one of the wellknown classification schemes (IAEA, 2009a). Most waste categorisation/classification
schemes developed in the past were (correctly) devised to tackle technical issues and werenecessarily expressed in technical terms. They do not readily indicate the scale of hazard
associated with the waste and therefore do not help to explain to stakeholders thesignificance of the hazards (NEA, 2016).
Any new proposed scheme can benefit by considering relevant good practice and lessons
learnt at sites of previous major accidents and legacy situations. The scheme shouldincorporate all waste arisings and include classification of those wastes which do not need
to be managed or regulated as radioactive waste, i.e. account for hazardous features so thatany planning, regulatory or safety management aspects can be appropriately considered. If
the waste classification scheme is revised, it will be beneficial to consult relevant
stakeholders and record the justification behind the decision for accountability purposes.
For UL wastes, there might be an opportunity to revise the classification scheme to include
a component which relates the hazard of each class of waste to another readily
understandable or commonly encountered hazard. Ideally this would include considerationof chemical as well as radiological hazard and a proportionate approach would then be
applied to manage the risk (IAEA, 2013). Where wastes do not readily fit into an existing
waste category it will also be important to ensure that sufficient characterisation data iscollected to enable waste to be classified by alternative categories if this becomes necessary.
This is important so that all stakeholders including regulators and members of the public
have confidence in the management techniques that are applied based on representative
sampling.
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The key drivers in developing a strategy for management of UL wastes are understandingthat there are potentially large amounts of waste that will need to be managed as radioactive
waste (probably as VLLW or LLW, or potentially clearance waste), and identification of
the generally much smaller amount of waste that is acutely and highly hazardous (both interms of radiation and other hazards). In respect to the contaminated material which would
typically be classed as VLLW or short-lived LLW there is a question around whether itshould be allowed to remain in situ provided it is supported by a robust safety andenvironmental assessment and is identified as the preferred management option. This might
be supported by measuring specific features of the waste and require innovative methods
to be applied.
Under certain conditions, restricted site release based on specific land use may be acceptedby regulatory authorities, particularly if the cost of subsurface remediation is prohibitive.
In this instance, the radiological characterisation needs to inform the process of devising
long-term monitoring strategies (NEA, 2013). In the United Kingdom, the environment
agencies released a guidance document for operators that describes what they need to dowhen carrying out decommissioning and clean-up activities. The operators must prepare awaste management plan and a site-wide environmental safety case. The regulators have set
standards for public and environmental protection to be consistent with international anddomestic law, guidelines and policies. As part of the guidance, an operator may apply for
authorisation to dispose of radioactive waste, such as a buried object or structure, by leaving
it permanently in situ (Environment Agency et al., 2018).

Waste acceptance criteria
The WAC in most countries have, as is to be expected, been designed for conventional
treatment, transport, storage and disposal which may not be applicable or convenient for
UL wastes. The case studies have highlighted that the formation of these radioactive wastes
means that quite often they are in a physical or chemical form that has not been encountered
before and may contain materials that are hazardous. Also, some of these wastes containnon-hazardous chemicals and/or biological material that does not fit current waste
treatment, storage and disposal WAC. It is therefore necessary to consider what is needed
to know about the waste in each of these areas and how that information needs to beevaluated.
It is important to clarify at an early stage who is responsible for the development of the
WAC for storage and disposal and how this will be developed so that characterisation
information can best inform this process. Without this information it is difficult to specify
the characterisation programme which might lead to too much or too little information
being gathered leading to the inefficient use of resources. A degree of iteration in theprocess is to be expected.
Characterisation, sorting, treatment and conditioning of waste should be performed
considering WAC, but also, the WAC should be defined to support a safe and effective
waste management strategy so that it can be effectively regulated. In some countries, theWAC for LLW disposal sites has been reviewed in recognition of the fact that it would be
beneficial to provide more flexibility of disposal of certain types of waste but still meetwith the requirements of the environmental safety case. This has been achieved throughgood interactions between the regulators and policy makers, operators and waste producers.
In 2015, the US Nuclear Regulation Commission issued revised guidance on the constraintsof blending LLW as well as the mathematical averaging of radioactivity concentrations(NRC, 2015). Also, in the United Kingdom following the application for an environmental
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permit by the Low Level Waste Repository for continued disposal at the site in 2013,
significant changes were made to the WAC to incorporate specific restrictions required by
the regulator on some non-radiological components, for example, chemical complexing or
chelating agents, ion exchange, and biological materials (LLW Repository Limited, 2016).
In the case studies, in the absence of a geological disposal facility (GDF), further
management (including retrieval and interim storage) of HLW and ILW has been deferred
resulting in characterisation being focused on monitoring any chemical and physicalchanges to the radioactive waste. If a permanent disposal option is not yet available and
WAC do not exist for the disposal of the expected waste forms, it is still possible to retrieve
the waste and condition it for safe storage while maintaining reconditioning flexibility inaccordance with future WAC. This is particularly important for sites where the changing
state of the waste might preclude treatment, conditioning and disposal options. In the
United Kingdom, in the absence of WAC for a GDF, an assessment process has been
defined for packaging specifications so waste can be converted into a passively safe
disposal form that meet the needs for storage, transport and disposal (NDA, 2010). These
specifications define the boundary features and performance requirement of the waste
package based on the “overview of the generic disposal system safety case” (NDA, 2016).

Defining end-states
Where possible “end-states” and words used to describe them should be clearly defined in
respect to the physical state and any controls required to protect people from any residual
contamination, so that the right characterisation objective is selected. By establishing andimplementing clear definitions of end-states it makes it clear to all stakeholders what you
are trying to achieve. At the same time, it is beneficial to adopt an iterative and adaptive
approach to enable progress with the characterisation work, when the end-state will not be
achieved for decades and/or the starting assumptions are uncertain (DSA, 2020). Thisapproach is explored in more detail in Chapter 5.
For large, complex sites, it is likely that this will involve a staged process to get to thepreferred end-state. This should be supported by a robust safety case which continues to be
developed through staged implementation. It is useful to include these stages in a “roadmap” which provides details of the characterisation objectives and the execution time-line.
This is extremely useful when discussing the objectives and plan with all the various
stakeholders. In some circumstances, for example, it might not be practicable to allow
unrestricted access to sites and there might need to be institutional controls applied. In these
instances, it is essential that the decision-making is understood by all stakeholders and thatintergenerational equity is appropriately considered (IAEA, 2017).
In the case studies it was highlighted that some of the terms that are being used to describe
the preferred end-state have yet to be defined resulting in decisions being deferred. This
may limit the selection of future characterisation tasks and may also lead tomisinterpretation of definitions by stakeholders. It has been illustrated that having effective
dialogue with stakeholders, especially those directly affected by remediation/clean-up
decisions, provides valuable input into selecting and supporting achievable end-states(DSA, 2020). It is therefore important to start the dialogue with stakeholders on thedefinition of end-states as early as possible.

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Defining priorities on complex sites and safety analysis
The case studies illustrate that there are many complex sites with various types of UL
wastes that need to be managed, many of which are novel in formation and composition.
Safety analysis, radiological and environmental impact assessments are necessary tosupport the identification of priorities, feasible management options and select preferred
options from available alternatives. They should also provide justification for the WAC.
This prioritisation needs to be technically underpinned but at the same time should be
informed by stakeholder engagement.
There are many challenges on what should be prioritised on sites. If the focus is on
reduction of high hazards or risks, then balanced thinking is needed: it may be better tomanage the acute risk first and then worry about the longer-term management later. At thesame time, actions taken now should, if possible, not prejudice future management options.
These analyses must, where possible, be based on existing regulations and regulatory
guidance. Only in exceptional circumstances and based on a safety case that demonstrates
compliance with the safety basis of the applicable regulation(s) should exemption fromthese criteria be permitted. In such circumstances there may be need for adaptive approachto the application of the regulations.
Thus, the design and content of these analyses should be specific to the purposes of theassessments, including the interest of the intended audience.

