Subsidy Project of Decommissioning and Contaminated Water
Management Started From FY2021

Research and Development for Treatment andDisposal of Solid Waste
Final Report for FY2021
September 2022
International Research Institute for Nuclear Decommissioning(IRID)All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)

Contents of the Report
Page No.
1. R&D approach2
2. Project details
12a. Storage management
12(a) Investigation and evaluation of storage methods
13[1] Investigation of storage methods for high dose waste
14[2] Investigation of the requirements for containers and the storage facilities
45(b) Technological development of contamination evaluation to segregate solid waste
126b. Establishment of concepts of treatment and disposal, and development of safety assessment methods 195(a) Establishment of selecting advance treatment methods
196[1] Low-temperature treatment technique
199[2] Study on approach to assessing applicability for treatment technology
261(b) Provision of disposal methods and development of safety assessment methods
273[1] Information organization to study the disposal concepts and to establish safety assessment methods
273[2] Development of techniques for assessing impact of affecting substances, etc. on disposal
324
c. Characterization of waste
339(a) Efficient characterization
341[1] Investigation of analytical inventory estimation methods
341[2] Proposal of mid-and-long-term analysis plan
366(b) Development of sampling technology
388
d. Integration of R&D results
420
3. R&D management
437©International Research Institute for Nuclear Decommissioning11. R&D Approach©International Research Institute for Nuclear Decommissioning21. R&D approach-Background and purpose of R&DBasic policies of waste management in the “Mid-and-Long-Term Roadmap”◆

It is necessary to understand the properties of solid waste including nuclide
composition and radioactive concentration, so as to study its treatment and disposal.

◆

The solid waste generated shall be stored and managed safely and rationally inaccordance with its properties.

◆

To ensure safety in the storage and management of solid waste, a method for rationallyselecting treatment (preceding processing) for waste stabilization andsolidification shall be developed, and preceding processing methods shall beselected before determining the technical requirements for disposal.

◆

To effectively proceed with research and development (R&D) project related to treatment
and disposal of solid waste, the R&D project teams of characterization, treatment anddisposal of solid waste shall closely work together. The progress and issues of eachproject shall be shared to proceed with R&D among the project members whileobtaining an overview of activities pertaining to solid waste management andidentifying the need for R&D tasks.
Until around the end of FY2021

Technical prospects of measures for the treatment and disposal ofsolid waste, and their safety are indicated.
*Mid- and Long-Term Roadmap Toward the Decommissioning of Fukushima Daiichi Nuclear Power Station Units 1-4, TEPCO(September 26, 2017), excerpts from 4-5, partly revised.
©International Research Institute for Nuclear Decommissioning31. R&D approach
Goals for achieving the milestones of the Mid-and-Long-Term Roadmap◼ The following specific goals for the milestones to be achieved until around the end of FY2021 are indicatedin the strategic plan* of the Nuclear Damage Compensation and Decommissioning Facilitation Corporation(hereinafter referred to as NDF). R&D initiatives are undertaken to contribute to the attainment of the goals.

Specific goals related to technical prospects
1.

2.
3.

4.

5.

6.

Safe and rational disposal concepts shall be formulated based on the properties and amountof solid waste generated at the Fukushima Daiichi Nuclear Power Station (NPS) and thetreatment technologies that can be used for such solid waste. In addition, safety assessment
methods reflecting the properties of the disposal concept shall be developed based on
examples from various foreign countries.
The analysis and evaluation methods for characterization shall be clearly specified.
Treatment technologies that are expected to be introduced in actual conditions for stabilizationand solidification of waste with disposal in mind for various important waste streams (e.g.,secondary waste generated from water treatment) shall be clearly specified.
A system for the rational selection of treatment (preceding processing) methods forstabilization and solidification shall be established based on No. 3 above before thedetermination of the technical requirements for disposal.
With respect to solid waste for which treatment technology that considers disposal is notclearly specified, establishment of treatment and disposal measures using the series ofmethods developed prior to FY2022 shall be foreseeable.
The challenges pertaining to storage management of solid waste until waste conditioning andcorresponding measures shall be clearly specified.

* 3.2.2.4. in Technical Strategic Plan 2019 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company Holdings, Inc.
(Nuclear Damage Compensation and Decommissioning Facilitation Corporation, September 9, 2019
©International Research Institute for Nuclear Decommissioning41. R&D approach
Goal of R&D• Solid waste management at the Fukushima NPS shall be stored and operated safely andrationally depending on its properties.
• A rational way for selecting preceding treatment methods shall be developed.
•* Goal excerpted from “2. Contents of Project” of Guidelines for applying to the “Project of Decommissioning and Contaminated Water Management(Research and Development of Processing and Disposal of Solid Waste)” (March 1, 2021), This goal is same as it of the existing subsidized project.
This research is a continued project to achieve the same goal.

Development of
selection system

Study on storage andmanagement
Waste
characterization
methodReflection on the storageand management methodsof Fukushima Daiichi NPS
*1
*2

*3

Process composition of treatment
technology, operating conditions, etc.
Conditions for a more rational systemsuch as specifications of waste, focused
nuclides*, etc.
Combination of applicable disposal
concepts and treatment technologies

Method ofassessing theapplicability oftreatment
technologies

Options ofdisposal concept

Disposalsafetyevaluation

ScenarioParameters

Waste package
specifications

Viable options ofstreams 3

Disposal safety
evaluation method

Properties ofwaste

Viable options ofstreams 3• Investigation and evaluation ofstorage methods• Technological development ofcontamination evaluation for
sorting solid wastes

Feedback to
studies 2

Process
information.1* In this research, 27 types of accident wastes that cover a wide range of wastes were selected, and focus was given to 67 nuclides (e.g., C-14 and Pu-239) that weredeemed probable in the statistical inventory study©International Research Institute for Nuclear Decommissioning51. R&D approach
Reflection of R&D results on the decommissioning
Waste management (project)

Mid-and-Long-Term Roadmap

R&D• Contaminationassessment technology
for sorting solid wastes[milestones of waste management]

Generation of waste• Efficient characterization• Development ofsampling technology

Storage2019• Investigation andevaluation of storage
methods

2020

Stabilization• Integration of
R&D results

Storage• Establishment of
selecting advancetreatment methods

2021
Present

Treatment*

Disposal• Provision of disposal
methods and development of
the safety evaluation method

Milestone

Indicate treatment
and disposal
measures andtheir technicalprospects
pertaining to thesafety*Solidification and storage of waste according to the results of treatment and disposal study©International Research Institute for Nuclear Decommissioning61. R&D approach
Association with other research【 Input into this project 】
ID

Implementation items and specificdetails (use of information)

Necessary informationTime
periodSource from1

Understanding of storage conditionsof high dose waste to study for their
management.

Consumables of equipment to be used for fueldebris retrieval, and specifications of water
treatment and air-conditioning systems, andcollecting containers.

As
neededExchange
Fuel debris
information as
retrieval project
necessary2

Understanding of storage conditionsof high dose waste to study for waste
management.

Specifications of systems to be used forcanisters of collecting, containing and storing
fuel debris

Exchange
As
Canister project information asneedednecessary

3

Study on waste management stream
and acquisition of basic knowledge
concerning approach to applicability
assessment of treatment technology

Information on high-temperature treatment
technologyAsneededProject
proposers

Cooperate throughadministration
officePurpose of use at output destinationTime
periodDestination

Remarks

Remarks【Output from this project】
ID

Implementation items and output
details1

Research results of subsidized
projects

Acquisition of basic data on study of hightemperature treatment technology

2

Sharing of storage management
results and adjusting the connecting
conditions

Confirmation of compatibility with the fueldebris retrieval process3

Confirmation of evaluation conditionsof the amount of hydrogen generation

Comparison with evaluation of hydrogenamount generated from fuel debris

4

Knowledge of waste characterization

Results of evaluations on analysis data ofsamples taken from Fukushima Daiichi andstudy on generation process of the samples

Cooperate throughadministration
officeExchange
As
Fuel debris
information asneeded retrieval project
necessary
Exchange
As
Canister project information asneedednecessary
Fuel debris
Participate
Ascharacterization assessment
neededproject
committee
AsneededProject
proposers

Note: Fuel debris retrieval project: Development of Technology for Further Increasing the Scale of Retrieval of Fuel Debris and Internal Structures Project
Canister project: Development of Technology for Collection, Containing and Storage of Fuel Debris Project
Fuel debris characterization project: Development of Technology for Analysis and Estimation for Fuel Debris Characterization Project
Project proposers: Organizations that proposed to study applicability of practical high-temperature treatment technology for solid waste to establish selecting methods for advance treatment method.

©International Research Institute for Nuclear Decommissioning71. R&D approach
Implementation schedule (1/3)
Detailsof Implementation
実施内容

Planned計画
Actual実績

FY20212021年度

4

5

6

7

8

9

10

11

12

1

2

3

a. Storage
managementa.
保管管理(a) (a)Investigationand evaluation of storage methods
保管・管理方法の検討・評価①
Study of storage methods for high dose waste①高線量廃棄物の保管方法の検討

Drafting of theinvestigation plan
調査計画立案

Organization of the properties of waste andparameterization
廃棄物性状の整理とパラメータの検討

Summary
まとめ・廃棄物性状の整理とパラメータの検討
・Organizationof the properties of waste and parameterization
Drafting of theinvestigation plan
検討計画立案

Evaluation of the amount of hydrogen gas generated by typicalhigh dose waste
代表的な高線量廃棄物の水素ガス発生量の評価

Summary
まとめ・Evaluationof the amount of hydrogen gas generated by typical high dose waste
・代表的な高線量廃棄物の水素ガス発生量の評価② ②容器及び保管設備の要求事項の検討
Study of the requirements for containers and storage facilities

Drafting of theinvestigation plan
検討計画立案

Investigationof problem cases
トラブル事例の調査

Organization of issues andstudyof measures
課題の整理と対策の検討

Summary
まとめ・Investigationand organization of problem cases
・トラブル事例の調査、整理
Drafting of theinvestigation plan
検討計画立案

Investigation of performance assessment
methods
for vent filters
ベントフィルタの性能評価方法の検討

Summary
まとめ・Investigationof performance assessment methods for vent filters
・ベントフィルタの性能評価方法の検討(b)Technologicaldevelopment of contamination evaluation for sorting solid waste(b)
固体廃棄物の分別に係る汚染評価技術開発

Draftingof plans
計画立案

System design
システム設計Summaryまとめb. Establishment of concepts of treatment and disposal, and development ofb.
処理・処分概念の構築と安全評価手法の開発
safetyassessment methods(a) (a)Establishmentof selecting advance treatment methods
先行的処理方法の選定手法の構築①①低温処理技術
Low-temperature treatment technologies
i. Establishment of methods for examining the possibility of solidification by means of lowi.temperature
低温固化処理固化可能性検査手法の検討
solidification treatment

Drafting of the experimental planand carrying out outsourcing
formalities
検討計画立案・委託手続き

Implementation of tests and analyses(outsourced,
in-house)
各試験及び分析実施（委託、内部実施）

Test completion
andreport
試験完了・報告

Summary
まとめ
Drafting of the experimental planand carrying out outsourcing
formalities
検討計画立案・委託手続き

Implementation of tests and analyses(outsourced,
in-house)
各試験及び分析実施（委託、内部実施）

Test completion
andreport
試験完了・報告ii. Investigationof degradation of solidified substances by heating, etc.
ii.加熱等による固化体変質の調査

Summary
まとめ

Drafting 検討計画立案of the experimental plan

Implementationof analysis (in-house)
解析実施（内部実施）

Test completion
andreport
試験完了・報告iii. Evaluation of the relationship between the inventory of Cs, etc. and theiii.Cs等インベントリと固化体温度の関係性評価
temperatureof solidified substances

Summary
まとめ
Drafting of the experimental planand carrying out outsourcing
formalities
検討計画立案・委託手続き

iv. Investigation
into and evaluation of factors influencing long-term degradation
iv.長期的な変質に与える影響因子の調査と評価

計画

Implementation of tests and analyses(outsourced,
in-house)
各試験及び分析実施（委託、内部実施）

Test completion
andreport
試験完了・報告
Summary
まとめ

The target completion deadline of each work step is indicated by “▽”.

©International Research Institute for Nuclear Decommissioning81. R&D approach
Implementation schedule (2/3)
実施内容
Detailsof Implementation

Planned計画
Actual実績② ②処理技術の適用性評価アプローチの検討
Investigation of approach for assessing applicability for treatment technology

FY20212021年度4

5

6

Draftingof the study plan
検討計画立案

7

8

9

10

11

Investigationand organization work (in-house)
調査整理作業（内部実施）

12

1

2

3

Test completion
and report
試験完了・報告i. Consolidationof acquired data and evaluation method used for technologicali. 技術比較に用いる取得データ及び評価軸の整理

Summary
まとめcomparison
Drafting of the experimental planand carrying out outsourcing
formalities
検討計画立案・委託手続き

Implementation of tests and analyses(outsourced,
implemented in-house)
各試験及び分析実施（委託、内部実施）

Test completion

試験完了・報告and reportii. Investigation related to Cs volatilization volume during high-temperatureii. 高温処理時のCs揮発量及び抑制に関する調査
treatment
and its control(b)of disposal concepts and development of safety assessment methods(b)Provision
処分概念の提示及び安全性評価手法の開発① Information organization to study the disposal concepts and to establish safety① 処分概念の検討と安全評価手法の構築に向けた情報整備
assessment
methods

Summary
まとめ

Draftingof the study plan
検討計画立案Studyof disposal concepts
処分概念の検討

Summary
まとめ

Draftingof the study plan
検討計画立案Information
organization
情報整備

Summary
まとめ・Studyof disposal concepts for typical waste streams
・代表的な廃棄物ストリームを対象とした処分概念の検討

・Information
organization to establish safety assessment methods
・安全評価手法の構築に向けた情報整備② ②処分への影響物質等の影響評価手法の開発
Development of techniques for assessing impact of affecting substances, etc. on disposal

Draftingof the study plan
検討計画立案

Organizationof sorption impact, and study of parameters
収着影響の整理、パラメータ検討

・Evaluationof nuclide sorption impact
・核種収着影響の評価
Drafting検討計画立案of the study plan

Sensitivity
evaluation of parameters including sorption impact
収着影響を含むパラメータの感度評価

・Sensitivity
evaluation of the dose evaluation parameters
・線量評価パラメータの感度評価
Summary
まとめ

ccCharacterization
性状把握 of solid waste(a)characterization(a)Efficient
性状把握の効率化①
Investigation of analytical inventory estimation methods①解析的インベントリ推算手法の検討
i.ⅰ.
Implementationof analysis
分析の実施Collectionof rubbles
瓦礫等の採取（i）試料の採取、保管、輸送(i)
Collection, storage and transportation of samples
・Collection,
storage
・採取・保管is carried out throughout the year. The date of transportation will be changed appropriately depending on the**Collection
採取は通年実施する。輸送実施日は試料の準備状況に応じて適宜見直す。
status of preparation of the samples.
Analysis samples
transportation (First time)
分析試料輸送（1回）

Analysis samples
transportation (Second time)
分析試料輸送（2回）

・Transportation
・輸送
Study of the plan for次年度計画検討
the following year
・Studyof the plan for the following year
・次年度計画検討

計画

The target completion deadline of each work step is indicated by “▽”.