Characterisation project plan including sampling and analysis
The characterisation plan should set out the detailed sampling, monitoring and analyticalrequirements to provide enough information to meet with the characterisation objective(s).
All stakeholders should be involved in developing the characterisation plan from thosedefining the sampling strategy through to the laboratory carrying out the analysis. This isparticularly important when there are lots of different organisations/contractors involved incarrying out each of the tasks. This co-ordination should ensure that there are no surprises
during the execution of each of the tasks and that resources are used efficiently.
The principal challenge for characterising UL wastes is that less is known about its
provenance and composition than for other radioactive wastes, which means additional
characterisation is likely to be needed. There are also fewer opportunities to infer thecharacteristics of UL wastes from reviewing existing information, understanding ofprovenance or through comparison with similar waste elsewhere. Also, the case studies
have demonstrated that the sampling and analysis may be needed for multiple waste
streams that vary from VLLW all the way up to HLW. This needs careful planningaccompanied by a conservative safety assurance process to manage all the different hazards.
Dismantling of unstable building structures may require more detailed characterisation
prior to commencement, both in terms of radiological hazards but also physical safety. In
the case studies it was noted that there are significant hazards associated with highly active
fuel and other radioactive particles being distributed unevenly across the site. Also,
penetration of surface contamination into the body of materials is likely to be different fromthat found in conventional decommissioning.
The sampling strategy should, where practical, be designed with all the characterisation
objectives in mind. This is particularly important for legacy and post-accident sites where
you want to limit the time taking the samples due to high EDR and/or hazardous
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physical/chemical environments. The sampling strategy needs to consider theheterogeneity of the waste (including the process of formation) and stratification ofradiological and other hazardous/non-hazardous properties throughout the profile of thewaste. Common approaches used to define sampling and measurement methodology suchas judgement and probabilistic approaches requires a reasonable amount of priorknowledge. Therefore, when considering UL wastes where there is little or no informationon the provenance, a composite or multi-implemental sampling approach may need to beconsidered (NEA, 2017). In Japan, a stochastic sampling method is being developed
applying Bayesian statistics which is likely to be more suited to some of the challenges
presented on the site.
The case studies have highlighted that there are several challenges associated with
developing a representative sample strategy. In the Ukrainian case study, the minimum
number of containers that needed to be characterised were based on the process history of
the plant and then calculated using statistics based on the accepted confidence probability.
This demonstrates that legacy sites where some documentation is sometimes available, andthe waste is relatively homogenous, it is possible to apply this type of approach. However,this approach does not lend itself to dealing with heterogeneous waste. The only practicalmeasure that can be taken currently is to pre-treat the waste such as dividing the waste into
different streams based on material composition, before defining a representative sampling
method. However, a balance needs to be struck between the cost of segregating andcharacterising many homogeneous waste streams against a smaller number of lesshomogeneous waste streams. The level of representativeness required to inform decisionmaking needs to link back to the characterisation objectives. Other approaches are
discussed in Chapter 5.
There are also challenges with developing appropriate methods for measuring large
volumes of radioactive waste for non-radiological properties that might not meet theconditioning or disposal WAC. A method of estimation is being designed to achieve thisin Japan.
It might be necessary to use different tools or robotics to sample the radioactive waste dueto the physical and EDR hazard. This may require the adaptation of existing tools or thedevelopment of new ones. Using the wrong tool can create all manner of problems
including delays to project delivery, increased costs and unnecessary exposure risks. It isalso worth considering that the equipment and tools used to sample the waste may at somepoint in the future become waste. Therefore, when specifying what tools and equipment toselect for the project, it might be important that they can be decontaminated to operate onother parts of the site or other sites.
It should also be clear within the schedule how samples will be preserved and whether there
are any pre-analysis techniques that need to be completed. A lesson learnt in one of thecase studies highlighted the need to have laboratory equipment available to pyrolyze
samples for H-3 and C-14 analysis. This demonstrates that it is beneficial to haverepresentatives from all disciplines on the planning team.
It is important that all the characteristics of the samples taken are documented and where
there are large numbers of samples to be taken, logged onto a centralised database. This
information can then start to be used in conjunction with CCTV and geostatisticalmodelling to identify the different zones of radiological and non-radiologicalcontamination.

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The analytical schedule should consider each of the characterisation objectives and where
possible, results can be used to inform future decisions such as treatment and storage
options. In the case studies, it was clear, that (short-lived and naturally occurring)radionuclides were being analysed that did not seem to present a barrier to any of thetreatment, storage or disposal options. It is worth revisiting why these decisions were taken
and determine whether the characterisation objectives were clear to all stakeholders
involved in the project or whether there were other reasons that need to be explored in moredetail.
The case studies have illustrated that some radioactive waste has DTM radionuclides. It is
important to determine early on when planning the characterisation project whether there
is suitably qualified experienced resource available to carry out the radiological and nonradiological analysis. It is also equally important to establish whether the laboratory(ies)have the right equipment to complete the analysis as most of them will be used to analysing
conventional waste streams. In one of the case studies, several laboratories were used tocarry out the analysis on the waste samples due to the large variety of radiological isotopes
that needed to be measured.
For some radioactive waste, this will also mean the development of new analyticaltechniques and equipment that can operate in high radiation fields which may take time. It
will be beneficial to continue to share knowledge and information internationally on thedevelopment of new analytical techniques. In addition, new laboratories can take lessons
learnt from facilities (especially those handling ILW and HLW) on how to maximise
operational efficiency as well as how to adequately consider decommissioning during thedesign stage to minimise future costs

Nuclide Vectors (fingerprints)
As discussed in several of the case studies the development of fingerprints (nuclide vectors)
based on the radiological composition of waste is useful when developing the clearance
and waste disposition process. However, the application of a fingerprints to UL waste canbe particularly challenging due to uncertainties in their composition. Where there is large
variability in the waste stream the uncertainty in the fingerprint could vary significantly. It
is therefore essential to determine the range of radionuclides that you need to measure anddefine the level of accuracy with which they require to be reported to meet thecharacterisation objective.
In the Ukrainian case study regarding the characterisation of the saline bitumen compoundvat residue, a combination of scaling factors and regression of logarithms analysis wereused to define the fingerprint. Several countries use regression of logarithms analysis when
there is a non-linear relationship between the key nuclide and the DTM nuclide. This is a
useful approach to accurately model complex, non-linear relationships between theradionuclides (IAEA, 2009b).
It is recommended that fingerprints are reviewed regularly i.e. as a minimum they should
be validated annually so that they continue to meet with the characterisation objectives andconsider radioactive decay and possible ingrowth of radioactive species. If, however, there
is any suspicion that the nature of the radioactive material has changed significantly more
frequent reviews will be required. Gamma spectrometry can be used to review and compare
the relative activities of gamma-emitting species. Also, it is possible to get an indicationthat there is more drift than acceptable by using Health Physics instruments to measure therelative activities of alpha or beta-emitting radionuclides (SDF, 2017).
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The case studies highlighted, as discussed above, that a fingerprint could be derived forsome of the homogeneous waste streams where there was some information on theformation process. However, a large volume of UL waste is heterogeneous which meansthat it does not fit with the normal application of a radioactive waste fingerprint. It istherefore important to understand the acceptable variability in the fingerprint and whether
it is appropriate to use “best-estimate” proportions of radionuclides or a cautious estimateof the proportions based on those that effect waste management decisions.

Characterisation reporting and review
One of the case studies highlighted the need to establish a framework and guidance forindependent characterisation review by a third party. In several countries, following aninternal review by a dedicated team, independent of the delivery team, the characterisation
report and supporting documentation is then reviewed or assessed by the regulatoryorganisation(s).