©International Research Institute for Nuclear Decommissioning91. R&D approach
Implementation schedule (3/3)
Detailsof Implementation
実施内容

FY20212021年度Planned計画
Actual

実績

4

5

6

9

8

7

10

3

2

1

12

11

Implementation
分析実施 of analysis(ii)（ⅱ）分析の実施
Implementation of analysis**Analysis
is carried out throughout the year. Moreover, the analysis plan is changed appropriately as needed.
分析は通年実施する。また分析計画は必要に応じて適宜見直す。
As needed随時As
needed随時

Asneeded随時

As needed
随時(iii)
Analytical database（ⅲ）分析データベース*Analytical
database is compiled and then reported as needed.
* 分析データは取りまとめ次第随時報告する。
iiⅱ.
Study
on analytical inventory estimation methods
解析的インベントリ推算手法の検討(i)
Status of generation and status of storage of solid waste（i）固体廃棄物の発生状況・保管状況(ii)
Contamination mechanism（ⅱ）汚染メカニズム(iii)
Study on analytical inventory estimation methods（ⅲ）解析的インベントリ推算手法の検討

Study of classification based on the generation
発生・保管形態等による分類検討and storage
form

Study水による汚染に関する検討
related to contamination through waterStudyrelated to contamination through air
空気による汚染に関する検討

Drafting検討計画立案of the study plan

Study related to classification of waste and estimation of
廃棄物分類に関する検討と廃棄物インベントリの推算
waste
inventory

Summary
まとめ

Compilation and summarization

推算の課題の整理とまとめof estimation issues②②中長期的な分析計画の提案
Proposal for mid-and-long-term analysis planiⅰ.
Developmentof mid-and-long-term analysis plan
中長期的な分析計画の作成

Consolidation of information related
Consolidation of sample information,
分析に関係する試料情報、廃炉工程等の整理分析能力に関する情報の整理
toanalytical capabilities
decommissioning schedule, etc. related to analysis

Development of mid- and long中長期的な分析計画の作成
term
analysis plan

iiⅱ.
Study
of analytical goals based on needs
ニーズを踏まえた分析目標の検討

Consolidation of analytical needs, order of priority, identification of
分析ニーズ、優先度、対象核種の抽出、分析手法等の整理
targetnuclides, analysis methods, etc.

定量的な分析目標の検討
analytical goals

Study of quantitative

Summary
まとめ

Evaluation of the validity of

Summary
まとめ

iii
Study
of analytical planning methods
分析計画法の検討ⅲ.

(b)of sampling technology
サンプリング技術の開発(b)Development

試行結果の妥当性の評価
the trial run results

Trial run of the study ofanalytical planning methods
分析計画法検討の試行

Drafting
up the study plan
検討計画立案

Summary
まとめ

Design, manufacturing
and testing of the sampling system
試料採取装置の設計・製作・試験

Study and 採取計画の検討・策定
formulation of sampling plans

Summary
まとめ

Summaryまとめ

d.
of R&D results
研究開発成果の統合
d. Integration
Drafting
up plans
計画立案

Study of case studies and consolidation of the viewpoints fornarrowing
down
ケーススタディの検討と絞り込みの視点の整理

Summary
まとめ

The target completion deadline of each work step is indicated by “▽”.

©International Research Institute for Nuclear Decommissioning10

1. R&D approach
Project organization chart
Tokyo Electric Power Company Holdings,
Inc.
〇 Coordination from the perspective of on-site application(Coordination to reflect the decommissioning status of the site,
etc.)Japan Atomic Energy Agency〇 Management of technologicaldevelopment (a - d)
〇 Development (b, c, d)〇 Operational assistance (e)IRID

Hitachi-GE Nuclear Energy,Ltd.
〇 Development (b, d)〇 Operational assistance (e)Fujimoto Chemicals Co., Ltd. (b.(a)②①)*
Inspection and calibration of CHN analyzer

QJ Science Co., Ltd. (b.(b)②)*
Investigation of methods for assessing the impactof affecting substances on nuclide migration

Visible Information Center Inc. (b.(b)②)*
Investigation of methods for analyzing andassessing the transition in chemical environmental
due to disposal and nuclide migration

Tokyo Nuclear Services Co., Ltd. (b.(b)②)*
Acquisition of basic data on the sorption of nuclides
under coexistence with affecting substances

Kaken Co., Ltd. (b.(b)②)*
Acquisition of basic data on uranium sorption tobarrier materials under coexistence with affecting
substances

CRIEPI A7 (b.(a)①)
Investigation related to Cs volatilization volume
during high-temperature treatment and its control(outsourced research)

CRIEPI A7 (b.(a)①)
Acquisition and organization of data used instudying the evaluation approach (outsourced
research)

Taiheiyo Consultant Co., Ltd. A8 (b.(a)①)
Study related to solidification properties of cementbased material and its long-term property changes(outsourced research)
Hitachi Transport System East Japan Co., Ltd(c.(a)①)*
Transportation of analysis samples from the
Fukushima Daiichi Nuclear Power Station to theanalysis facility

FY2021〇 Formulation of overall plan and technical management (a - e)
〇 Technological management such as progress of technologicaldevelopment (a - e)
〇 Coordination related to technological development (a - e)

Ascend Co., Ltd (c.(a)①)*
Service contract concerning analysis and testing related to theinvestigation of properties of secondary waste generated fromcontaminated water treatment in Fukushima

Inspection Development Company Ltd. (C.(a)①)*
Service contract concerning β-nuclide analysis andcompilation of results regarding contaminated water inFukushima and secondary waste generated fromcontaminated water treatment

MHI Nuclear Development Corporation :Note (c.(a)①)
Work of radioactivity analysis of the waste from the
FY2021 accident
Note: Known as Nuclear Development Corporation until
November 2022.
Nippon Nuclear Fuel Development Co., Ltd. (c.(a)①)
Analysis and evaluation of highly contaminatedsubstances related to the accident at Fukushima Daiichi
Nuclear Power Station

Tokyo Power Technology Ltd. (c.(a)①)
Collection of analytical samples such as rubbles, etc.
from Fukushima Daiichi Nuclear Power Station

CRIEPI (c.(a)①)
Research related to streamlining waste classification
during inventory estimation (outsourced research)

National Nuclear Laboratory (UK) (c.(a)①)
Statistical sampling of wastes at Fukushima Daiichi
Nuclear Power Station and development of analysis plan(outsourced research)＊Decided based on competitive bidding, etc.

Toshiba Energy Systems &Solutions Corporation

Atox Co., Ltd.
〇 Development (c, d)〇 Operational assistance (e)〇 Development (a, d)〇 Operational assistance (e)JGC Japan Corporation (b)*Information organization to study the disposal concepts
and to establish safety assessment methods

General Engineering Co., Ltd. (d)Assistance in the organization of information related to thestudy on waste streams

Hamamatsu Photonics K.K. (a)Designing and manufacturing of fluorescent detector of
the surface α contamination detector

Sigmakoki Co., Ltd. (a)Designing and manufacturing of the optics system of thesurface α contamination detector
Techno AB (a)Designing and manufacturing of the circuit for the surface
α contamination detector

Mechano Breit Co., Ltd (a)Designing and manufacturing of the housing of thesurface α contamination detector

Tokiwa Copy Co., Ltd.
Printing of an overview of the research results

Orano ATOX & SOLUTIONS (c)Designing and manufacturing of the samplingsystemKK Kouyou (c.(b))
Designing and manufacturing of the work dock[Legend indicating the work for which the organization is responsible]a.
Storage managementb.
Establishment of concepts of treatment and disposal, anddevelopment of safety assessment methodsc.
Characterizationd.
Integration of R&D resultse.
R & D management©International Research Institute for Nuclear Decommissioning

11

2. Project details (No.1)
a. Storage management(a) Study and evaluation of storage[1] Investigation of storage methods for high dose waste[2] Investigation of the requirements of storage containers andfacilities(b) Technological development of contamination evaluation forsegregate solid waste©International Research Institute for Nuclear Decommissioning

12

(a) Investigation and evaluation of storage and management methods• According to the Mid-and-long-term Roadmap*, solid waste generated in the Fukushima Daiichi Nuclear
Power Station (NPS) shall be stored and managed in a safe and effective manner considering the properties.
*Excerpted from 4-5, Mid-and-long-term Roadmap towards the decommissioning of the Fukushima Daiichi NPS, Tokyo Electric Power Company (TEPCO) Holdings, Inc.(Sep. 26, 2017)

Study of storage andmanagement methods✓ Study on the safety functions
Significant issues for the storage and management
・Hydrogen release (Ensuring air permeability (hydrogen
permeable function)
Coping with conflicting
requirements of the safetyfunctions・Prevention of spreading nuclides Sealability (Ensuring theMain study items since last
fiscal year

✓ Study and update of the process flow

function of particle collection)(Ensuring the function ofhydrogen permeability)

Study on the requirements of the container andstorage facilities

Evaluation of the amountof hydrogen generation(Ensuring the function ofparticle collection)

Examining the requirements
for a vent filter[1] Storage method for high dose waste
Evaluation of hydrogengeneration behaviors(Clarification of issues to evaluate hydrogengeneration behaviors for a wider range oftarget objects)(A detailed study)[2] Investigation of the requirements for the container and storage facilities

Survey of trouble cases for thecontainer and storage facilities

Performance evaluation of ventfilters(Clarification of issues and measures, andinvestigation of trouble cases to expand therequirement information)(Clarification of environments to affectperformance and lifetime of filters, and studyof the filter performance test conditions andmethods)(A new investigation)(A detailed study)

Study items for FY2021

A detailed study and a new investigation is conducted in FY 2021.

Reflecting to the storage and management method for the Fukushima Daiichi©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste◼ Achievements up to FY2020
 Referring one example of the storage method for high dose waste, hydrogen generation rate wasestimated based on decay heat that Cs-137 is set as a representative of a contaminated nuclide.
◼ Goal Issues of evaluation on hydrogen generation behaviors for high dose waste are clarified.

◼ Details of implementation (Details are described in the next pages.) Information acquisition
Parameters including properties of high dose waste and G values of hydrogen generation areclarified.
 Study of factors required for evaluation of the amount of hydrogen generation
High-level radiation waste to be required for measures of hydrogen generation during storage isselected and factors that will affect evaluation of hydrogen generation rates are studied.
 Evaluation of the amount of hydrogen generation✓ Parameters of the hydrogen generation evaluation model that would affect hydrogen generation
rates are identified and the effects of the hydrogen generation rates are investigated.
✓ Parameters for each factor that would affect the evaluation of hydrogen generation rates are set inmore practical manner and the amount of hydrogen generation is evaluated.
✓ On a base of evaluation results of the hydrogen generation amount of a representative high dose
waste, issues of the evaluation are clarified.
◼ Indicators for the achievements of goal Issues of the hydrogen generation behaviors for high dose waste are clarified (Target at thecompletion: TRL 2 →3).
©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste<Study on estimation of the hydrogen generation amount in previous years and in FY2020>
Evaluation formula (linear model) and parameters of hydrogen generation rate as well as concepts of setting are described as below.
𝑯𝒚𝒅𝒓𝒐𝒈𝒆𝒏 𝒈𝒆𝒏𝒆𝒓𝒂𝒕𝒊𝒐𝒏 𝒓𝒂𝒕𝒆 = ෍(𝒅𝒆𝒄𝒂𝒚 𝒉𝒆𝒂𝒕 𝒐𝒇 𝒏𝒖𝒄𝒍𝒊𝒅𝒆 𝒊 𝒂𝒅𝒊𝒂𝒕𝒊𝒐𝒏 𝒕𝒚𝒑𝒆 𝒋 × 𝑮 𝒗𝒂𝒍𝒖𝒆 𝒐𝒇 𝒓𝒂𝒅𝒊𝒂𝒕𝒊𝒐𝒏 𝒕𝒚𝒑𝒆 𝒋 × 𝒘𝒂𝒕𝒆𝒓 𝒂𝒃𝒔𝒐𝒓𝒑𝒕𝒊𝒐𝒏 𝒓𝒂𝒕𝒆)𝒊,𝒋[3] Water absorption rate[2] G value*[1] Nuclide and radiation type*G value: The G value refers to the number

𝑫𝒆𝒄𝒂𝒚 𝒉𝒆𝒂𝒕 𝒐𝒇 𝒏𝒖𝒄𝒍𝒊𝒅𝒆 𝒊 𝒓𝒂𝒅𝒊𝒂𝒕𝒊𝒐𝒏 𝒕𝒚𝒑𝒆 𝒋= 𝑰𝒏𝒗𝒆𝒏𝒕𝒐𝒓𝒚 𝒐𝒇 𝒏𝒖𝒄𝒍𝒊𝒅𝒆 𝒊 × 𝑬𝒎𝒊𝒔𝒔𝒊𝒐𝒏 𝒓𝒂𝒕𝒆 𝒐𝒇𝒓𝒂𝒅𝒊𝒂𝒕𝒊𝒐𝒏 𝒕𝒚𝒑𝒆 𝒋 × 𝒆𝒎𝒎𝒊𝒔𝒔𝒊𝒐𝒏 𝒆𝒏𝒆𝒓𝒈𝒚 𝒐𝒇 𝒓𝒂𝒅𝒊𝒂𝒕𝒊𝒐𝒏 𝒕𝒚𝒑𝒆 𝒋 )[4] Inventory[5] Emission energy
Fig. 1 Concepts of parameters setting from FY2019 to FY2021
[2 ]G
value

Parameters[1] Nuclide iradiation type j

FY
2019

Nuclide:Represented by
Cs-137
Radiation type:Β and γ

FY
2020

FY
2021

No change

α nuclideexcluding Cs137, β,γ and αrays including
accidentderived
nuclides.

γ-and βrays
shall
both be0.45.

NochangeSetting
for each
nuclide[3] Water
absorption rate
Both γ-and βrays arecalculated as 1
for aconservative
approach.

Considering
waste cutting,storage situationof the innercontainer andthickness ofwater film

Considering typeof radiation
nuclides andwater existence.

[4] Inventory
Estimation of theamount ofradioactivity basedon weight of thewaste per a storage
container (0.6 ton)
and concentrationof radioactivity
weights**[5] Emission
energyThe β-rayenergy iscalculated by
the maximum
energy.

Estimation for FY 2020of molecules of reactant consumed(or product formed or reduced) per
100eV of energy absorbed whenradiation reacts.

Estimation for FY 2019

Storage container
Inner container

Metal waste(Top-accessretrieval)

Items
Radiation source(Bq/t)
Capacity of storage container (t)Radiation source(Bq)
Energy released(MeV/s)Hydrogen generation rate (m3/d)・Estimation of theamount ofradioactivity basedon areaconcentration ofradioactivity **・Estimation of thepossibility tocontain 8 innercontainers per one
unit.

The β-rayenergy isspectrally given
and calculated
by PHITS.

Waste excluding
waste to begenerated from
retrieving duel
debris shall be
estimated.

Evaluation
model
considering
properties ofeach waste
shall beestimated.

7 sheets

Decay heat
(MeV/s)
FY2019 model
FY2020model**Estimated inventory of waste to be generated during retrieving fuel debris.

Water screen:<Reference>
Conceivable wastes

Rate of absorption
into water

Hydrogengeneration rate
(m3/d)

Dryer

Nonconservative
estimation(hydrogengeneration rate:
Approx. 1/13.5)

Waste objects have been expanded in FY2021 anda more practical evaluation will be conducted toclarify issues for evaluation of hydrogen generation
behaviors.
©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste<Study of the storage methods for high dose waste>
Study items and the procedures for FY2021 are as below.
1. Information acquisition
１．情報収集Acquisition of information on high-radiation level waste
高線量廃棄物の情報収集• Specificationof the scope of・情報収集対象物の明確化
informationacquisition
・廃棄物性状の整理• Organization of the properties of waste

Organization of parameters of hydrogen
水素生成パラメータ整理
generation• Consolidationof results of existing studies
・既往検討の整理• Clarification of mechanism
・メカニズムの明確化• Organizationof information on the・G値のパラメータ情報の整理
parametersof G values

2. Study of factors required for

２．水素発生量評価に
evaluation of the amount ofhydrogengeneration
必要な要因の検討

Selection of wastes that require
対策が必要となる廃棄物のmeasures
選定• Study of waste
requirements•

・廃棄物要件の検討
Selectionof high dose waste
・高線量廃棄物の選定

Study of factors influencing hydrogen generation

水素発生因子検討

Study of factors influencing the evaluation of the
水素発生速度評価の影響因子の検討
hydrogen generation rate
(Radioactivity
inventory, type of radiation,（放射能インベントリ、放射線の種類、
moisture content,
status of water adhesion,
etc.)
含水率、水の付着状況
等）

Evaluation models and impact parameter
評価モデルと影響パラメータsetting••

Study of evaluation
設定models
Impact parameter setting
・評価モデル検討(Identification)
・影響パラメータの設定（抽出）3. Evaluation of amount
３．水素発生量評価
of hydrogen generation

Evaluation of amount of hydrogengeneration
水素発生量評価••

Specific
settings of impact parameters
・影響パラメータ値の具体的な設定
Calculationand evaluation of amount・発生量計算評価
generated

Considerations

考察 of the issues
Study and consolidation
relatedto evaluation methods
・評価方法に関する課題の検討・整理(・Re-examination,
as needed, of wastes（・必要に応じ、対策が必要となる廃棄物の再検討）
that require measures)•

1. Information acquisition• According to the waste list that was studied based on an
integration of R&D results (waste stream) , properties ofwaste (compositions of waste, radioactivity inventory andwater content rate, etc.) required to study the storage
management method are clarified from aspect of hydrogengeneration (high-radiation dose and water content or not).
• To examine parameters of hydrogen generation, thehydrogen generation mechanism is clarified and theninformation related to results of existing studies andparameters of G values are examined.
2. Study of factors required for evaluation of hydrogengeneration (hydrogen generation rates)• As a result of clarifying the hydrogen generation parameters,
impact factors of the hydrogen generation rate evaluation
are studied.
• Wastes needed for countermeasure during storage are
selected based on the results of information acquisition forhigh dose waste and study of hydrogen generation factors.
• The evaluation model for selected waste is studied andparameters (impact parameters) that will affect hydrogengeneration parameters especially hydrogen generation rateare extracted.
3. Evaluation of the amount of hydrogen generation (hydrogengeneration rate)• The impact parameters (practical estimation value) that will
be used for specific calculation evaluation are set based on
the study results and the amount of hydrogen generation for
representative high-radiation waste is evaluated by
calculation.
• Issues for the evaluation model and parameter setting arestudied based on a result of the calculated evaluation.
• The selection of waste to be required for countermeasure
are also re-examined whether it is appropriate or not basedon the evaluation results.

©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste<Clarification of hydrogen generation parameters>
Information acquisition
Hydrogen generation evaluation formula from radiolysis(1) Linear model
Hydrogen generation rate= Σi,j（Decay heat of nuclide i / radiation type j×G value of radiation type j×Water absorption rateof radiation type j)
Decay heat of nuclide i / radiation type j= Inventory of nuclide i×Emission rate of radiation type j×Emission energy ofradiation type j
Hydrogen generation rate is proportional to 1) absorption energy (decay heat),2) G value, and 3) absorption rate.
Decay heat depends on nuclide inventory, G value depends on radiation types
and absorption rate depends on water existence (thickness of water membrane,etc.) .Hydrogen continues generating at a certain rate.

(2) Radiolysis model [1]
Ci、 Cl、 Cm、 Cs：Concentration of material i, l, m and s (mol/l), giγ：G value of γ ray ofmaterial i (unit/100eV)、
ξ：Number of molecules and conversion factor of mol (mol/unit), Qγ： Absorption energy ofγ ray (eV), N：Number of materials affecting reactions (-)
klm：Reaction rate constant of material l and m (l/mol/sec)、ks： Reaction rate constant ofmaterial i and s (l/mol/sec)

Hydrogen generation rate depends on 1) Absorption energy (decay heat andproportion of energy absorption), 2) G value and 3) Reaction rate constant among
materials and thereby it is function of time.

Fig.1 : Aging changes of degradation productconcentration calculated by radiolysis model [2]*(L) and (g) in the figure represent liquid phase andgas phase.

[1] Yoichi Wada, Masahiko Tachibana, Atsushi Watanabe, Kazushige
Ishida, Nobuyuki Ota, Naoto Shigenaka, Hiromitsu Inagaki & Hiroshi Noda (2016) Effects of seawater components on radiolysis of water atelevated temperature and subsequent integrity of fuel materials, Journal of Nuclear Science and Technology, 53:6, 809-820[2] Junichi Takagi and Seiji Yamamoto “ Alpha-radiolysis Simulation to Predict Hydrogen Accumulation under Fuel Debris Condition ”,The
International Workshop on Radiolysis and Electrochemistry NPC 2016, Brighton, United Kingdom - October 7, 2016
©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Information acquisition<Clarification of hydrogen generation parameters>[Clarification of mechanism]
Applicability of the linear and radiolysis models
Hydrogen is produced by irradiating into water. Hydrogenproduced from irradiation mitigates into the gas phase of a container
and the partial pressure of hydrogen is increased.
The concentration of hydrogen underwater achieves equilibrium astime passes. The concentration can be calculated by the radiolysis
model.
The concentration of the gas phase hydrogen in the container isequal to the liquid phase hydrogen concentration in accordance with
the Henry’s Law.
When the concentration is less than 4 vol %, the countermeasurefor hydrogen generation is not necessary.
On the other hand, when the hydrogen concentration of the gas
phase is more than 4 vol %, hydrogen in the container has to bereleased into outside of the container by using a filter vent.
The hydrogen generation rate shall be conservatively calculated by
the linear model.

Ｈ2 (gas phase outsideof the container)

Ｈ2 (gas phase insideof the container)

Ｈ2Ｏ → Ｈ2
(underwater)
Fig.1 : Illustration of hydrogen
mitigation inside/outside of thecontainer

In the both cases,
It is important to set values of decay heat, G value andabsorption rate for the estimation.
©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Information acquisition<Clarification of hydrogen generation parameters>
Clarification of the G value parameter information [1]The distribution of chemical species are nearly homogeneous in space for a microsecond after physical processes thatwater absorbs radioactive energy (excitation and ionization), physical chemistry processes (spar formation) and chemical
processes (chemical reaction in a spar).
•Chemical yield of chemical species represents as G value and [number of species //100 eV] or [mol/J] are used as units.
•
After completing a spar reaction (up to 10-7 second), G value is called as the primary G value.
•
On the other hand, the amount of these degradation products that interact after long time has passed is called as“apparent G value” or “ effective G value.” Generally, the primary G value is larger than apparent G so that the primary G
value is used for evaluation.
Time (s)Physicalprocess

Excitation

Ionization

Physical chemistry
process
Spar formation

Chemicalprocess
Spar reaction

Spar formation① Low LET radiation(Secondary electron in the case ofelectromagnetic waves)② High LET radiation(Recoil proton in the case ofneutrons)

Spar diffusional reaction(Spatially uneven)
Diffusion
Reaction within the spar
Reaction between spars

Spatially evenreaction

End of spar reaction(Spatially even)

Reaction with the solute (in the case of a dilute solution)

Fig. 1 Overview diagram of the radiolytic reaction scheme of water (room temperature)

Fig. 2 Spar diffusional reaction process(Spatially uneven chemical process)[1] Yusa Muroya, “Radiation chemistry of water and water solution”, RADIOISOTOPES vol. 66 (2017）No. 10
https://www.jstage.jst.go.jp/article/radioisotopes/66/10/66_661006/_article/-char/ja©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Information acquisition<Clarification of hydrogen generation parameters>[Clarification of the G value parameter information]•
The primary G value depends on radiation types and energy, etc. (Table 1 [1]).
•Handling of free water and other type of water (i.e. Cement crystal water) is accordance with rules in each country (Table 2[2]).
[Examples]•
U.S. and U.K.: Hydrogen generation is evaluated considering the reduction of free water caused by cement solidification andchanges of the water condition•
France: G value for free water is used for estimation even after cement solidification* G values for estimating hydrogen generation are set after classification depending on properties of waste (water status, andtypes of organic substances and radioactivity) in each country
Table 1. Primary G value of major radiation degradation
products (molec./100eV)

Table 2. Clarification results of the concepts of G value setting in each country
The G-value was confirmed to be set according to the radioactive waste classificationand the state of water.

Types of radiation

US

Gamma rays

・At the time of drying (5 to
20wt%): 1.6
・When cement solidifies: 0.6 to

Carbonate slurry

0.85
G-value
(H 2)

UK

*1

Iron coprecipitation
Same as above
slurryReactor internal
structures

Concept of G-value (H 2)・Adhesion water: 1.6(After drying: 0)・Organized as per the waste
properties (presence of organic
matter, occurrence ofsolidification, moisture contentetc.). * 2・Even if the waste does notcontain α nuclides, the α ray Gvalue is conservatively assumed
to be the base.

FranceEstablished based on・Free water, cement hydration
discussions with the regulatorywater: 0.452 (Documented value)authorities
・Mg(OH)2: 0.051 (Documented(Example of initial settings is 0.4 orvalue)
0.45 for free water)・Free water, cement hydrationwater: 0.452 (Reference value)Same as above・Fe(OH) and water: 0.00529(Reference value)Same as above

・ Adhesion water: 0.45

・G-value of free water is used even
・G-value has been set in the past
after cement solidification.
as 0.05 to 0.5 considering the・A lower G-value can be set whenchange of state of water due toactual measurement is carried out.
cement solidification.
(0.032 for the DSC *3 dry sludge)
・Whether to adopt the・The G-value of waste is calculated
documented value or the actuallyfrom the sum of products of the massmeasured value is discussed withfraction of the generation source andthe regulatory authorities.
each G-value.

*1 Set according to the solidification conditions (depends on the type of cement material). In the case of Type I (OPC), G-value of
0.6 is used when the ratio of water to cement by weight is 0.45.
*2 CH-TRU payload appendices (books)*3 Dry Storage Container[1] ] Yusa Muroya, “Radiation chemistry of water and water solution”, RADIOISOTOPES vol. 66 (2017）No. 10
https://www.jstage.jst.go.jp/article/radioisotopes/66/10/66_661006/_article/-char/ja[2] Achievement reports for FY2018: Subsidy project of Decommissioning and Contaminated Water Management started since FY2016 “R&D for
Treatment and Disposal of Solid Waste” and Subsidy Project of Decommissioning and Contaminated Water Management started since FY2017 “R&D for
Treatment and Disposal of Solid Waste (R&D for advanced treatment and analysis methods ) (study on waste characterization, storage, management,
disposal and waste stream) https://irid.or.jp/_pdf/20180000_15.pdf?v=2©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Information acquisition<Procedures of information acquisition of high dose waste>(1)
廃棄物リスト①解体廃棄物

List of waste items based on integrated resultsof R&D (waste stream)

S1 圧力容器

S2 格納容器金属

S3 格納容器コンクリート

S4 建屋内金属

S5 建屋内コンクリート

S14 デブリ回収廃棄物

N1 4号機解体廃棄物

N2 ５・６号機解体廃棄物

1

圧力容器

解体金属

解体コンクリート

金属ガラ

解体コンクリート

DSPスロットプラグコンクリート

通常炉相当（S1～S5参照）

通常炉相当（S1～S5参照）

2

蒸気乾燥器

配管類

汚染粉塵（表面除染）

配管類

汚染粉塵（表面除染）

保温材

事故汚染物（S3～5参照）

3

気水分離器

4

上部格子板

5

炉心シュラウド

6

炉心支持板

7

その他機器類8

燃料ｻﾎﾟｰﾄ

9ｼﾞｪｯﾄﾎﾟﾝﾌﾟ10

制御棒案内管

11

CRDﾊｳｼﾞﾝｸﾞ

12 CRD1315

鉄筋

T/B壁（内面）・床・柱等

PCVヘッド高温焼却設備ＢＳＷコンクリート（放射化）R/B機器

T/B壁（外面）

シールドプラグコンクリート

焼却工作建屋

ＰＣＶ本体

ＲＰＶペデスタルコンクリート（表面汚染） R/B高線量機器・配管（DHC系・AC系・FCS）

Rw/B壁（内面）・床・柱等

RPVヘッドサイトバンカ建屋

ＰＣＶ内機器

トーラスコンクリート(表面汚染)