Retrieval, conditioning and packaging radioactive waste
The objective of the initial characterisation is to gather enough information about thephysical state of the facility as well as the radioactive and non-radioactive properties of thewaste to develop a safe, optimised retrieval plan. Safety and environmental protection
factors need to be considered in detail when developing a plan for the retrieval of UL waste
as there are likely to be some unknowns even after the initial characterisation phase. Thismeans that a thorough hazard and risk assessment needs to be conducted with a level offlexibility and conservatism built into the plan.
It is important to evaluate which equipment is most suitable for the job based on
information including the class and types of waste to be retrieved (geometry, size, mass,
LLW, ILW or HLW), the radiological hazard as well as other potentially hazardous
materials (dangerous chemicals). There also needs to be consideration of the unknowns andthe need for continued iterative characterisation to inform the retrieval work and possible
conditioning and packaging of the waste (IAEA, 2007a).
Inadequate initial characterisation can be very costly both from a safety and financial
perspective. It is useful to gather information through CCTV before retrieving UL waste todetermine whether it is appropriate to use conventional retrieval equipment. The
characteristics of the waste or the operating environment might require conventional
retrieval equipment to be adapted or completely redesigned. In one of the case studies it
was found that by using conventional decommissioning equipment (excavator withclamshell grab) this disturbed the degraded radioactive waste packages during the retrieval
operation resulting in an airborne dust hazard. It is important to spend adequate time on theinitial characterisation of facilities where there is limited historical records so that the mostappropriate retrieval equipment can be selected for the project supported by a robust safetyassessment. It is advisable that some flexibility is built into the project plan to enablechanges to be made part way through especially when dealing with UL waste. This will
ensure that hazards are satisfactorily managed and that costs are minimised.
The initial characterisation is also important to support decisions on any conditioning thatmight be required, the monitoring needed to be performed to meet the WAC as well as the
specification of the storage/disposal package. In one of the case studies, when some of theradioactive waste was retrieved it was found to be too big to fit into the designated storage

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container. It is therefore important that there is a degree of flexibility in the retrieval planto account for unexpected situations.
The packaging/containers that are going to be used to store and/or dispose of the waste
need to be considered in detail prior to retrieving the waste so that they meet with the WACof the proposed conditioning, storage or disposal requirements. Where possible it isadvisable to monitor the waste prior to it being bulked into one container as it can beproblematic after packing. Also, due to high background levels in the facility, monitoring
equipment might need to be located in another facility or be assembled with additionalshielding to enable waste to be cleared. In one of the case studies, the high backgroundlevel prevented the clearance of radioactive waste which led to higher project costs.

Radioactive waste inventory
The case studies have demonstrated that the task of identifying the radioactive waste
inventory is particularly difficult for legacy and post-accident sites. The identification of
the sources of radioactive waste including the estimation of the amounts of existing wastethat has been characterised and those which have yet to be characterised presents problems
for its overall waste management. In Japan, a mathematical model to help to estimate highvolume waste inventories is in development.

Research and development
There are examples provided in the case studies illustrating how important R&D is insupporting UL waste characterisation and decision-making. Several countries have welldeveloped R&D programmes which focus on areas such as waste characterisation, waste
minimisation, treatment and conditioning, storage and disposal (NEA, 2016). There aremany challenges when dealing with such a wide range of complexities and uncertainties
that UL waste presents which requires new sampling and analytical methodologies
supported by innovative and sometimes novel engineering. It will be important to share theoutputs from the R&D projects internationally to support other countries who areembarking on waste characterisation projects.

Conclusion
UL waste may not fit into the current waste classification/categorisation system or the
existing treatment, storage or disposal WAC resulting in characterisation objectives
concentrating on managing the steady state. It is evident in the case studies that there arestill barriers present that hinder progress towards characterising and managing UL waste.
Many of these can be linked to the absence of national policy and associated legislative andregulatory framework, a lack of clear responsibilities and/or the lack of definition of endstates.
There is also insufficient access to appropriate analytical facilities, or those that areavailable do not have the expertise to analyse novel waste samples. In addition, thecharacteristics of these UL waste require the development of new analytical techniques
which require suitably qualified and trained resource as well as funding.
It will be important that the international community continue to share experiences and bestpractice on dealing with the characterisation challenge so that progress can be made onthese complex sites. This collaboration will also inform future research and development.

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References
DSA (2020), “Regulatory Framework of Decommissioning, Legacy Sites and Wastes from Recognition
to Resolution: Building Optimization into the Process Report of an international workshop”, Tromsø,
29 October-1 November 2019, Norwegian Radiation and Nuclear Safety Authority and the Nuclear
Energy Agency, May 2020, number 5, www.DSA-rapport 05-2020 Regulatory Framework
komplett.pdf
Environment Agency et al. (2018), “Management of radioactive waste from decommissioning of nuclear
sites: Guidance on Requirements for Release from Radioactive Substances Regulation”, Version 1.0,
www.sepa.org.uk/media/365893/2018-07-17-grr-publication-v1-0.pdf
EUR (2016), “Advancing Implementation of Nuclear Decommissioning and Environmental Remediation
Programmes: Policy Support Document”, European Commission’s Joint Research Centre, preparedin collaboration with the International Atomic Energy Agency, EUR 27902 EN,
https://publications.jrc.ec.europa.eu/repository/bitstream/JRC101139/cider_policy%20support%20do
c_online.pdf
IAEA (2017), “Advancing the Global Implementation of Decommissioning and Environmental
Remediation Programmes, Proceedings of an International Conference, Madrid, Spain, 23-27 May
2016”, International Atomic Energy Agency, Vienna, www-pub.iaea.org/MTCD/Publications/PDF/
Pub1759_web.pdf
IAEA (2013), “Model Regulations for the Use of Radiation Sources and for the Management of the
Associated Radioactive Waste: Supplement to IAEA Safety Standards Series No. GS-G-1.5”, IAEATECDOC-1732, International Atomic Energy Agency, Vienna, www-pub.iaea.org/MTCD/
Publications/PDF/TE-1732_web.pdf
IAEA (2007a), “Retrieval and Conditioning of Solid Radioactive Waste from Old Facilities”, Technical
Reports Series no. 456, International Atomic Energy Agency, Vienna,
www-pub.iaea.org/MTCD/Publications/PDF/TRS456_web.pdf
IAEA (2007b), “Strategy and Methodology for Radioactive Waste Characterization”, IAEA-TECDOC1537, International Atomic Energy Agency, Vienna, www-pub.iaea.org/MTCD/publications/
PDF/te_1537_web.pdf
IAEA (2009a), “Classification of Radioactive Waste: General Safety Guide No. GSG 1”, International
Atomic Energy Agency, Vienna, www-pub.iaea.org/MTCD/Publications/PDF/Pub1419_web.pdf
IAEA (2009b), “Determination and Use of Scaling Factors for Waste Characterization in Nuclear Power
Plants”, IAEA Nuclear Energy Series No. NW-T-1.18, International Atomic Energy Agency, Vienna,
www-pub.iaea.org/MTCD/Publications/PDF/Pub1363_web.pdf
LLW Repository Ltd. (2016), “Waste Acceptance Criteria – Low Level Waste Disposal”, Waste Services
Contract, WSC-WAC-LOW Version 5.0, Issue 1.
NDA (2016), “Geological Disposal: Overview of the generic Disposal System Safety Case”, NDA
Report no. DSSC/101/01, Nuclear Decommissioning Authority
https://assets.publishing.service.gov.uk/government/uploads/system/uploads/attachment_data/file/63
4857/NDA_Report_no_DSSC-101-01_-_Geological_Disposal__Overview_of_the_generic_Disposal_System_Safety_Case.pdf
NDA (2010), “Geological Disposal. An overview of the generic Disposal System Safety Case”, NDA
Report no. NDA/RWMD/010, Nuclear Decommissioning Authority,
https://rwm.nda.gov.uk/publication/geological-disposal-an-overview-of-the-generic-disposal-systemsafety-case-december-2010/?download
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 141

NEA (2017), Radiological Characterisation from a Waste and Materials End-state Perspective:
Practices and Experience”, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_15056
SDF (2017), “The UK Nuclear Industry Guide To: Clearance and Radiological Sentencing: Principles,
Process and Practices”, produced by the Clearance and Exemptions Working Group and published onbehalf of the Nuclear Industry Safety Directors Forum,
www.nuclearinst.com/write/MediaUploads/SDF%20documents/CEWG/Clearance_and_Exemption_
GPG_2.01.pdf
US Nuclear Regulatory Commission (2015), “Concentration Averaging and Encapsulation Branch
Technical Position, Revision 1”, Volume 1, US Nuclear Regulatory Commission, Office of Nuclear
Material Safety and Safeguards, www.nrc.gov/docs/ML1225/ML12254B065.pdf

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Chapter 5. Characterisation methodology and adaptations and enhancements
to address UL waste challenges
UL radioactive waste can consist of high activity components and particles from fuelcladding, fuel residues, sludges and flocs and contaminated or activated metal. It can also
consist of large volumes of contaminated land, soil, vegetation, structures and debris. The
focus of this chapter will be on lower activity solid UL waste and how characterisation
methodology can be adapted and enhanced to address the challenges this waste poses.
It is important to understand both the radiological properties (such as radionuclide activity
concentrations, total activity inventories and external dose rates) and hazardous nonradioactive properties (the concentrations of hazardous chemical compounds). The
hazardous non-radiological properties of the waste can be flammable, explosive, reactive
and toxic to people and the environment. Therefore, it is important to have information on
all the properties to enable correct classification and sentencing of waste and to enable safe
treatment, packaging, transport, storage and disposal options to be developed. Hazardous
substances can therefore be a significant factor when developing waste management
options for UL waste.
The physical, chemical and biological characteristics will determine how the waste will
behave when it is processed, transported, stored or disposed. For example, an understandingof the physical size, material components and flow characteristics/viscosity of raw waste isneeded to enable the design of a waste retrieval process. Chemical and biological processes
can result in significant changes to the waste and any encapsulant which can then impact
on the performance of the waste package.
This chapter focuses on the key differences between UL waste and conventional waste andhow the existing characterisation methodologies can be adapted or enhanced to cater forthis. To achieve this and enable decisions to be taken, it is also important to have the rightlegislative and regulatory framework supported by technical guidance.

Site prioritisation
The case studies support the conclusion that post-accident and legacy sites are each uniqueand require novel approaches to try and solve the technical or programmatic challenges
faced. The use of screening analyses can be useful in helping to prioritise sites and identify
where additional effort may be needed. To inform this additional characterisation is likely
to be needed to develop a site conceptual model taking account of the source-pathwayreceptor model. The main considerations when developing a conceptual model are:•understanding the state of each contaminant and potential mobility considering soil
chemistry and the proximity to the water table;•meteorological conditions including precipitation, wind speed and direction,
humidity, and temperature. (This should also consider extreme weather events);•site topographical including all water bodies and their flood extents;•

proximity to humans or wildlife.
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In addition to these conventional contaminated land considerations, it is also important toconsider the risks of not prioritising certain parts of the site such as the:•Physical state of the site where the materials and wastes are being stored. Theremight be a risk that the building/store might collapse.

•

Chemical interactions between different materials resulting in a hazard to humans
or the environment i.e. reactive metals (such as sodium, magnesium andaluminium) and volatile organic compounds.

•

Biological properties such as bird or bat guano which can lead to the generation of
the organic-rich and biologically hazardous waste streams; and the presence of
microbes which can produce various noxious gases.

One of the key issues is understanding the risks across a wide variety of different hazards(DSA, 2020). Independent analysis of all these criteria is important to develop a better
understanding of the risks and ensure the right questions are being asked. This will probably
require involvement from different regulatory bodies and scientific disciplines applying
holistic approaches that consider remediation across the life cycle of the site.
If the focus is on reduction of high hazards or risks, then balanced thinking is needed: it
may be better to manage the acute risk first and then worry about the long-term
management later. At the same time, actions taken now should, if possible, not prejudice
future waste management options. Although it is important to consider all the hazards andrisks discussed, it is also important to think about processing waste that is easy tocharacterise especially where sites have space constraints. A staged or phasedcharacterisation process is needed for any complex site, with further decisions on next step
being taken in the light of the most recent information, e.g. on the wastes (NEA 2019). This,in turn, will support more optimal decision-making, ensuring that limited resources are
spent cost-effectively.

Characterisation objective(s)Any characterisation activity should be aimed at achieving the predetermined objectives
and completing the assigned tasks. Characterisation of UL waste, performed without
clearly formulated objectives and tasks, can be costly and inefficient, as well as lead to
unexpected consequences. In general, characterisation is an iterative process whereby the
results of completed tasks help to formulate future tasks.
A “life cycle approach” to waste characterisation is recommended, as this approach isdesigned so that all the information that may eventually be required on UL waste isconsidered at the outset. A “life cycle approach” does not mean a “just in time” approachor always acquiring only the data needed to achieve the immediate characterisation
objective. The objective of the approach is to optimise the number of characterisation
phases, the timings of these phases and the information required to support each phase.
Information is required on the UL waste for the analysis of technical options for treatment,
transport storage and disposal. Feasible options must meet safety criteria and so theinformation needed must also be sufficient to support the safety analyses to meet with theregulatory requirements. Radionuclides of interest for disposal safety can be different fromthose for operational safety. For example, some radionuclides only make up a small amountof the specific activity so do not present an operational safety risk but may represent a
disposal issue. Also, the radionuclides in the WAC are often among the DTM.

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By understanding a large proportion of the characterisation information requirements at the
project outset for all phases across the life cycle, this can help to optimise resources.
Where the WAC is still being developed, there is an opportunity for radioactive
characterisation of UL waste to inform the process. This will ensure that resources are notwasted measuring short-lived or irrelevant radionuclides. What is irrelevant will depend on
the phase of the project and characterisation objective that needs to be achieved. The key
point is to consider whether the methods used for conventional waste management areappropriate for UL waste and if they are not, what can be done to improve them. There issignificant experience on what is likely to be important as demonstrated in the case studies.
As described in Chapter 2 it is important to take a holistic approach to waste
characterisation so that the non-radiological properties of the waste are considered in thewidest sense covering physical, chemical and biological properties. These properties may
impact on the waste treatment, conditioning, storage and disposal options or decision toleave the waste in situ. Once an understanding of the non-radioactive inventory is available,
it will be important to establish whether the presence of any of these substances limits the
management options and whether all or just some of them need to be measured.
Applying a structured process to define objectives is relevant to UL waste characterisation.
This is likely to be an iterative process as some provisional samples might be needed to get
an understanding of the extent, types and form of contaminants you need to deal with. It isalso likely that the structure will change for each post-accident site of legacy situation.

Historical information
The main difference between UL waste and conventional waste as discussed in Chapter 2
is that the provenance and material characteristics of the waste are less well-known as thewaste generation was unplanned. The waste is likely to contain a mixture of radiologicaland non-radiological hazardous chemicals. There are generally fewer or no equivalent
wastes elsewhere from which the properties of the waste can be inferred, and it is more
likely that techniques for characterisation will not have been developed or be commerciallyavailable.
One of the primary characterisation objectives will be to develop knowledge of thecontamination state i.e. radiological, chemical, biological and mechanical properties. The
first thing to establish, using where practical, conventional characterisation methodologies,
is what information you have on the properties of the material and waste on the site. Thiscan be derived by carrying out a desk-based study by reviewing:•

Operational records of the radiological inventory and location on the site. At anuclear power plant, this is likely to consist of loaded fuel, used fuel and radioactive
waste.