R/B高線量機器・配管（DHC系・AC系・RCW）

Rw/B壁（外面）RPV下鏡※

補助建屋

ＲＳＷ鋼板

ウェルプラグ(表面汚染)

R/B高線量機器・配管（AC系）

デブリ取出し関係機器類

共用プール建屋

ＢＳＷ鉄筋

追加遮蔽材

SPT建屋

ＲＰＶペデスタル鉄筋

タービン本体雑固体焼却設備

発電機本体

17
18

20
21
22

②瓦礫等
S6 瓦礫金属
1 金属がら2 配管（金属）
3 鉄筋
4 解体タンク類
5 ワイヤー類
6 空容器類

置）

4 銀含浸ハーシュライト(AGH)

5 フェロシアン化コバルトハーシュライト(KH)
6 珪チタン酸塩(TSG)
S7 瓦礫コンクリート

S8 可燃物

S15 汚染土壌

7 ろ過砂(硅砂）コンクリートがら

使用済み保護衣

汚染土壌

8

紙・ウェス

礫・採石

酸化チタン(GX194)
難燃物（塩ビ管等）

SSフィルタ(ポリプロピレン製)
減容処理設備プラスチック、ビニール類

浚渫土

不燃物（ガラス，陶磁器等）
フェロシアン化コバルト(Cs-treat)

3号燃料取出用カバー
ウルトラフィルタ(ポリエーテルサルフォン製)

木材

アスファルト

機器類（ケーブル、電池等含む）

SARRY
コンクリート

汚染粉塵（表面除染）
9 チャバサイト(IE-96)

10 珪チタン酸塩(IE-911)
11 珪チタン酸塩(EN-101,HS-726,ET-501)
12 ろ過砂+ゼオライト系:クリノプチロライト
13 ろ過砂+活性炭系:アンスラサイト

13

18
19
20

17

21
22

除染装置スラッジ

蒸発濃縮装置高性能ALPS

脱塩用ＲＯユニット
ろ過フィルタ（ポリプロピレン製）
脱塩器

高分子系系吸着材:水酸化セリウム(ReadE2,Read-B)

Csフィルタ(ナイロン樹脂製)
円筒フランジ型タンク

珪チタン酸塩(IE-911,FST)(放水路浄化)

高分子系系吸着材:イノミジ酢酸型キレート樹脂(IRC748i)

Srフィルタ(ナイロン樹脂製)
横置タンク

珪チタン酸塩(FST)(タンク浄化)

高分子系系吸着材:水酸化鉄系吸着材(CN4000)

角型タンク
RO濃縮水処理設備

ろ過フィルタ（ポリプロピレン製）
固体廃棄物貯蔵庫

キレート樹脂系吸着材:水酸化鉄(Ly-RT)
N5 その他瓦礫等

モバイル型Sr除去装置
スラリー安定化設備

ゼオライト系系吸着材:銀ゼオライト(AgZ）
左記に含まれない全ての瓦礫

伐採木・草

什器類

焼却灰

建屋スラッジ（水抜きに伴い発生）3号燃料取出装置
サブドレン他浄化設備

デブリ取出し関係設備
ろ過フィルタ（ポリプロピレン製）
デブリ一時保管庫・保管庫

Csフィルタ(ナイロン樹脂製)
汚染水貯蔵タンク④その他廃棄物
S16.1 震災前廃棄物（固体廃棄SARRY-2

16 珪チタン酸塩(ET-501)

22

S12 フィルタ
逆浸透膜処理ユニット１

キレート樹脂系吸着材:水酸化鉄(Ly-RT)

1721

除染装置（ＡＲＥＶＡ）

S11 除染装置スラッジ

ラム）

珪チタン酸塩系吸着材(FST,Sr-treat)

15 ろ過砂+活性炭系:アンスラサイト

20

S10.3 多核種除去装置③（処理カ材）

活性炭系系吸着材(GAC,AGC,クラレコール)

14

19

S10.2 多核種除去装置②（吸着炭酸塩スラリー
鉄共沈スラリー

第2モバイル型ストロンチウム除去装置

9

18

リー）

モバイル型ストロンチウム除去装置(KMPS)

珪チタン酸塩(ISM)

KURION

12

16

Rw/B機器・配管・等

S9.2 吸着塔②（モバイルSr除去装 S10.1 多核種除去装置①（スラ（KURION,SARRY）

11

15

第二セシウム吸着装置（ＳＡＲＲＹ）

第三セシウム吸着装置（ＳＡＲＲＹ-2）

S9.1 吸着塔①7 工事資機材（鉄板、鉄パイプ等）
8 車両・重機

油分離装置

高線量配管（MS系）

3 ハーシュライト(H,EH)

10

14

セシウム吸着装置（KURION）

その他機器・配管等

2 界面活性剤改質ゼオライト(SMZ)

1

19

震災後設置

増設固体廃棄物焼却設備

湿分分離器③水処理二次廃棄物

16

震災前設置
集中環境施設プロセス建屋

Dismantled
waste, rubbles
and etc.
Secondary waste generated
from water treatment
ＲＳＷコンクリート（放射化）

鉄筋

給水加熱器

14

N3 その他施設解体廃棄物

機器類(ケーブル、盤等含む)

復水器

廃棄物リストWaste list
(Excluding
spentfuel, fuel debris)（使用済み燃料、燃料デブリは除く）

Srフィルタ(ナイロン樹脂製)
S16.2 震災前廃棄物(サイトバン

物保管施設等）カ、燃料プール等）

除染装置スラッジ抜出し装置

X1 使用済み燃料

X2 デブリ

X3 環境影響物質（分別したもの）

セシウム吸着塔一時保管施設

S13蒸発濃縮装置廃スラリー
蒸発濃縮装置廃スラリー

N4 その他水処理設備
高性能ALPS(2)

珪チタン酸塩(HS-726,FST)

水中で保管されており
Waste
items stored under water,
cut*1 or *1
handledunder water,
水中切断
や水中で取り扱うもの
waste items
containing
水分を含むもの water

活性炭系系吸着材(K-MAC)
高分子系系吸着材(CRB05,CN4000,READ-B,WA30)

ゼオライト系吸着材(AgZ,KW-500G-7)

RO濃縮水処理設備
珪チタン酸塩(HS-72)
サブドレン他浄化設備
ゼオライト系(A51-JHP)
珪チタン酸塩(HS-726)
活性炭系(K-MAC)
高分子系(READ-B)
モバイル式処理装置
珪チタン酸塩(IE-901)(SFP浄化)
珪チタン酸塩(IE-901)(トレンチ浄化)

1 運転廃棄物

造粒固化体（日立製濃縮廃液ペレット）

使用済み燃料デブリ

PCB

2 運転廃棄物（基準不適合or性状不明）

造粒固化体（東芝製濃縮廃液ペレット）

原子燃料（供用中）

デブリ（部材混合）除去金属類（鉛、アルミ等）

3

造粒固化体（東芝製廃樹脂ペレット）

原子燃料（使用前）

4

タンク貯蔵廃樹脂類

損傷燃料

5

タンク貯蔵廃スラッジ類

6

タンク貯蔵廃液類

7

制御棒

8

チャンネルボックス

9

ヒューエルサポート10

中性子検出器11

使用済燃料プール

12

チャンネルボックス

13

ポイズンカーテン14

その他

デブリ（部材付着）

アスベスト含有物
鉛含有物

ゼオライト系(ET201)(トレンチ浄化)

珪チタン酸塩(HS-716)
高分子系(CRB05,WA30)

Yes

水酸化第二鉄系吸着材(As-CATCH2)

Other waste

その他

Excluded from the

調査対象から除外
scope ofinvestigation*1：切断部の冷却による散水は含まない*1: Does not include the water sprayfor cooling the cut portion

高性能ALPS検証試験装置

ゼオライト系(AgZ)

No

ゼオライト土嚢(ゼオライト)
ゼオライト土嚢(活性炭土嚢)

15
16
17
18
19
20
21
22

Waste items*2 that are assumed to be
通常炉の解体廃棄物とは異なる
stored differently than dismantling waste
generatedfrom an ordinary reactor*2
保管が想定されるもの(1) List of waste items
The information is organized based on the list of waste
items. However, information on spent fuel and fuel debris
are not acquired.
(2) Waste required for information
Waste to be required for information acquisition are
selected from the aspect of the following major factor forhydrogen generation.
➢ Water content or not(3) Information acquisition of high dose waste
The following information of selected waste is clarified➢ Properties: Lump-like and particle-like, etc.
➢ Radioactivity: Concentration of radioactivity foreach nuclide➢ Others: Containers to be used for storage, etc.

No

Excluded from the
調査対象から除外
scopeof investigation*2: Assuming that they are stored in*2：通常炉の廃棄体製作に準じて

accordance with the process of
保管されることが想定
creating
waste packets from the wastefrom ordinary reactors.

Yes
Waste pertaining to which
情報収集対象廃棄物
information is to be acquired

Acquisition of information such as
性状・放射能・保管方法案等のproperties, radioactivity, proposed
情報収集storagemethods, etc.

(3)

Acquisition and consolidation of information on
高線量廃棄物の情報収集
整理
high dose waste(Proposed
waste items for which measures for hydrogen gas（水素ガス発生対策要の可能性のある廃棄物案）
generation are likely to be required)©International Research Institute for Nuclear Decommissioning

21

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Information acquisition<Targeted information: 1. Dismantled waste, rubbles and etc.>
Waste list: The amount of hydrogen gas generation is evaluated for Units 1 to 3 assuming contamination that is different from ordinary reactors.
① Dismantling waste
S1 Pressure vessel

S2 Containment vessel metal

S3 Containment vessel
concrete

S4 Metal from inside thebuilding

S5 Concrete from inside thebuilding

N1 Dismantling waste from
S14 Waste from debris collection
Unit 4

N2 Dismantling waste from
Units 5 and 6

N3 Dismantling waste from otherfacilities

Dismantled concrete

Metal scraps

Dismantled concrete

DSP slot plug concrete

Equivalent to ordinary reactor(Refer to S1 to S5)

Contaminated dust (surface
contamination)

Pipes

Equivalent to ordinary reactor(Refer to S1 to S5)
Contaminants from theaccident (Refer to S3 to S5)

Installed before the earthquake
disaster
Central radioactive waste treatmentbuilding
High-temperature incineration
facility
Incineration work-shop building

3

Equivalent
Pressure vesselto L2 waste Dismantled metal
in anordinary
Steam dryer
Pipes
reactor(includes
Steam-water separator L3)
Equipment (including cables, panels, etc.)

RSW concrete (activated)

Rebars4Upper grid plate

Rebars

BSW concrete (activated)

5Core shroud

Main body of PCV

RPV pedestal concrete(surface contamination)

PCV internal components

Torus concrete (surface
contamination)

R/B equipment
R/B high-level radioactive
equipment and pipes (DHC
system, AC system, FCS)R/B high-level radioactive
equipment and pipes (DHC
system, AC system, RCW)

1
2

6Core support plate

7

Other equipment

8

Fuel support

9

Jet pump

Equivalen RSW steel platet to L1
waste in BSW rebars

10 Control Rod Guide Tube

Well plug (surface
contamination)

Equivalent to
L2-L3 waste inan ordinary
reactor

Rw/B walls (inner), floors,pillars, etc.

RPV head

Rw/B walls (outer)

RPV lower hemispherical
dome*

Auxiliary building

Equipment for debris retrieval

Common pool building

R/B high-level radioactive
equipment and pipes (AC system)

On-site bunker building

Additional shielding materialan
RPV pedestal rebars
ordinary
reactor(includes
L2)

SPT building
Miscellaneous solids incineration
facility
Installed after the earthquake
disaster
Additional miscellaneous solid
waste incineration facility
Cesium adsorption apparatus(KURION)
Oil separator
Second cesium adsorption
apparatus (SARRY)
Third cesium adsorption apparatus(SARRY-2)
Decontamination equipment(AREVA)
Reverse osmosis membrane
treatment unit 1
Evaporative concentration
apparatus
RO unit for demineralization
Demineralizer
Cylindrical flanged tank
Horizontal tank
Rectangular tank

Turbine body
Generator

11 CRD housing

Condenser

12 CRD

Moisture separator

13

Waste items circled in
red are studied.

Feed water heater
Other equipment and pipes,
etc.
High-level radioactive pipes(MS system)
Rw/B equipment and pipes,
etc.

Not studied as it involves fuel debris handling

14

Contaminated dust (surface
Heat insulating material
contamination)
T/B walls (inner), floors, pillars
PCV headetc.
T/B walls (outer)
Shield plug concrete

15
16
17

Information on the PCV, turbine and the reactor building concrete, etc.
is not studied because they are estimated to be dismantled in the air.

18
19
20
21
22
② Rubbles, etc.
S6 Metal rubble

S7 Concrete rubble

S8 Combustible materials

S15 Contaminated soil

1

Metal scrap

Concrete scrap

Used protective clothing

Contaminated soil

2

Pipes (metal)

Concrete

Paper and waste cloths

Gravel and quarry

3

RebarsContaminated dust (surface contamination)

Plastics, vinyl, etc.

Dredged soil

4

Dismantled tanks

Wood

Asphalt

5

Wires

Felled trees and grass

6

Empty containers

Incinerated ash

7Equipment and materials required forconstruction work (steel plates, steel
pipes, etc.)

8Vehicles and heavy machinery

Solid radioactive waste storage
facility

N5 Other rubbles, etc.
All types of rubbles not
mentioned on the left
Non-combustibles (PVC pipes,
etc.)
Non-combustibles (glass,ceramics, etc.)

Slurry stabilization equipment
Volume reduction facility
Unit 3 fuel discharge cover

Equipment (including cables,
batteries, etc.)

Unit 3 fuel discharge equipment

Fixtures and fittings
Building sludge (generated
when water is drained)

Facilities related to debris retrieval
Debris temporary storage facility
and storage facility
Storage tanks for contaminatedwater
Equipment for removing sludge
from decontamination equipment
Cesium adsorption vessel
temporary storage facility

9
10
11
12
13
14
15
16
17
18

✓
✓

Information on the waste that includes the reactor pressure vessel and the
DSP slot plug concrete generating during duel debris retrieval is clarified.
Rubbles are not studied because the radiation level is low.