•

Operational information on levels of contamination likely to be present on the site.
In the case of a nuclear power plant this is likely to include the level ofcontamination in the reactor and surrounding circuits (based on neutron fluencecalculations), components and areas inside and outside the building. Some of thisinformation could be gathered from existing decommissioning projects of similar
reactors and from non-destructive and destructive tests of the internals carried out
as part of statutory outages. Similar types of information are likely to be required
for manufacturing, processing and treatment facilities. For temporary radioactive
waste stores/disposal sites, this may include an inventory and maps.

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Records of the hazardous and non-hazardous chemical inventory and the locationon the site.

•

Review the materials inventory derived from the as-built construction of thebuildings/waste stores/disposal facilities on the site. This should consider for
example, asbestos and reactive metals.

•

Operational information on the condition of the plant/facilities. For sites that haveexperienced an accident, this may include the consideration of any mitigation
actions that were taken. Post-accident this will also need to consider thecommencement of decommissioning activities.

All this information needs to be used in conjunction with wider site understanding basedon, for example, computer modelling to start the iterative process and obtain initial ideason the temporal and spatial boundaries of contamination. The next step is to start to plan
the waste characterisation project to develop the sampling and measurement strategy andimplementation plan. As discussed in Chapter 4, it is essential that a robust safetyassessment is performed before any physical work is conducted on the site. Legacy andpost-accident sites present numerous physical hazards due to the state of the facility as wellas radiological and chemical hazards. Safety and environmental protection factors need tobe considered in detail when developing the characterisation plan as there are likely to be
more unknowns compared to dealing with conventional decommissioning. This means, it
is advisable to conduct a thorough hazard and risk assessment with a level of flexibility andconservatism built into the plan.

Site reconnaissance and monitoring
On conventional decommissioning sites it may be possible to use the historical information
to describe: the context of the site, identify areas of concern that would need further
assessment and highlight areas where there were gaps or uncertainties in the information
regarding potential risks or hazards (NEA, 2013). However, on many sites holding UL
waste, it is likely that there will be lots of gaps and uncertainties in the information gathered
due to the lack of data on the formation and properties of the radioactive waste, thisinformation can only be used to start to understand the source term.
To further assist in the delineation of the temporal and spatial boundaries it is useful to
begin with site reconnaissance. This should be based on non-intrusive techniques such as avisual inspection of the whole site as well as the use of non-intrusive remote sensing
techniques. The use of mobile gross alpha, beta and gamma instruments can be used to start
to build a radiological map of the site. These methods were successfully applied in someof the case studies including visual surveys, beta, gamma, direct neutron and magnetic
measurements, and anthropogenic and geobotanical surveys, using a GPS receiver to create
2-D maps of the site. Wide-area gamma scanning surveys have also been exploited on
several sites with the ability to detect particles at 100 mm depth with >95% probability ofsuccess at practical ground speeds (SEPA, 2005).
The results from surveys have also determined that in situ measurement methods are betterat hotspot detection than ex situ laboratory sampling analysis. However, careful choice ofthe spatial model for the conversion of the raw counts is essential for estimation of activity
concentrations over average areas (Rostron, P.D et al., 2012).
This information can then be used to start to plan the sampling strategy and plan for thesite.

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Sampling plan and strategy
As described in Chapter 2, the DQO process and others designed using these principles
provides a systematic approach to data collection which assures that the type, quantity andquality is appropriate for the intended application. It defines the criteria that the survey
design should satisfy, including when and where to perform measurements, the level ofdecision errors for the survey and the number of measurements to perform (NEA, 2013).
The DQA process which was designed to focus on environmental decision-making still
provides a systematic set of steps to carry out a preliminary data review and verification of
statistical assumptions that can be applied to UL waste. Overall, these approaches are useful
to apply to UL waste with some modifications.
The statistical approaches used in DQO which enables the formulation of a sampling plan
are based on frequentist statics (sometimes referred to as frequentist statistical inference)
rather than Bayesian statistical methods or geostatistics. However, these methods can beused although this may involve slight adjustments to the activities within the DQO Process,which should result in equivalent but different outputs (EPA, 2006). The use of frequentist
statistics may not be applicable to UL waste where there is limited information on theprovenance and material properties of the waste, and the processes that led to its generation.
It might be appropriate as identified at Fukushima to apply a stochastic method of analysis
using Bayesian statistics. This statistical method enables the probability of a hypothesis to
be updated as more information becomes available unlike frequentist inference where themodel parameters and hypothesis are fixed (IAEA, 2016). Whichever statistical methods
are used they need to be supported by measurement, assumption and professional
judgement to demonstrate they are sufficiently representative to allow a decision to bemade such as “Does the waste comply with the WAC”?
Uncertainties in waste characteristics becomes greater as waste heterogeneity increases. It
is essential to gain an understanding of the scale and type of heterogeneity in the UL
material and waste as this has implications on the sampling plan. There can be differencesin the composition of particles distributed throughout the waste which can also concentrate
to form hotspots or differences in particle distribution of radionuclides and chemicals
throughout the same materials which are distributed throughout the waste i.e. surface
contamination of concrete. It is important to understand how the activity is distributed as
this affects how exposure can arise. Dose and risk assessment should consider different
possibilities and the results should inform characterisation plan in respect to what degreeof heterogeneity is acceptable. It may therefore be beneficial to segregate the waste during
pre-treatment to simplify and give a higher level of confidence in subsequentcharacterisation and waste management. It is important to recognise that a balance needs
to be struck in the cost of segregating and characterising many homogeneous waste streams
against a smaller number of less homogeneous waste streams.
The methodologies provided in DQO are not designed with the objective of determining
whether radioactive waste meets the WAC for a given treatment or disposal facility. Both
MARSSIM and MARSAME are used principally for making decisions on whether to clear
land, materials and equipment from control under radioactive substances regulation.
However, the principles of DQO and DQA, together with some of the survey
methodologies described in MARSSIM and MARSAME, can be used to achievesystematic and justifiable planning, implementation, assessment and decision-making forcharacterising UL waste. The other consideration is that characterisation is likely to require
determination of the chemical contaminants as well as the physical and chemical properties

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Geostatistics
Experience has shown that UL waste is likely to be heterogeneous and this is likely toincrease due to the presence of more materials or of waste derived from more than one
source or process. Also, the distribution of radionuclides and hazardous chemicals throughthe waste mass is likely to be influenced by the material composition of the waste as wellas the environmental processes that it has been exposed to following formation
i.e. evapotranspiration, filtration. This means that to gather enough information to estimate
the source term, the average activity levels to assist with defining the waste categories anddetermining the location of hot spots it is important to optimise the sampling strategy. One
way to do this is by using geostatistics which considers the spatial continuity of thecontamination to predict it at unsampled locations and to quantify the prediction
uncertainty.
Historical information, in situ (alpha/beta/gamma) measurements and results fromlaboratory analysis of samples can be used in geostatistical models of spatial distributionof activity levels (NEA, 2017). It is possible to integrate the physical, historical and in situ
monitoring measurements to improve the understanding of the main variable which in mostcases will be laboratory data. This approach will also reduce the estimation uncertainty(Desnoyers, 2014).
Geostatistics, can also provide guidance on designing sampling and surveying plans. Thismethod can therefore be used to supplement MARSSIM by providing optimisation ofsampling points as well as proposing the most appropriate statistical tests (Desnoyers,2014).
At Fukushima Daiichi, post-accident mapping using dose rate monitoring figures andgeostatistics identified areas that required further investigation in relation to population
exposure limits. By integrating information on the topography of the area it was possible
to better understand the air dispersion as well as the deposition of the radioactive dust
(Desnoyers, 2014).