Units 4 to 6 not studied
Not studied as it involves fuel debris handlingNo underwater storage and underwater cutting
Same as waste from ordinary reactors©International Research Institute for Nuclear Decommissioning

22

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Targeted information: 2. Secondary waste generated from water treatment and other waste>③ Secondary waste generated from contaminated water treatment
S9.1 Adsorption vessel ①(KURION, SARRY)
1

2

KURION

S9.2 Adsorption vessel ②
S10.1 Multi-nuclide removal(Mobile Sr removal equipment) system ① (slurry)
Mobile strontium removal
equipment (KMPS)

Surfactant modified zeolite (SMZ)

Silico-titanate (ISM)

Hershelite (H, EH)

Second mobile strontium
removal equipment

4

Silver impregnated hershelite (AGH)

Silico-titanate (IE-911, FST)(For cleaning the drainage
canal)

5

Cobalt ferrocyanide hershelite (KH)

Silico-titanate (FST) (Cleaningof tanks)

6

Silico-titanate (TSG)

7

Filter sand (silica sand)

8

SARRY

3

9

Chabazite (IE-96)

10

Silico-titanate (IE -911)
Silico-titanate (EN -101, HS -726,
ET-501)
Filtration sand + zeolite series:
Clinoptilolite
Filtration sand + activated charcoal
series: Anthracite

11
12
13
14

SARRY-2

15

Filtration sand + activated charcoal
series: Anthracite

16

Silico-titanate (ET -501)

Carbonate slurry

Iron coprecipitation slurry

S10.2 Multi-nuclide removal
system ② (adsorbent)

S10.3 Multi-nuclide removal
S11 Sludge fromsystem ③ (treatment column) decontamination equipment

Chelate resin series
Silico-titanate series adsorbent
adsorbent: ferric hydroxide(FST, Sr-treat)(Ly-RT)
Activated charcoal series
adsorbent (GAC, AGC,
Kuraray coal)
Polymer series adsorbent:
Cerium hydroxide (ReadE2,
Read-B)
Polymer series adsorbent:
Iminodiacetic acid chelate
resin (IRC748i)
Polymer series adsorbent:
Ferric hydroxide series
adsorbent (CN4000)
Chelate resin series
adsorbent: ferric hydroxide(Ly-RT)
Zeolite series adsorbent: Silver
zeolite (AgZ)

Sludge from thedecontamination system

S12 Filter

High performance ALPS

S13 Slurry discharged fromthe evaporative concentration
apparatus
Slurry discharged from theevaporative concentration
apparatus

Cobalt ferrocyanide (Cs-treat)

Silico-titanate (HS -726, FST)

Cs filter (made of nylon resin)

Activated charcoal series
adsorbent (K-MAC)

Sr filter (made of nylon resin)

Polymer series adsorbent:(CRB05, CN4000, READ-B,
WA30)

RO concentrated water
treatment facility

Zeolite series adsorbent (AgZ,
KW-500G-7)

Barrier filter (Made ofpolypropylene)

RO concentrated water
treatment facility

Mobile Sr removal equipment

Silico-titanate (HS -72)
Sub-drain cleaning equipment
Zeolite series (A51-JHP)
Silico-titanate (HS -726)
Activated charcoal series (KMAC)

Cs filter (made of nylon resin)

Polymer series (READ-B)

Sr filter (made of nylon resin)

Mobile treatment apparatus
Silico-titanate (IE -901)(Cleaningof SFP)
Silico-titanate (IE -901)(Cleaning of trenches)
Zeolite series (ET201) Cleaningof trenches)
High-performance ALPS
verification test apparatus
Silico-titanate (HS -716)
Polymer series (CRB05, WA30)
Zeolite series (AgZ)
Ferric hydroxide series
adsorbent (As-CATCH2)
Other
Zeolite sandbag (zeolite)
Zeolite sandbag (activatedcarbon sandbag)

17
18
19
20
21
22

④ Other waste
S16.1 Waste from before theearthquake disaster (Solid waste
storage facility, etc.)

45

S16.2 Waste from before theearthquake disaster (on-sitebunker, fuel pool, etc.)
Granulated solidified
substances (Concentrated
liquid waste pellets
manufactured by Hitachi)
Granulated solidified
substances (Concentrated
liquid waste pellets
manufactured by Toshiba)
Granulated solidified
substances (Waste resin
pellets manufactured by
Toshiba)
Waste resin stored in tanks
Waste sludge stored in tanks

6

Liquid waste stored in tanks

7

Control rods

8

Channel box

9

Fuel support

10

Neutron detector

11
12
13
14
15

Spent fuel pool
Channel box
Poison curtain
Other

1

Waste from operation

2

Waste from operation (nonconformance to standards orproperties unknown)

3

High performance ALPS

Barrier filter (Made ofpolypropylene)

SS filter (Made ofpolypropylene)
Ultra filter (made ofpolyethersulfone)
Sub-drain cleaning equipment
Barrier filter (Made ofpolypropylene)

Titanium oxide (GX194)

N4 Other water treatment
facilities

X1 Spent fuelX2 Debris

X3 Substances affecting theenvironment (segregated)

Spent fuelDebris

PCB

Nuclear fuel (in-service)

Debris (components mixed)

Removed metal (lead,
aluminum, etc.)

Nuclear fuel (before being used)Debris (components adhered)

Substances containing
asbestos

Damaged fuel

Substances containing lead

Waste items circled inread are studied.

Spent fuel, debris (fuel debris)
is not studied
No underwater storage andunderwater cutting
Same as waste from ordinary
reactors✓✓

Information on secondary waste generated from water treatment is selected and clarified.
Wastes generated before the Great East Japan Earthquake are classified as Level 1 waste thatare equivalent to waste generated from the normal reactor.
©International Research Institute for Nuclear Decommissioning

23

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information items for high dose waste1>

Vent

The information of selected items is clarified as below.
➢Properties
Information to be used for studying evaluation model.
✓ Material (metal or concrete or resin, etc.)✓ Shape (block-like or porous or particles less than several mmdiameter)(for dismantled waste) [1]Vent➢Storage conditions
Information on the storage container (anticipated container) andwater content required for evaluation✓ Volume of the storage container (estimated) [m3]Approx.
✓ Volume of stored waste (estimated) ［t]
1.5 m✓ Amount of stored radioactivity per nuclides [Bq]Approx. 2m(calculated from concentration of radioactivity defined from theApprox. 2madvanced inventory estimation and amount of stored radioactivity)(for secondary waste generated from
✓ Amount of sored water content (estimated) [L]water treatment) [2](setting estimated from water content rate, possibility of drytreatment, pre-treatment and etc.)
Fig.1: Storage container for waste(illustration) ）[1]：Final report of Subsidy Project of Decommissioning and Contaminated Water Management in FY 2016 Supplementary
Budget, Advanced Fundamental Technology for Retrieval of Fuel Debris and Reactor Internals[2]：Estimation setting of the storage container based on the information related to secondary waste generated from water
treatment©International Research Institute for Nuclear Decommissioning

24

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information items for high dose waste2>➢ Storage condition of the storage container
Estimation of storage method for dismantled waste<Image of the storage container>

Vent

Inner container
Fig.1: Conceptual image of storage case for the container used for
estimating the amount of hydrogen generation in FY2020

A container model was created considering the shape of waste to begenerated during fuel debris retrieval in FY2020 and the amount ofhydrogen generation was evaluated. In FY2021, the following
estimated items are studied to store in the container.
✓ Metal waste including the reactor pressure vessel, etc.
A detailed cutting method is not studied as demolition methods aredifferent depending on the installation situation. The waste is cut to fit
the size of the storage container under water by remote-operation and7 pieces block-like wastes are assumed to be stored in the storage container✓ Concrete waste to be generated during collecting fuel debris
The concrete waste is cut to fit and store the container by using a wired
saw. Water is assumed to exist in the container as below.
・Adjacent to contaminated water and water impregnates in theconcrete void. (The concrete voids is assumed to be less than 10%. In this case, the void

Assumed issues

Proposal 1: Packaging in dedicatedcontainers and transporting•••

Large dedicated containers are needed
Handling of removed wastes
Concepts pertaining to the safetyfunctions at the time of directly
packaging waste in dedicatedcontainers (dose, contamination,amount of hydrogen generation)

Proposal 2: Cutting the waste to a size thatcan fit in the storage container•••

Proposal 3: Cutting the waste to a size thatcan fit in the retrieval container

Cutting of removed wastes (method
and location)
Handling of removed wastes
Concepts pertaining to the safetyfunctions at the time of directly
packaging waste in storage containers(dose, contamination, amount ofhydrogen generation)• Cutting of removed wastes(method and location)
Can be handledwith the current
collection, transfer
and storage flow forremoved wastes

is conservatively set as 10%.)・Water adhered to the surface when lifting from water after cutting
under water. (The conservative estimation of water thickness is 2mm of the wholesurface although water is not evenly adhered the whole surface.)

・Water using when producing concrete is not considered（Assuming thatwater-cement ratio is around 50. Free water may possibly remain right after concrete
production, however remaining water can be ignored as free water is assumed to evaporate
after time has passed.）➢ Storage volume, voids and water content of wastei.

The following conditions are set for the waste.
Storage ratio is set based on track records 200 liter drumof JPDR demolitions [1])Metal

1m3 steel container✓✓

Handling at the storage vault needs to be studied
Detailed studies shall be carried out as needed in accordance with theprogress of the studies on retrieval methods

It is assumed that waste is cut and stored to fit
the size of the storage container.
Fig. 2: Handling flow of large waste studied in FY2020[1]
Kunio Shiraishi, Takenori Sukekawa and Satoshi Yanagihara, (Japan Atomic Energy Research Institute)“, Dismantled Waste
Data Analysis of the Japan Power Demonstration Reactor (JPDR) (Commissioned Research),” JAERI-Data/Code 99-050.

Metal waste: 20vol% (conservative setting
Concreterather than the record)
3m steelcontainer
Concrete waste: 50vol%Flexibleii. Storage volume of waste is set based oncontainer
volume inside the container and its
Filling factor (%)
proportion.
iii. Water volume generated from the waste is
the isopen
literatures.
*:set
Thebasedinternal on
volumeset assumingthat thickness of the shield is 60mm. (Inner3

dimensions: 1.3m×1.3m×0.88m=1.5m3）©International Research Institute for Nuclear Decommissioning

25

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information items for high dose waste3>
Assuming a storage method of secondary waste generated from the water treatment
Material: SS400
Inner surface: Polyethylene lining
Outer surface: Resin coating

Vent

Capacity: Approx. 5.1m3
Approx.
1.5m

Iron coprecipitation

Approx. 2mApprox. 2m

Carbonate

Fig.1: Waste storage container(assumed) (Secondary waste
generated from the water treatment［1］Filtered water

A filter press dehydrates slurryfor dehydration.

Likely to change after basic designing

Feedwater tankAnti-dispersing cover

Flushing pump

Filter press machine

Water for cleaning
filter cloth

Back-up
extraction unit

Unit cleaning
equipment

Slurrytank

SlurryDischarge ofdehydrated productFiltrate

Slurrytank

Return of filter clothcleaning water
Return of
backwashing water(Cleaning of internal structures,
extraction unit, etc.)

Return of filtrate and filter
plate cleaning water
Extraction of waterfrom compression

Cleaning unit

Diaphragm
pump

Backwashing water

Water fromcompression

SlurryHighpressure
washerSlurry return line

Filling
Closing of lid
Storagecontainer

Dehydration
product

Transfer

Dehydration
product

Pump forwater fromcompression

Pump for extraction of waterfrom compression

Liquid waste transfer pump

Tank forwater fromcompression

Slurry pump

Feed watertank
Liquid wastetransfer pump

(ALPS treatedwater)

Liquid wastetank

Liquid waste
tank
Filtratetank

ALPS transfer

Cleaning
water tankFeed water pump

Filterunit

Liquidwaste
Liquid waste transfer pump

➢ Storage volume, voids and water content of waste
The following conditions are set for the waste.
Volume inside the container*: Assuming that thickness of the shield is 50mm, inner volume is set.
(Inner diameters: 1.9m×1.9m×1.4m=5.1m3)

i.
Compressed air

Waste water from
HIC cleaning
Waste water from
HIC cleaning➢ Storage condition of the storage container
Secondary waste generated from water treatment is assumed to be
stored based on results studied by Tokyo Electric Power Company(TEPCO) Holdings, Inc. as below.
✓ Slurry (carbonate and iron co-precipitation):
After the filter press dehydration treatment, slurry is assumed to be
stored in the storage container described in the left figure.
✓ Waste adsorbent:
It is assumed that the waste is removed from HIC and stored in thestorage container described in the left figure. Adsorbent is dehydrated
through dehydration equipment during storage in HIC therefore,
additional dehydration is not performed during replacement.

Storage ratio:
Slurry: 80vol.% (based on the result studied by TEPCO HD.)