Developing radionuclide vectors (fingerprints)
As described in Chapter 2, there are several challenges when developing fingerprints(nuclide vectors) for large volumes of UL solid waste compared to conventional waste.
This is due to the heterogeneous nature of UL waste coupled with the lack of provenancewhich therefore has implications on how to take a representative sample. It is therefore
important to understand the acceptable variability in the fingerprint which will depend on
what it is going to be used for. If the observed variability exceeds this level, segregate thewaste or sub-divide the waste volume and produce individual fingerprints for sub-set of thewaste. Any decisions taken should be justified and documented so that it is appliedcorrectly.
It is beneficial to use both scaling factors and regression of logarithms analysis to define
the fingerprint. This should provide a strong basis to manage both linear and non-linear
relationships between the key radionuclide and the DTM radionuclides. A representative
number of samples should be collected covering the highest activity concentration range.
It is possible to apply a statistical test to determine whether different data populations are
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statistically the same. Ideally there should be enough samples collected to limit theuncertainty to within an acceptable range usually a factor of 10 (IAEA, 2009).
This information can then be used to estimate the radioactive properties of the waste
through inference by:•direct measurement of the key radionuclides and then applying the scaling factor
for DTM radionuclides;•

direct measurement of gross radiation or radioactivity measurement and then by
applying a conversion factor for key radionuclides then the scaling factor for the
DTM radionuclides (NEA, 2017).

Decision-making
A key part in the decision-making process is determining and defining the end-state as
discussed in Chapter 4. An end-state is a set of site-specific conditions that areappropriately protective of people and the environment from a regulatory perspective,
considering the intended, continuing and future use of the site. It must also consider thewider interests of people potentially impacted by the process of achieving the end-state andits future use (NEA, 2019). It is beneficial to apply this process holistically to optimise theprotection of people and the environment addressing all hazards as well as the wide rangeof economic and social factors. This is likely to be an iterative process when the end-state
will not be achieved for decades and/or the starting assumptions are uncertain (DSA, 2020).
The proposed end-state will depend on current legislation, regulations and guidance, safety
and environmental assessment, waste management infrastructure and available funding.
There are several options that should be considered before defining the characterisation
objectives and plan bearing in mind that the initial characterisation options may help define
the feasibility options. As described in the International Commission on RadiologicalProtection (ICRP) principle of justification, “any decision to that alters the radiation
exposure situation should do more good than harm” (ICRP, 2007). Therefore, whenevaluating these options consideration should also be given to:
Leaving the material/waste in situ:•

To make this decision, it might be necessary to review and revise the current
legislative and regulatory framework. It will also be essential to clearly define “endstates” and conduct effective stakeholder engagement.

•

The material/waste will need to meet safety and environmental assessment criteria
and may require continued management including further monitoring andregulatory oversight.

•

The waste may require some form of pre-treatment and/or engineered barriers
installed to prevent migration of radiological or non-radiological chemicals into thesurrounding environment.

Retrieving the waste for interim storage and/or disposal:•

Where there are large volumes of waste, or waste of different categories it will bebeneficial to carry out some form of pre-treatment including: sorting andsegregation, physical or chemical treatment or a combination of all of them. This
will maximise the use of the existing waste infrastructure but could also mean thatnew infrastructure needs to be developed.
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It may be necessary to immobilise the waste so that it meets with existing WAC forinterim storage and/or final disposal. As discussed in the case studies, it might also
be necessary to develop new immobilisation techniques and/or waste packages tocater for both the radiological and non-radiological chemical/biological properties
in the waste. It also may require a review and potential revision of the existingstorage and disposal WAC.

For complex sites, the remediation strategy may consist of a mixture of the options outlined
which could take a considerable amount of time to carry out depending on the current
regulatory framework and waste infrastructure. In some circumstances, a staged approach
might be beneficial where each stage is clearly defined with an achievable delivery
timescale. Also, it is possible that the end-state will change over long periods of time due
to changes in stakeholder preferences, regulatory or policy requirements, or as further
characterisation information becomes apparent (NEA, 2016). It is therefore recommended
that an adaptive approach is used and that the safety case continues to be developed throughthis staged implementation.
It is important that the decision-making framework considers how to accommodate andcommunicate the likelihood of significant uncertainty especially in the early stages ofcharacterisation and remediation planning/implementation to stakeholders (NEA, 2019).
A holistic and proportionate risk assessment approach should be adopted to evaluate theradiological and non-radiological impacts, and this should be communicated in languagethat is understood by all stakeholders.
To develop the decision-making process, it is essential that there is a sufficient regulatory
framework in place and a waste management infrastructure.

Regulatory framework including waste acceptance criteria and categorisation/
classification of waste
The regulatory framework should provide the structure that informs the characterisation
objectives and ultimately the decisions that are taken. As discussed in the case studies andChapter 4, the regulatory framework and guidance on how it should be applied is at various
stages of development for UL waste. It is important that the regulatory framework has
flexibility as it needs to cover a wide variety of conditions and evolving technologies. It isuseful to link remediation/clean-up and decommissioning activities with waste
management programmes (DSA, 2020). It should, as far as reasonably practicable, be
consistent to the approaches already implemented for the management of non-radioactive
contaminants in land and groundwater (NEA, 2016). A goal-setting or performance-based
approach may be more appropriate when managing UL waste (NEA, 2019). However, abalance needs to be struck so that there is enough definition in the regulatory framework toengender confidence from all stakeholders that the appropriate safety (including
environmental) standards are being applied but at the same time it is not too prescriptive sothat the regulator is not taking responsibility for adequate waste characterisation away fromthe operator. To develop an effective legislative and regulatory framework with a structured
R&D programme, it may be beneficial for radiation and nuclear safety regulators to workwith those regulators responsible for physical and chemical hazards (NEA, 2018).
WAC are commonly in place and they work well for planned conventional waste streams
but not always for UL waste. In particular, the application of WAC for conventional
treatment, transport, storage and disposal may not be applicable to UL waste, requiring
variation in methods and revision of safety arrangements and their regulatory supervision
CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

150 | NEA/RWM/R(2020)2(NEA, 2016). Although, the WAC is relatively well established for LLW radionuclide
content, the non-radiological components may be less so.
The case studies have illustrated that there can be high volumes of solid LLW and VLLW,
and there might not be the capacity in the existing waste infrastructure to enable disposal.
This represents a significant issue to any UL waste decommissioning project from both awaste management perspective as well as and probably more importantly a stakeholder
perspective. It is therefore pertinent to consider whether:
1. The volume of radioactive waste can be greatly reduced so that it can be managed
within the existing waste infrastructure through sorting and/or treatment.
2. A new regulatory framework can be developed which enables the waste to be stored
or disposed of in situ on the site while meeting strict requirements (this might mean
implementing long-term stewardship restricting access).
3. Construct a new engineered interim storage site and/or a disposal facility.
Whichever, of these decisions is taken will impact on the characterisation objectives.

Conclusion
Overall, the current characterisation methodologies discussed in Chapter 2 are applicable
to UL waste as they are to conventional waste. However, there are some adaptations andenhancements that can be made to the methodologies to benefit those dealing with UL
waste. Figure 5.1. illustrates adaptations and enhancements that can be made in the case ofcharacterisation of accident waste based on the experience and the case of Fukushima
Daiichi.
Standard methods are, by definition, not likely to be fully applicable to UL wastes. As noted
in Challenges in Nuclear and Radiological Legacy Site Management, Towards a Common
Regulatory Approach (NEA, 2019) every legacy situation has its unique features, and thesame can be said for accidents. Therefore, it is likely that a decision tree like that in Figure
5.1 would need to be specifically designed to address each situation. At the same time, thecase studies show that a phased approach to characterisation may be needed and that
iteration of the process of characterisation may be needed at each decision point in thedecommissioning or remediation programme. It is recommended that the design of thecharacterisation programme is fully integrated with the wider management programme andthat effort is made to engage with all stakeholders to obtain relevant input into that design.
This may complicate actions at the early stages but improve effectiveness in the longer
term. Figure 5.2 illustrates the style of road map and interactions that are likely to be
relevant, based on the above discussion.