Waste adsorbent: 80vol.% (regarded as same as slurry
because of insufficient information)
ii. Voids
Setting by inner volume of the container and storage volume ofwasteiii. Water content
Slurry: 60vol.% (based on the result studied by TEPCO HD.)
Waste adsorbent: 50wt. %(based on properties of long-term storage spent resin of JPDR[3])[1]：Assuming of the storage container based on the information related to secondary waste
generated by water treatment.
[2]：The figure was created by reference to information provided by TEPCO HD.
[3] Kato, etc. “ Disposal of spent ion replaced resin during long-term storage in JPDR andtheir investigation of the physical property,” Journal of Atomic Energy Society of Japan, vol.22,
No.6、pp.386 -391(1980)

Fig. 2: Treatment process of secondary waste for water treatment (slurry)［2］©International Research Institute for Nuclear Decommissioning

26

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information items for high dose waste4-1>
Nuclides required to evaluate hydrogen generation➢ Conventional evaluation✓ When evaluating hydrogen generation from high dose waste caused by accidental contamination, Cs-137 is regarded
as a target nuclide (evaluation for FY2020). ⇒ Other nuclides except Cs-137 are not regarded not to contributecontamination.
✓ According to analysis results of heat value of fuel debris (fuel component and fuel cladding tubes), heating values(energy) of β-, γ- and α-rays can be evaluated (evaluation for FY2020) .⇒ The heat value of fuel debris cannot be
used to evaluate high dose waste because there are differences in heat values (energy) of contamination between
accidental contamination adding to dismantled waste and secondary waste generated from water treatment.
Current data is used to evaluate wide
➢ Concepts of the evaluation
variety of nuclides.
✓ Hydrogen generation of high dose waste is evaluated considering
nuclides except Cs-137 contributing to heat generation.
✓ Nuclides for evaluation of hydrogen generation are selected based on evaluation results of heat value of fuel debris.
✓ The concentration of selected nuclides is evaluated from radioactivity concentration data for each waste.
⇒Nuclides that are used for evaluation are selected from both of the aspects of heat generation (absorption energy) andradioactivity concentration of waste. The heat generation rate for β-, γ- and α-rays is set.
a. According to evaluation results of fuel heat value evaluated by the Japan Atomic Energy Agency (JAEA) [1],nuclides that contributes to hydrogen generation (absorption energy is large) are selected.
b. According to 10-year-attenuation radiation concentration data that is back data of inventory estimation results [2] foreach waste (waste generated from demolition, fuel collection and water treatment) evaluated by JAEA and theCentral Research Institute of Electric Power Industry (CRIEPI), high-level radioactive nuclides are selected.
c. Nuclides that are attributed to evaluation on hydrogen generation from the aspect of (a) the heat generation rateand (b) concentration of radioactivity are covered and then the heat generation rate of β-, γ- and α-rays are set
based on multiplication of concentration and decay energy.
[Supplement]
With regard to target dismantled waste (metal and concrete), when evaluating hydrogen generation caused by nuclides that are not accidental andactivated-/-secondary contamination nuclides, Co-60 must be added in evaluation target nuclides in accordance with “Requirements for the marginal depth
disposal waste and its inspection method: 2015”(AESJ-SC-F014:2015).
[1] JAEA-Data/Code 2012-018, Evaluation on fuel composition for Fukushima Daiichi Nuclear Power Station, https://jopss.jaea.go.jp/search/servlet/search?5036485[2] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka, Yoshikazu Koma & Youko Takahatake (2021) Stochastic estimation of radionuclide composition in wastes
generated at Fukushima Daiichi nuclear power station using Bayesian inference, Journal of Nuclear Science and Technology, 58:4, 493-506,©International Research Institute for Nuclear Decommissioning

27

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition➢ From the aspect of the heat generation rate
Aging variation of the heat generation rate of the Unit 3
reactor core fuel at the Fukushima Daiichi Nuclear Power
Station (NPS) evaluated by JAEA[1] describes in the
figure classified by a) Radiated material, b) Actinides andc) Fission products.
The contribution to the whole heat generation rate (referto the right figure and table) is shown Fission products >Actinides > Radiated material in an early stage.
Compared with Actinides that has longer half-life time,
the heat generation rate of the fission product decreases
and it becomes Actinides > Fission product > Radiatedmaterial 50-100 years later.
Among the nuclides that have passed 10-50 years, major
nuclides (top five nuclides) that generates a large amountof heat were evaluated from the data of fuel compositions
as below.
・Radiated materials: Sb-125,Te-125m,Co-60,Ni-63,Fe-55
・Actinides: Pu-238,Am-241,Cm-244,Pu-240,Pu-239
・Fission products: Y-90,Ba-137m,Cs-137,Sr-90,Cs-134

Radiated material
Actinides
Fission product

Generation amount rate (W/core)<Information items for high dose waste4-2>
Nuclides required to evaluate hydrogen generation

Time period (year)

Fig.1: Aging variation of the heat generation rate of the Unit 3
reactor core fuel at Fukushima Daiichi NPSHeat generation rate（W/core）
Time period

10-year20-year50-year

Radiatedmaterial

1.10E+01

1.01E+00

4.16E-02

Actinides

1.31E+04

1.42E+04

1.47E+04

Fission product 5.37E+04

3.95E+04↓

1.91E+04

Heat generation rate（W/g）
Time period

10-year20-year50-year

Radiatedmaterial

8.78E-08

8.01E-09

3.31E-10

Actinides

1.04E-04

1.13E-04

1.17E-04

Fission product 4.27E-04

3.14E-04

1.52E-04[1] JAEA-Data/Code 2012-018 “Evaluation on fuel composition for Fukushima Daiichi NPS” https://jopss.jaea.go.jp/search/servlet/search?5036485©International Research Institute for Nuclear Decommissioning

28

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information items for high dose waste 4-3>
Nuclides required to evaluate hydrogen generation➢

Nuclides of dismantled waste

In reference to back data of the inventory estimation results estimated by
JAEA/ CRIEPI[1], the concentration of radioactive nuclides which half-life
period have passed 10 years is accumulated and target nuclides for
estimation are the ones that the accumulated concentration is up to 99%.
Table 1 Nuclides from the reactor internal structures(equivalent to L2)<Assumed to be the same as the pressure vessel
studied>No.
順位

Nuclide
核種Type of
線種
radiation

上位%

Cumulative
concentration➢

Nuclides of secondary waste generated from water
treatment

Major nuclides that are collected from water treatment ofeach slurry and sludge are clarified.
Table 3 Nuclides selected from the secondary waste generated from water treatment

Table 2 Nuclides from the containment vessel concrete (equivalent
to L2)<Assumed to be the same as the concrete studied>No.
順位

Nuclide
核種Type of
線種
radiation

核種
Nuclide

AREVA
AREVA
sludgeスラッジ

炭酸塩
Carbonateslurry
スラリー

Iron
co鉄共沈
precipitation
slurryスラリー

Cs-137

Sr-90

Sr-90

Sr-90

Sr-90

Cs-137KURION

SARRY

Cs-137
Sr-90

上位%

Cumulative
concentration

Silver銀ゼオライトzeoliteH-3

1 Cs-137 β、γ39.21%

1 Cs-137 β、γ52.37%

2 Pu-241 α

65.00%

2 Pu-241 α

72.73%

Cs-134

Cs-134

3

88.76%

3

89.67%

Pu-241

Pu-241

Pu-241

Pu-241

Pu-241

Pu-238

Pu-238

Pu-238

Pu-238

Pu-238

Pu-238

Cm-244

Cm-244

Cm-244

Cm-244

Cm-244

Cm-244

Pu-240Pu-240Pu-240Pu-240Pu-240Pu-240

Sr-90

β

Sr-90

β

4 Pm-147 β

94.09%

4 Pm-147 β

94.14%

5 Cs-134 β、γ95.73%

5 Cs-134 β、γ96.30%

6 Pu-238 α

96.57%

6 Pu-238 α

97.00%

7 Eu-154 β、γ97.40%

7 Sb-125 β、γ97.68%

8 Sb-125 β、γ97.91%

8 Eu-154 β、γ98.30%

9 Cm-244 α

98.30%

9 Cm-244 α

98.62%

10

Co-60

β、γ

98.56%

10β、γ98.84%

11 Eu-155 β、γ98.79%

11 Eu-155 β、γ99.01%

12 Pu-240 α

98.98%

13

99.17%

Fe-55

β、γ

Co-60

β、γ

α

核種
NuclideSilico-titanate
珪チタン酸塩
フェロシアン
Cobaltβ、γIminodiacetic

Titanium
酸化チタン
oxide

Ferric
水酸化鉄hydroxide

H-3

H-3

H-3

H-3

Pu-241

Pu-241

Pu-241

Pu-241

ferrocyanide
化コバルト

H-3

Pu-241

イノミジ酢酸resin

series系吸着材
adsorbentacid chelate型キレート樹脂

Sr-90
Pu-241

αSr-90

Pu-238

Pu-238

Pu-238

Pu-238

Pu-238

Cm-244

Cm-244

Cm-244

Cm-244

Cm-244

Pu-240Pu-240Pu-240Pu-240Pu-240➢ Nuclides required to evaluate hydrogen generation (plan)
Black letter: Nuclides selected based on the heat generation evaluation of fuel compositions
Blue letter: Nuclides added the above
Radiated material：Sb-125,Te-125m,Co-60,Ni-63,Fe-55,H-3,Eu-154,Eu-155
Actinides：Pu-238,Am-241,Cm-244,Pu-240,Pu-239,Pu-241
Fission products：Y-90,Ba-137m,Cs-137,Sr-90,Cs-134, Pm-147[1] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka,
Yoshikazu Koma & Youko Takahatake (2021) Stochastic
estimation of radionuclide composition in wastes generated at
Fukushima Daiichi nuclear power station using Bayesianinference, Journal of Nuclear Science and Technology, 58:4,
493-506,©International Research Institute for Nuclear Decommissioning

29

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information acquisition of high dose waste (investigation method)>
Selected from list of thewaste list that was
integrated from R&D results(the waste stream)

廃棄物性状
廃棄物区分

解体廃棄物

保管状況

廃棄物名称

総放射能濃度[Bq/t]

材質

Estimation of storage
conditions, storage volume,
void rate, water content
rate and etc.

形状

現在

保管前処理（計 保管前処理（想 保管容器仕様 容器内容積（想 廃棄物収納率
画）
定）
（想定）
定）[m3][vol%]

保管容内状況（想定）
廃棄物収納量収納放射能量
廃棄物収納量[t]
[m3]
[Bq]

空隙量
[m3]

水分量
[L]

備考
（想定根拠等）

圧力容器
蒸気乾燥器
気水分離器

解体廃棄物

DSPスロットプラグコンクリート
保温材
PCVヘッドシールドプラグコンクリート
RPVヘッドRPV下鏡※
デブリ取出し関係機器類

水処理二次廃棄物

界面活性剤改質ゼオライト(SMZ)
ハーシュライト(H,EH)
銀含浸ハーシュライト(AGH)
フェロシアン化コバルトハーシュ
ライト(KH)
珪チタン酸塩(TSG)
・・・・
・・・・

Information on studyof evaluation model

Analysis results or the inventory
estimation results evaluated by
JAEA and CRIEPI©International Research Institute for Nuclear Decommissioning

30

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Information acquisition of high dose waste(investigation status)>
廃棄物区分

解体廃棄物

解体廃棄物

廃棄物性状
廃棄物名称

大区分

中区分
S1 圧力容器S14 デブリ回収廃棄物

水処理二次廃棄物

S9.1 吸着塔①（KURION,SARRY） KURIONSARRYSARRY-2

S9.2 吸着塔②（モバイルSr除去装置） モバイル型ストロンチウム除去装置(KMPS)
第2モバイル型ストロンチウム除去装置

S10.1 多核種除去装置①（スラリー）

S10.2 多核種除去装置②（吸着材）

形状

圧力容器

金属

塊状

蒸気乾燥器

金属

塊状

気水分離器

金属

塊状

DSPスロットプラグコンクリート

コンクリート

多孔状

保温材

金属

多孔状

PCVヘッド

保管状況

総放射能量*(イン

金属

塊状

現在

10年減衰[Bq/t]

9.10E+13

1.27E+14
9.10E+13
1.27E+14－切断

外寸1.42m×1.42m×1mh－

切断

廃棄物保管容器－切断（ベント付き）

据付中

－

切断

据付中

－

切断

据付中

－

多孔状

金属

塊状

RPV下鏡※

金属

塊状

デブリ取出し関係機器類

金属

塊状

界面活性剤改質ゼオライト(SMZ)

焼結体

粒状(数mm径)

ハーシュライト(H,EH)

焼結体

粒状(数mm径)

銀含浸ハーシュライト(AGH)

焼結体

粒状(数mm径)

フェロシアン化コバルトハーシュライト(KH)

焼結体

粒状(数mm径)

珪チタン酸塩(TSG)

焼結体

粒状(数mm径)

ろ過砂(硅砂）珪砂・活性炭粒状(数mm径)

チャバサイト(IE-96)

焼結体

粒状(数mm径)

珪チタン酸塩(IE-911)

焼結体

粒状(数mm径)

珪チタン酸塩(EN-101,HS-726,ET-501)

焼結体

粒状(数mm径)

ろ過砂+ゼオライト系:クリノプチロライト

焼結体

粒状(数mm径)

ろ過砂+活性炭系:アンスラサイト

珪砂・活性炭粒状(数mm径)

ろ過砂+活性炭系:アンスラサイト

珪砂・活性炭粒状(数mm径)

情報なし

専用吸着塔保管情報なし

専用吸着塔保管(ベント

珪チタン酸塩(ET-501)

焼結体

粒状(数mm径)

情報なし(ベント孔あり)

情報なし

孔あり)のまま

珪チタン酸塩(ISM)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(IE-911,FST)(放水路浄化)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(FST)(タンク浄化)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

炭酸塩スラリー

炭酸塩

粒状(数mm径)

7.72E+11

HIC保管

フィルタープレス－

鉄共沈スラリー

水酸化鉄

粒状(数mm径)

7.88E+12

HIC保管

フィルタープレス－

珪チタン酸塩系吸着材(FST,Sr-treat)

焼結体

粒状(数mm径)

5.53E+10

HIC保管

活性炭系系吸着材(GAC,AGC,クラレコール)

活性炭

粒状(数mm径)

4.81E+10

HIC保管

高分子系系吸着材:水酸化セリウム(ReadE2,Read-B)

樹脂

粒状(数mm径)高分子系系吸着材:イノミジ酢酸型キレート樹脂(IRC748i)

樹脂

粒状(数mm径)高分子系系吸着材:水酸化鉄系吸着材(CN4000)

樹脂

粒状(数mm径)

キレート樹脂系吸着材:水酸化鉄(Ly-RT)

樹脂

粒状(数mm径)

1.26E+11

HIC保管

ゼオライト系系吸着材:銀ゼオライト(AgZ）

焼結体

粒状(数mm径)

4.81E+10

HIC保管

酸化チタン(GX194)

焼結体

粒状(数mm径)

4.70E+10

HIC保管

20

1.2

20

50

0.74

1.8

0.74

88

50

0.74

1.8

1.2

20

1.4872

20

0.3

2.34

1.2

20

(約1.5)

50

0.74

1.8

20

0.3

2.34

0.74

88 DSPスロットプラグコンクリートと同一設定

1.2

20

別途整理

圧力容器廃棄物（事故前L2相当+滞留水汚染）の10年後減衰放射能濃度を想定
比重7.8とし、L1廃棄体2.9t収納時の表面付着水等：約4~20kgから20kg（学会標準参照）
格納容器内廃棄物（コンクリート）（事故前L2相当+滞留水汚染）の10年後減衰放射能濃度を想定
比重2.4とし、内空隙10%全てに水浸漬, □1.2m×0.8m高≒6.8m2全面に2mm厚の水(約14L)が付着と仮定
圧力容器と同一設定

2.34

1.2

20 圧力容器と同一設定

0.3

2.34

1.2

20

（追加遮蔽あ情報なしり）（ベント付き）専用吸着塔保管(ベント孔あり)のまま（情報無し）保管容器移し替え後の保管は想定されていない
2.11E+13

情報なし
2.30E+10
情報なし

専用吸着塔保管(ベント孔あり)

情報なし

専用吸着塔保管(ベント孔あり)のまま

樹脂

多孔状

貯槽D

2900

ベントフィル付保管容器

HIC保管
HIC保管

淡水置換後水切りをして保管（含水率約50％)外寸2×2×1.5mh

HIC保管

Csフィルタ(ナイロン樹脂製)