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Figure 5.1. A characterisation decision tree for accident waste

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Figure 5.2. Characterisation roadmap

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References
Desnoyers, Y. and D. Dubot (2014), “Characterization of radioactive contamination using geostatistics”,
Nuclear Engineering International, March 2014, www.geovariances.com/wpcontent/uploads/2016/08/nei_characterization_of_radioctive_contamination_using_geostatistics_cea_
geovariances.pdf
DSA (2020), Regulatory Framework of Decommissioning, Legacy Sites and Wastes from Recognition to
Resolution: Building Optimization into the Process, Report of an international workshop, Tromsø, 29
October-1 November 2019, DSA Report May 2020, number 5, Norwegian Radiation and Nuclear
Safety Authority, https://dsa.no/publikasjoner/dsa-rapport-5-2020-regulatory-framework-ofdecommissioning-legacy-sites-and-wastes-from-recognition-to-resolution/DSA-rapport%20052020%20Regulatory%20Framework%20komplett.pdf
EPA (2006), “Guidance on Systematic Planning Using the Data Quality Objectives Process”, EPA
QA/G-4, United States Environmental Protection Agency,
www.epa.gov/sites/production/files/documents/guidance_systematic_planning_dqo_process.pdf
IAEA (2016), “IAEA Safety Glossary: Terminology used in Nuclear Safety”, 2016 Revision,
International Atomic Energy Authority, Vienna, wwwns.iaea.org/downloads/standards/glossary/iaea-safety-glossary-draft-2016.pdf
IAEA (2009), “Determination and Use of Scaling Factors for Waste Characterization in Nuclear Power
Plants”, IAEA Nuclear Energy Series: No. NW-T-1.18, International Atomic Energy Agency,
Vienna, www-pub.iaea.org/MTCD/Publications/PDF/Pub1363_web.pdf
ICRP (2007), “The 2007 Recommendations of the International Commission on RadiologicalProtection”, ICRP Publication 103. Ann. ICRP 37 (2-4), Elsevier,www.icrp.org/publication.asp?id=ICRP%20Publication%20103
NEA (2019), Challenges in Nuclear and Radiological Legacy Site Management, Towards a Common
Regulatory Approach, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_40359
NEA (2018), Preparing for decommissioning during operations and after final shut-down, Radioactive
Waste Management and Decommissioning, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_15058
NEA (2017), Radiological characterisation from a waste and materials end-state perspective, OECD
Publishing, Paris, www.oecd-nea.org/jcms/pl_15056
NEA (2016), Strategic Considerations for the Sustainable Remediation of Nuclear Installations, OECD
Publishing, Paris, www.oecd-nea.org/jcms/pl_14984
NEA (2013), “Radiological Characterisation for Decommissioning of Nuclear Installations”, OECD
Publishing, Paris, www.oecd-nea.org/jcms/pl_19330
Rostron, P.D., M.H. Ramsey and J.A. Heathcote (2012), “Comparison between measurement methods
for the characterisation of radioactively contaminated land”, presented at Radiological
Characterisation for Decommissioning Studsvik, Nyköping, Sweden, 17-19 April 2012, www.oecdnea.org/rwm/wpdd/rcd-workshop/C-3___OH_Comparison_between_ex-situ_and_in-situRostron.pdf.pdf
SEPA (2005), “An evaluation of the sensitivity of the Groundhog evolutionTM bench monitoring system”,
Scottish Environmental Protection Agency, (available on request: www.sepa.org.uk).

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Chapter 6. Conclusion and recommendations on good practiceThis report provides a comprehensive overview of the key aspects, decision-making andactivities that need to be completed for the characterisation of large volumes of legacy andunconventional (UL) waste. It also considers the interrelationships between key issues toprepare and plan for characterisation based on experience and good practice in NEA
member countries.
The key aspects to consider include:•

Management of post-accident and legacy sites should be driven by a clear strategy,
underpinned by a regulatory framework to implement the policy. Characterisationof UL waste may represent a major aspect of this strategy. If the strategy does notexist or it is being updated, then there is an opportunity to include key
characterisation aspects. It is also worth considering whether the current legislativeand regulatory framework constrain progress in managing UL wastes and how
policy makers can direct change with the support of waste characterisation experts.

•

It is recommended that the waste classification scheme is revised to enable themanagement of UL waste. It would be useful if the scheme were flexible to address
uncertainty in characterising UL waste. Part of that flexibility should provide forcases where radiological classification implies one set of actions, but other andmore significant hazards imply a different set of actions.

•

Establishment of clear definitions of end-states makes it clear to all stakeholders
what you are trying to achieve. This is likely to be an iterative process consistingof various stages especially on large complex sites. It is important that this is
supported by a robust safety case which continues to be developed through staged
implementation.

•

Different teams (either from one or multiple contractor organisations) working
within a radioactive waste management programme may have different specificobjectives, for example, historically, those concerned with waste storage have notnecessarily been concerned with waste disposal. Also, the information required
from characterisation can be quite different in these two areas. It is important thatboth these aspects are integrated.

•

Examples of UL waste have been described in terms of the characteristics that makethem different from waste arising in conventional decommissioning and otherplanned activities carried out in accordance with up-to-date standards. It should berecognised that many of these facilities will not have been, at least in part, operated
in accordance with modern standards.

•

Understanding the waste form is fundamental to the entire waste management
programme. However, it is essential that characterisation is carried out in a safemanner. There needs to be an informed decision-making process established so that
it is clear to all stakeholders when action is necessary and when delay is appropriate.
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To understand how best to apply and not over-stretch resources on characterisation,
more information is need on the proportion of remediation/decommissioning
project costs that are typically needed for waste characterisation and how muchdifferent techniques cost. This information would then enable reasonable, costeffective characterisation objectives to be set. Iteration of objectives may be needed
to find the appropriate balance between the amount of useful information that canbe obtained, the cost to obtain it against the urgency to take action to reduce thehazard. It is unlikely that a project would look at relaxing the characterisation
objectives, so it is recommended that when setting the initial objectives, it is better
not to set them too high. However, whichever approach is taken, all stakeholders
should be engaged in that discussion from an early stage (see Figure 5.2).

•

Identifying resource constraints and gaps in scientific technical capability will help
to inform the national training campaigns a well as R&D programmes.

Overall, the current characterisation methodologies discussed in Chapter 2 are applicable
to UL waste as they are to conventional waste. Within their activities and observations, theexpert group has identified several areas for adaptation and enhancement to themethodology for characterising of unconventional waste that would merit further attention.
Among the topics are:•setting a clear strategy underpinned by a regulatory framework to implement thepolicy;•having an integrated waste management strategy to optimise resources and time;•

having a clear and flexible waste classification system and WAC such that UL
waste could be included and that addresses radiological and other hazardous
content of the waste in a proportionate manner;•having a flexible and adaptable sampling methodology;•

understanding the end-state goal prior to undertaking characterisation activities;•

early and open dialogue with all stakeholders;•

understanding that characterisation is a continuous process;•having a clear review and validation process.

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Annex A: Analytical methods for assessing radionuclides in
the waste samples
FranceSample preparation comprised of:•Microwave or acidic digestion: An aliquot of the homogenised sample (around
0.2 g dried sample) was dissolved in a microwave system using a mixture of nitricacid (HNO3), hydrochloric acid (HCl), hydrogen peroxide (H2O2) and hydrofluoric
acid (HF). The precise composition of the acids was adjusted depending the sample
matrix. Due to the mass limitation of the microwave system, the acidic digestion
was carried out at atmospheric pressure.