5.1

80

(遮蔽5cmを想定）
情報なし

処理なし

遠心分離

－

4.1

5.8

(脱水物)

（脱水物）

ー

SARRYの実績を基に容器収納率、脱水物のかさ密度等より設定

1.0（含水率約50％)（含水率約60％)

ALPSスラリーと同等と推定

コンクリートまたは金属製の保情報なし

管容器（ベント
孔あり）に収納情報なし

現行の保管容器(ベント孔あり)のまま

Compiled list of acquired information
ウルトラフィルタ(ポリエーテルサルフォン製)

樹脂

多孔状

ろ過フィルタ（ポリプロピレン製）

樹脂

多孔状

Csフィルタ(ナイロン樹脂製)

樹脂

多孔状

Srフィルタ(ナイロン樹脂製)

樹脂

多孔状

蒸発濃縮装置廃スラリー

炭酸塩

粒状(数mm径)

して瓦礫エリア（情報無し）保管容器移し替え後の保管は想定されていないで保管

横置きタンク

情報なし（ベント管付情報なしき）

現行のタンク

(ベント管あり)のまま

珪チタン酸塩(HS-726,FST)

焼結体

粒状(数mm径)

情報なし

HIC保管

活性炭系系吸着材(K-MAC)

活性炭

粒状(数mm径)

情報なし

HIC保管

樹脂

粒状(数mm径)

情報なし

HIC保管

情報なし

処理なし

高分子系系吸着材(CRB05,CN4000,READ-B,WA30)

ゼオライト系吸着材(AgZ,KW-500G-7)

焼結体

粒状(数mm径)

情報なし

HIC保管

RO濃縮水処理設備

珪チタン酸塩(HS-72)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

サブドレン他浄化設備

ゼオライト系(A51-JHP)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(HS-726)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

活性炭系(K-MAC)

活性炭

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

高分子系(READ-B)

樹脂

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(IE-901)(SFP浄化)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(IE-901)(トレンチ浄化)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

ゼオライト系(ET201)(トレンチ浄化)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

珪チタン酸塩(HS-716)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

高分子系(CRB05,WA30)

樹脂

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

ゼオライト系(AgZ)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

水酸化第二鉄系吸着材(As-CATCH2)

焼結体

粒状(数mm径)

情報なし

専用吸着塔保管

情報なし

処理なし

その他

？

情報なし

情報なし

情報なし

処理なし

ゼオライト土嚢(ゼオライト)

焼結体

粒状(数mm径)

情報なし

未回収

情報なし

処理なし

ゼオライト土嚢(活性炭土嚢)

焼結体

粒状(数mm径)

情報なし

未回収

情報なし

処理なし

高性能ALPS検証試験装置

20

1.2

0.3

HIC保管

モバイル式処理装置

1.2
別途整理

20

多孔状

高性能ALPS

2.34

20

樹脂

多孔状

2.34

0.3

切断

ろ過フィルタ（ポリプロピレン製）

多孔状

2.34

0.3

20

切断

HIC保管

多孔状

0.3

20

－

8.17E+12樹脂

20

1.4872(約1.5)

備考（想定根拠等）－

1.26E+11

樹脂[L]－(今後導入)4.65E+10

樹脂

水分量

[m3]

据付中

粒状(数mm径)

SSフィルタ(ポリプロピレン製)

空隙量

[Bq]

切断

粒状(数mm径)

ろ過フィルタ（ポリプロピレン製）

収納放射能量[t]

切断

粒状(数mm径)

Srフィルタ(ナイロン樹脂製)

廃棄物収納量[m3]－

硫酸バリウムモバイル型Sr除去装置

廃棄物収納量[vol%]－

樹脂

RO濃縮水処理設備

廃棄物収納率

据付中専用吸着塔保管
1.27E+13

容器内容積（想定）[m3]

据付中

焼結体

その他水処理設備

廃棄物保管容器

コンクリート

除染装置スラッジ

N4

切断

RPVヘッドキレート樹脂系吸着材:水酸化鉄(Ly-RT)

S13蒸発濃縮装置廃スラリー

外寸1.42m×1.42m×1mh

シールドプラグコンクリート

9.10E+13

保管容器仕様（想定）据付中

フェロシアン化コバルト(Cs-treat)

サブドレン他浄化設備

保管前処理（想定）据付中S11 除染装置スラッジ
高性能ALPS

保管前処理（計画）据付中

S10.3 多核種除去装置③（処理カラム）

S12 フィルタ

保管容器内状況（想定）

ベントリ推算高度化)

材質

小区分

外寸2×2×1.5mh
ベントフィル付保管容器

6

80

4.8

ー

ー

1.2

（含水率約50％)

現在、休止中©International Research Institute for Nuclear Decommissioning31

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Selection of waste required for hydrogen generation management during storage>➢ Selection of the following waste that could possibly generate hydrogen✓ Dismantled waste that has to be stored and handled underwater, or cut underwater and water islikely to enter into the container.
✓ Waste that has not been contained and stored in the container (except waste that has been stored in thecontainer with vent)➢ Selected waste and waste objects for evaluation
The following waste among the selected waste is evaluated. In case that waste has different conditionsin the same classification, waste is selected from the aspect of materials, the amount of radioactivity inthe storage container and water content.
Table 1: Waste that is assumed to require hydrogen management and waste objects to evaluate hydrogen generation
Waste classification required for hydrogen management
Primary
classification
Dismantled
waste

Secondary waste
generated fromwater treatment

Classification

Type of waste

Waste objects for evaluation

Reason for selection

Reactor pressure vessel
(RPV)

RPV
Steam dryer
Steam separator

Metal waste including RPV, etc.

Metal waste contamination is derived from theaccident, radiated from base material, and secondary
contamination (equivalent to L2 waste before theaccident).

Waste generated during
fuel debris retrieval

DSP slot plug concrete
Heat insulating material
PCV/RPV heads, etc.

Concrete waste including DSP
slot plug concrete, etc.

Concrete waste (equivalent to L2 waste before theaccident and assuming that contaminated waste isderived from the accident,) and it has possibly highwater content.

Mobile Sr. removal
equipment (adsorption
tower)

Silicate titanium salt－

(No information on concentration of radioactivity)

Multi-nuclide removal
equipment (slurry andadsorption tower, etc.)

Carbonate slurry
Ion coprecipitation slurry, etc.

Other water treatment
facility

High-performance ALPS, RO
concentrated water treatment facility, etc.
(silicate titanium salt, activated carbon
and zeolite, etc.)

Carbonate slurry generated
from multi-nuclide removal
equipment－

High level of radioactivity among secondary waste
generated from water treatment and the amount ofgeneration is large.
(The concentration is assumed to be lower thanabove although there is no information on theconcentration of radioactivity.)

Issues concerning hydrogen generation of secondary waste generated from water treatment during storage can be studied based on evaluationof multi-nuclide removal equipment and carbonate slurry.
©International Research Institute for Nuclear Decommissioning32

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Study of factors for hydrogen generation>

Study on the following factors and the evaluation model effecting evaluation of hydrogengeneration rate.
✓ Type of radiation (G value)✓ Radioactivity inventory✓ The amount of heat generation and energy absorption ratio✓ Water content rate
Type of radiation (G value)
Nuclides for evaluation are mainly radiated materials and fission products that emit β- and γ-rays. On theother hand, Actinides are α-ray emitting nuclide.
β- and γ- rays are regarded as equivalent from the aspect of water radiolysis therefore, the common G
value is used. Fuel debris is assumed based on the previous case reports and G value that was used forevaluation of mixing α- and γ- rays is used [1].
Table 1: G value to be used for evaluation[1] Junichi Takagi and Seiji Yamamoto “ Simulation study on hydrogen generation by the alpha radiolysis from the fuel debris particles ”
Proceedings Symposium on Water Chemistry and Corrosion in Nuclear Power plants in Asia - 2017 (AWC 2017)©International Research Institute for Nuclear Decommissioning

33

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Impact parameter setting 1>
Radioactivity inventoryRadioactivity inventory (Bq)= Radioactivity per a unit weight (Bq/t)
×Waste weight (t)・Radioactivity per a unit weight (Bq/t)
Concentration setting of 10-year-decayed radioactivity
based on a maximum concentration of each waste(demolition, debris collection and water treatment)
according to the result of the inventory evaluated by
Japan Atomic Energy Agency (JAEA) and Central
Research Institute of Electric Power Industry (CRIEPI) [1].
・Waste weight (t)Volume inside the container (m3)×Filling rate (－)×
Waste density (t/m3)
Mass setting for each waste is as below.
・Reactor Pressure Vessel (metal waste) : Weight 2.34 ton*・Waste generating during fuel debris retrieval (concrete): Weight 1.8 ton **・Secondary waste generated from water treatment: Weight 2.65 ton ***Setting of volume inside the container 1.5m 3ｘFilling rate 20vol%ｘDensity7.8 ｔ/m3＝2.34 ton
Setting of volume inside the container 1.5m3ｘFilling rate 50vol%ｘDensity 2.4 ｔ/m3＝1.8 ton*** Setting of volume inside the container 5.1m 3ｘFilling rate 80vol%=Waste volume 4.1m 3、4.1m3ｘDensity 1.616 ｔ/m3ｘ0.4＝2.65 ton（0.4（＝100wt％-60wt %) ; Carbonate slurry with water content rate 60xt% by filter
pressing is assumed.
***

Fig.1: Example of the radioactivity inventory evaluation (concentration inside thecarbonate slurry) [1][1] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka, Yoshikazu Koma & Youko Takahatake (2021) Stochastic estimation of radionuclide composition in wastes generated
at Fukushima Daiichi nuclear power station using Bayesian inference, Journal of Nuclear Science and Technology, 58:4, 493-506,©International Research Institute for Nuclear Decommissioning34

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste
Information acquisition<Impact parameter setting 2>

The amount of heat generation and the energy absorption ratio
The amount of heat generation (W) ＝ The amount of heat generation per radioactivity of unit nuclide(W/Bq)×Radioactivity inventory (Bq）
According to 1) the number of Bq per a nuclide (Bq/core) and 2) the amount of heat generation per a nuclide (W/core) reported
in “Fuel composition evaluation of the Fukushima Daiichi Nuclear Power Station” (JAEA-Date/Code２０１２-018） by JAEA, theamount of heat generation per 1Bq of each nuclide (W/Bq) was calculated, and multiplying the calculation result gives by theradioactivity inventory described above. The calculation result of the amount of waste heat generation is given in the below table 1.
The inventory value is a maximum value of the evaluation. On the other hand, another evaluation using values of 95% tile and amode value was also conducted as evaluation values of secondary waste generated from water treatment widely vary.
Most of radiation energy is known to generate heat. Additionally, early composition products becomes water due to the binding
reaction therefore, some parts of the products contribute to decomposition. [1] Although there is no specific report case, 10% of theheat generation assumed to contribute to radiation decomposition for evaluation in this project.
Table 1: Type of waste and the amount of heat generation

Type of waste

The amount of heat generation[W]βγ nuclides

α nuclide

Reactor Pressure Vessel (metal waste)

MAX

7.96

6.98

Waste generated from collecting fuel debris(concrete)

MAX

10.69

9.82

MAX

31.2

25.2

(Reference)
95%ile

8.65

0.225

(Reference)
Mode

0.673

Secondary waste generated from water
treatment

1.01E-4

Considering heatgenerated from α nuclide.

[1]Shinichi Ohno “Hydrogen generation by water radiolysis”, RADIOISOTOPES Vol.29,No.8,P.401（1980)©International Research Institute for Nuclear Decommissioning

35

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Study of factors required for evaluation of hydrogen generation<Impact parameter setting 3>
Water content rate (water volume inside the container)
Water in the container can be source of hydrogen generation. Water volume inside the container is set
for evaluation in reference to the following setting.
➢ In case of dismantled waste✓ Reactor Pressure Vessel (RPV) (metal waste): Water adhered on the surface✓ Waste generating from collecting fuel debris (concrete): Free water (assuming that water isimpregnated at all voids inside) and adhered water.
➢ In case of secondary waste generated from water treatment
Secondary waste generated from water treatment (carbonate slurry): 60vol.% (according to theresult of study by TEPCO HD.)
➢ Water content rate of each waste (water volume inside the container) is set as below.
✓ RPV(metal waste)：20L*✓ Waste generating from collecting fuel debris (concrete) ：88L**✓ Secondary waste generated from water treatment:3980L**** Manufacturing requirements and inspection method for the marginal disposal waste in accordance with the Atomic Energy Society ofJapan standard（AESJ-SC-F014:2015）: Setting as 20L (adhered water on the surface when storing the L1 waste: Approx. 4-20kg)** Setting as 88L that totals 10vol% (74L) of voids inside the container and 2mm-thickness in the whole surface of the inner wall surface(14L).
*** Setting as filling rate 80vol% of the storage container and water content rate 60wt% of carbonate slurry by filter pressing.
(Volume inside the container 5.1m3ｘ Filling rate 80vol%= Waste volume 4.1m3 ,4.1m3ｘ Density1.616 ｔ/m3ｘ Water content rate 0.6＝
3980L）©International Research Institute for Nuclear Decommissioning36

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Study on factors required for evaluation of hydrogen generation<Hydrogen generation evaluation model>
Evaluation model for calculating hydrogen generation rate of high dose waste
・In case of dismantled waste (metal and concrete waste)
Waste and water have existed respectively, which are relatively separated. Setting of water and gas volume (air) depends onvolume of waste in the container. The water absorption rate of radiation energy generating from waste depends on type and state ofradiation.
・In case of secondary waste generating from water treatment
Conditions are set of homogeneously existing fine particles including slurry and adsorbent, and water. Water volume is set from thewater content rate depending on waste volume in the container. The water absorption rate of radiation energy generating from waste
is set depending on type of radiation.

Dismantled waste

Void ○m3

Secondary waste generated fromwater treatment
Void 1m3

Water △m3
Water: Approx. 4t
Waste □m3⇒■ton(◆Convert Bq to ▼W)

Filter (dehydration
products) : 2.65 t(Convert Bq to W)

Total volume: Approx. 1.5 m3

Total
volume:Approx. 6m3

Fig.1: Conceptual image of the evaluation model for calculation of hydrogen generation rate
©International Research Institute for Nuclear Decommissioning37

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Evaluation result of hydrogen generation 1>
2) Evaluated by the linear model(1 mol of hydrogen is estimated to be equivalent to 22.4L.
When hydrogen generates at a constant speed, all amount ofhydrogen transfers to the gas phase and concentration ofhydrogen increases.