•

Gamma spectrometry: After microwave digestion, the homogeneous solutions were
measured on a high-purity germanium detector (HPGe) using a calibrated
measurement geometry to determine the gamma emitters. All the radionuclides
between 40 keV up to 2Mev were of interest. The detection limit depended on thesample activity (Compton). The detection limit was optimised by adjusting thecounting time.

•

Transuranic nuclide measurements: After a second round of microwave digestion,
the alpha nuclides (americium, curium and plutonium) were precipitated by adding
cerium fluoride or ferrous hydroxides. The precipitate was filtrated. Plutonium wasseparated from disruptive nuclides including 241Am by using different ion exchange
resins or solid phase extraction chromatography using U/TEVA or TRU resins.
Alpha emitters were measured by alpha spectrometry with a passivated implanted
planar silicon (PIPS) detector. The chemical yield was determined by adding adefined amount of the 243Am and 242Pu tracer nuclides to the aliquot. The impurity
level in this tracer had to be measured and more specifically the quantity of 241Pu
present to prevent inaccurate 241Pu measurements. The purified plutonium solution
was mixed with a scintillation cocktail and 241Pu was measured by liquid
scintillation counting. The chemical yield was determined by alpha measurement.
A certified 241Pu standard solution was used for the efficiency calibration of theliquid scintillation counter. Due to the high content of alpha emitters in the waste,
the uranium and plutonium isotopic compositions were determined by mass
spectrometry measurement after selective chemical separation. The masscomposition was defined by thermal ionisation mass spectrometry or by inductively
coupled plasma with a mass spectrometer.

•

Alpha spectrometry can provide all isotopic information on all Pu isotopes. 238Pu,
239+240
Pu, and radioactive Pu tracers (e.g. 242Pu or 236Pu) were detected in the same
spectra. By this technique, the overlapping peaks of 239Pu and 240Pu could not be
resolved. Beta-emitting 241Pu could not be detected by alpha spectrometry but could
be easily measured after a simple procedure thanks to liquid scintillation counting.
However, TIMS or ICPS/MS measurements are more sensitive to the long-lived
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radionuclides, 239Pu and 240Pu: the 239Pu /240Pu ratio was then determined. As
explained, both the techniques are complementary; they do not replace each other.
•

By using nuclear data (half-live and specific mass), the results of both measurement
techniques can be crosschecked. The isotopic composition of the waste can then be
more precisely evaluated and the confidence level for the result is increased.

The samples were then analysed using the methods summarised in Table A1.
Table A1. Summary of methods and instruments used to analyse samples taken from CEA
CadaracheRadionuclide/chemical
63
Ni

Sr

90

H and 14C

3

Total organic carbon

Chemical composition

MethodNickel was separated from other nuclides by using theselective properties of the dimethylglyoxyme (DMG)-Ni
complex by different methods (liquid/liquid extraction orsolid phase resins such as “NI resin”). An aliquot of thepurified nickel compound was mixed with a scintillation
cocktail prior to measurement.
After microwave digestion, a defined amount of strontium
substrate was added to an aliquot of the digestion solution.
Strontium was separated from disruptive nuclides by using“SR resin”. Immediately after the separation of 90Y, 90Sr wasmeasured by liquid scintillation counting. After at least threedays, the samples were once again measured to determine
the renewable 90Y. The 90Sr activity was checked by the 90Y
activity equilibrium measurement. A certified 90Sr standard
was used for the efficiency calibration of the liquid
scintillation counter.
A total of 25 to 100 mg of the sample was covered withcatalyst to reduce the decomposition temperature and heated
up to 900-950°C. The resulting gases were drawn through agas bubbler filled with diluted sulphuric acid solution(H2SO4) and then through a sodium hydroxide solution(NaOH 1M) in which 3H and 14C were respectively trapped
prior to measurement.
The total organic carbon and carbonate content wasdetermined by applying European Standards methods on theliquid samples (DIN EN 1484) and on the solid samples(DIN EN 13137).
The chemical composition was determined on an aliquot ofsolution after microwave digestion. This was used todetermine the main elements, minor elements as well as
hazardous elements: Sb, Pb, As, etc.

Instrument
Inductively coupledplasma atomic emission
spectroscopy(ICP-AES) + Liquid
scintillation betaspectrometry
Inductively coupledplasma – mass
spectrometry(ICP-MS) + Liquid
scintillation betaspectrometry

Liquid scintillation
beta-spectrometry

TOC Analyser

ICP-AES and MS

Ukraine Rovno salt-bitumen compound
Sample preparation comprised of:•Decomposition of samples by wet ashing using a microwave digestion system forthe identification of radionuclides: 41Ca, 90Sr, 99Tc, 226Ra, 54Mn, 55Fe, 59Ni, 63Ni,
110m
Ag (this is specific to this waste stream), 93Zr and 94Nb.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

158 | NEA/RWM/R(2020)2•••

Decomposition of samples by dry ashing using a microwave muffle furnace,followed by acid leaching for the identification of radionuclides: 235U, 236U, 238U,
237
Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am.
Decomposition/isolation of radionuclides using acid decomposition to trap 3H and
C and dissolution of the analytical sample in an organic solvent followed byextraction of water-soluble components to identify 36Cl and 129I.

14

Decomposition products from the analytical samples were used to identify thegroup of radionuclides with similar chemical properties and radiochemical
characteristics.

The samples were then analysed using the methods in Table A2.

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTENEA/RWM/R(2020)2 | 159

Table A2. Method and analysis of samples taken from Rovno salt-bitumen compound
Radionuclide
Cs, 137Cs and 60Co
235
U, 236U, 238U, 237Np, 238Pu,
239
Pu, 240Pu, 241Pu, 242Pu, 241Am,
243
Am and 232Th
134

36

41

54

93

Instrument
Gamma Spec
12-channel alpha-spectrometer
Liquid scintillation betaspectrometry (for 41Pu)

Cl and 129I

Isolated by extraction column
chromatography on Cl-resin.
Eluate was then measured.

Ca, 90Sr, 99Tc and 226Ra

Determined by hydroxide
precipitation; isolation of
isotopes group by ion exchange
column chromatography on
specific resins; separation ofindividual isotopes by extraction
column chromatography (Srresin, MnO2-resin).

Liquid scintillation betaspectrometry (41Ca, 90Sr and99
Tc)

Determined by using hydroxide
precipitation. The isolation of
55
Fe was by column extraction
chromatography on specific resin
and the isolation of 110mAg byprecipitation in the form of
AgCl. The isolation of 59Ni and63
Ni was by extraction column
chromatography on Ni-resin.

Gamma spectrometry (54Mn and
110m
Ag)

Mn, 55Fe, 59Ni, 63Ni and 110mAg

Zr and 94Nb

H and 14C

3

10

MethodDirect measurement
Co-precipitation of isotopes andseparation by extraction column
chromatography on specificresins. The counting samples
were obtained by electrolytic
deposition.

Be and 135Cs

232

Determined by hydroxide
precipitation and the extraction
technique in “liquid-liquid”
system (extractant ‒ tenoyl
trifluoroacetone).
3H in the form of water
condensate.
14C in the form of Na2CO3.
10
Be was determined by liquid
extraction using tenoyl
trifluoroacetone. 135Cs was
isolated by co-precipitation withammonium phosphomolybdate

Th and 235U

CHARACTERISATION METHODOLOGY FOR UNCONVENTIONAL AND LEGACY WASTE

Liquid scintillation betaspectrometry

Liquid scintillation alpha
spectrometry (226Ra)
Liquid scintillation betaspectrometry (63Ni and 55Fe)

Mass spectrometry (59Ni wasmeasured followed byconversion of massconcentration units into activity
units)
Liquid scintillation betaspectrometry (93Zr)
Gamma spectrometry (94Nb)
Liquid scintillation betaspectrometry
Mass spectrometry (conversionof mass concentration units into
radioactivity units)
Mass spectrometry (they were
below the limit of detection on
the alpha-spectrometer)