1) Evaluated by the radiolysis model

1.E+02
1.E+02

RPV(metal waste)
圧力容器（金属廃棄物）

1.E+00
1.E-01
圧力容器（金属廃棄物）
RPV (metal waste)
燃料デブリ回収時に発生する廃Waste
generating from collecting fuel
デブリ回収廃棄物（コンクリート）
debris(concrete)
棄物（コンクリート）
Secondary waste generated from water
水処理二次廃棄物

1.E-02treatment

Secondarywaste generated from water
水処理二次廃棄物95%
treatment 95%

1.E-03

Secondarywaste generated from water
水処理二次廃棄物mode
treatment mode

1.E-04
1.E+00

1.E+02

1.E+04

1.E+06

1.E+081.E+10Time (second)Fig.1: Relationship between concentration and time of
the gas phase hydrogen (evaluation by the radiolysis)

Concentration of the gas phase hydrogen (%)

Concentration of the gas phase hydrogen (%)

Waste generating from collecting fuel

デブリ回収廃棄物（コンクリート）
debris (concrete)

1.E+01

1.E+01

Secondarywaste generated from water
水処理二次廃棄物
treatment

1.E+00

Secondarywaste generated from water
水処理二次廃棄物95%
treatment 95%
Secondarywaste generated from water
水処理二次廃棄物mode
treatment (mode)

1.E-01
1.E-02
1.E-03

1.E-04
1.E+00

1.E+02

1.E+04
1.E+06Time (second)

1.E+081.E+10

Fig.2: Relationship between concentration and time of thegas phase hydrogen (evaluation by the linear model)©International Research Institute for Nuclear Decommissioning38

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Evaluation result of hydrogen generation 2>
Table 1: Hydrogen generation rate evaluated by the linear model
Waste

Hydrogen generation
rate (L/sec)

RPV (metal waste)

2.94E-6

Waste generating from collecting fuel debris(concrete)

4.08E-6

MAX

1.09E-5

95%ile9.72E-7Mode

7.04E-8

Secondary waste generated from water
treatment (carbonate slurry)

Table 2: Concentration of equilibrium hydrogen evaluated by the radiolysis
Waste

Concentration ofequilibrium hydrogenin the gas phase (%)

RPV(metal waste)

14.4

Waste generating from collecting fuel debris(concrete)

15.1

MAX

12.3

95%ile

1.49

Mode

1.31

Secondary waste generated from water
treatment (carbonate slurry)

Fig.1: Concentration range of radioactive nuclides(example) [1]⚫ Several digits variations are found in theinventory evaluation values of eachnuclide.
⚫ There are large variations in hydrogengeneration rate and equilibrium
concentration depending on evaluation
values.

[1] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka, Yoshikazu Koma & Youko Takahatake (2021) Stochastic estimation of radionuclide composition in wastes generated
at Fukushima Daiichi nuclear power station using Bayesian inference, Journal of Nuclear Science and Technology, 58:4, 493-506,©International Research Institute for Nuclear Decommissioning

39

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Comparison of radiolysis with linear models>Linear model⚫ It is effective to evaluate the time ofreaching 4% in consideration of 0.4%
gas phase volume.
⚫ Hydrogen-oxygen recombination is notconsidered, therefore the concentration
values are excessively conservativecaused by the continuous increase of theconcentration.

100
Concentration of gas-phase hydrogen (%)

Radiolysis model⚫ It is effective to evaluate on theequilibrium concentration after passing
long time in consideration of hydrogenoxygen recombination reaction.
⚫ Assuming that gas-liquid equilibrium ofhydrogen concentration instantly
establishes, it was conservatively
evaluated.

10

1

0.1

0.01

ラジオリシスモデル
Radiolysis modelLinear model
リニアモデル

0.001
1.00E+00

1.00E+02

1.00E+04

1.00E+06

1.00E+08

1.00E+10

Time (second)Fig. Comparison of evaluation results of the RPV (metal
waste) hydrogen concentration©International Research Institute for Nuclear Decommissioning40

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Applicable plan of the radiolysis model in combination of the linear model>

teq

Time

Linear modelGas-phase hydrogen concentration (%)

Liquid-phase hydrogen concentration (%)➢ The radiolysis model is more
appropriate to evaluate equilibriumhydrogen concentration (①).
➢ The linear model is more appropriate
to evaluate the time of reaching 4%（②).

①

Radiolysis
model
4%

②

teq

Time

Fig.1: Applicable plan of the radiolysis in combination of the linear plan
©International Research Institute for Nuclear Decommissioning

41

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Issues for the evaluation model and parameter setting>
✓

✓
✓
✓

✓
✓

✓

There are two evaluation models to evaluate the gas-phase hydrogen concentration.: Radiolysis
and linear models.
The radiolysis model has characteristics of generating hydrogen and reaching equilibrium
concentration of the gas-phase hydrogen in consideration of water recombination.
In case of the linear model, hydrogen unilaterally generates, therefore the gas-phase hydrogenconcentration has no upper limit.
For above reasons, the radiolysis model is used to evaluate the gas-phase hydrogenconcentration inside the container whether hydrogen concentration exceeds 4% or not.
➢ In case of < 4%, ventilation may not be necessary.
➢ In case of >4%, ventilation design is necessary. In this case, the gas-phase hydrogenconcentration can be determined by a balance between the amount of hydrogen generation
and air exhaustion outside of the system.
Therefore, the linear model evaluation will be necessary to design ventilation.
Assuming that the radiolysis has established in accordance with Henry’s law, the gas phasehydrogen concentration is overestimated for the initial established region. Consequently, thelinear model is appropriate to evaluate the time of reaching 4%.
Currently, dynamic behaviors including accumulation and emission of hydrogen generated fromwaste of concrete and slurry are not reflected into both of the radiolysis and the linear models. A
method for reflecting these hydrogen generation evaluation should be considered as a futureissue.

©International Research Institute for Nuclear Decommissioning42

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Evaluation on the amount of hydrogen generation<Waste required for hydrogen generation measures>✓ The amount of heat generation (inventory) for three kinds of waste: i.e. metal waste
including the reactor pressure vessel, concrete waste such as the DSP slot plug
concrete, etc. generating from collecting fuel debris and secondary waste
generated from water treatment was conservatively evaluated. The result of theevaluation showed that hydrogen concentration inside the container could exceed
4% when sealing the container.
✓ These wastes could require measures for hydrogen generation including vent filter
installation depending on hydrogen generation rate.
✓ The amount of heat generation (inventory) is set as maximum values in thisestimation. To evaluate practical hydrogen generation rates, it is significant toappropriately set the amount of radioactivity (the amount of heat generation). For
practical and reasonable planning of storage and management, it is necessary toevaluate the amount of hydrogen generation based on the latest inventory.

©International Research Institute for Nuclear Decommissioning

43

(a) Investigation and evaluation of storage methods [1] Storage methods for storing high dose waste

Summary
Achievements so far now✓ Collecting and organizing information on waste characterization required for a storage
management method (waste compositions, inventory of radioactivity, type of radiation, watercontent rate and etc.) and hydrogen generation parameters.
✓ Waste required for storage management was selected according to the result of the study of waste
information and hydrogen generation factors, and evaluation models (radiolysis and linear models)and effects on the parameters were studied.
✓ Acquiring assumed values of equilibrium hydrogen concentration and hydrogen generation rate in
the container depending on type of waste based on evaluation of hydrogen generation amount.
✓ According to the evaluation results, installation of an appropriate vent filter is necessary becausehydrogen concentration in the container could exceed 4%.

Issues✓ The amount of waste heat generation (inventory) is significant to evaluate the value,
however the inventory evaluation values currently have variation in several digits. For
this reason, it is difficult to evaluate hydrogen generation rate and equilibrium
concentration that needs accuracy values.
✓ The improvement of evaluation accuracy for the waste inventory is needed for moreaccurate evaluation.
©International Research Institute for Nuclear Decommissioning44

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities◼ Achievements up to FY2020
 So far now, trouble cases in Japan and overseas have not been investigated and examined for the
purpose of reflecting into design of the containers, the storage facilities and the storage methods (system)
for the waste of the Fukushima Daiichi Nuclear Power Station (NPS).

◼ Goal Examining issues and measures during waste storage on the basis of the past trouble cases.

◼ Details of implementation (See details on next pages)
 Case study in Japan and overseas, and primary screening✓ Trouble cases during storing the radioactive waste container are surveyed and a referencecase is extracted to conduct a detailed survey for waste storage of the Fukushima Daiichi NPS.
 Detailed investigation✓ Detailed investigation for type of storage containers, information on stored waste, trouble
occurrence points and factors, environments of storage facilities, duration of storage areconducted and the result of the survey is examined.
 Study on issues and measures during waste storage✓ On the basis of the examined results, issues and measures for waste storage are studied.

◼ Indicators for the achievements of goal Completed examining issues and measures for waste storage based on the past trouble cases.
(Target at the completion: TRL：2→3）©International Research Institute for Nuclear Decommissioning45

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities＜Details of implementation＞
1.
Investigation in Japan and overseas, and the primary screening
The information on the following investigation items is acquired from a wide variety of public information in Japan and overseas. The
acquired information is examined to clarify trouble occurrence facilities and factors, etc. According to the investigation results, the primary
screening is carried out to extract cases that could reflect to design of the storage container, facility and method (system) of waste
generated from the Fukushima Daiichi Nuclear Power Station (NPS).
【Surveyed countries and public information】• Japan：Website of the JDream, Nuclear Information Archives (NUCIA) of Japan Nuclear Safety Institute (JANSI) and theInternational Nuclear Information System (INIS) database.
• U.S.: Trouble cases of Nuclear Regulatory Commission (NRC); Report of defects and incompatibility based on the 10 CFR Part21, and report related to the License Event Report (LER) provided by an operator.
• Europe✓ U.K.: Event notification for the Office of Nuclear Regulation (ONR) in every quarter and report for local communities atthe site and stake holders✓ France: Event notification for the Nuclear Safety Authority (ASN) as needed✓ Germany: Event notification for the Federal Office for the Safety of Nuclear Waste Management (BASE) on a regularbasis.
【Surveyed items】• Trouble cases of the storage facility for waste from Level 1 certificate (Class C waste) to Level 3 and metal containers.
• Fuel casks, fuel pools, flexible containers and concrete containers are not surveyed.
• In case that a trouble case caused by a vent filter installed for hydrogen management is detected, this trouble case would be
extracted as notable case.
2. Detailed survey
Investigation and information acquisition of extracted trouble cases related to occurrence factors including type of the container, storage
state of waste, trouble occurrence area and environment of the storage facility as well as additional information on recurrence prevention
measures.
3. Study on issues and measures at the time of waste storage
On a basis of the results of detailed investigations, potential issues that would cause a trouble when designing the storage container,storage facility and storage methods of the Fukushima Daiichi NPS are extracted to be considered and studied the measures.

©International Research Institute for Nuclear Decommissioning46

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey method in Japan>
According to the below database, key words regarding radioactive waste (waste material, etc.), the storage container (drum, etc.)
and the storage facility (storage house, etc.) are entered and retrieved 10 trouble cases (primary screening) of the container andthe storage facility of solid waste and during storage.
Database(organized by)
JDream (Japan
Science andTechnology
Agency: JST)

Overviews(survey period)
Approx. 70 million literatures in all fields of science technology
were acquired. Booklets (abstract journals and literatures) arecollected since 1981. The survey was conducted by Japanese key
word search.
(Duration of survey: From the beginning of registration August2021)

Nuclear Information
Archives(NUCIA)(Japan
Nuclear Safety
Institute: JANSI)The archive is intended to widely share the information on nuclear
power stations in Japan and operation of nuclear fuel cycle
facilities.
Report of trouble information (reports to the Japanese government
since 1966 in accordance with the regulation) and information on
maintenance quality (reporting to the government is notmandatory, but useful information has been shared since
September 2003) are registered.
(Duration of survey: From the beginning of registration to August
2021)

International
Nuclear Information
System (INIS)(Japan Atomic
Energy Agency:
JAEA)

INIS was established in 1970 aiming at promotion of technology
information sharing related to peaceful use of nuclear energy. The
INIS participating nations (24 international organizations from 132
nations) provided the nuclear literature information acquired fromtheir own nations for IAEA/INIS headquarter. The INIS
headquarter integrated these information and distributed theintegrated INIS database for each participant nation.
The total number of the nuclear literature provided fromparticipated nations was approximately 4.4 million literatures.
(Duration of survey: From the beginning of registration to August
2021)

Surveyed
number

Excluded survey case (except casesof solid waste)
•

0

•
•

9

• The leakage was discovered at thestorage area for solidification of
granular materials where is theintensive environmental facility in
the Fukushima Daiichi (Stored
materials are not solidification).

•

1

Radiation control for nuclear
medicine
The site improvement by
manufacturing jig
Report on increasing of storage
volume•Radioactive waste treatment at
the research institute in Tokai
area of Ibaraki Prefecture.
Facility management report at the
Aomori R&D Center of the Mutsu
Research Institute.

©International Research Institute for Nuclear Decommissioning47

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey results in Japan１： NUCIA >Nuclear power
plant

Date ofoccurrence

Summary of trouble cases

Primary screening

Tokai Ⅱ

March 19,2020

Damage caused by rust generated from a can of the Low-Level
Radioactive Waste (LLW) drum bottom which was carried out by
the Japan Nuclear Fuel Limited (JNFL).

Not found(Event caused from solidified materials)

Hamaoka

April 23, 2018

The chipped paint and water drops were found on the bottom of the
LLW drum(5 cans) carried out by JNFL. This was caused bycorrosion developed from cracks inside the drum bottom.

Yes(This event is caused by solidified materials
but an issue generated from storage andtransportation can be reference.)

Hamaoka

August 21,
2014

The expansion was found on the bottom of the LLW cement
solidification drum that is planned to be carried out.

Not found(Event caused from solidified materials)

ShimaneJune 13, 2011

Corrosion was found on the bottom of the drum at the solidified
waste storage facility, which was filled with the LLW waste.

Yes(Trouble case of stored waste)

KashiwazakiKariwa

November 2,
2007

The drums at the solidified waste storage facility were confirmed tohave fallen due to the Chuetsu-oki Earthquake (no contamination).

Not found(Event related to earthquake measures)

ShikaⅠ

October 30,
2007

The LLW waste storage drum were inspected after the Chuetsu-oki
Earthquake, and consequently the leakage of liquid was found fromthe drum.

Not found(Event related to earthquake measures)

KashiwazakiKariwa

July 17, 2007

Approximately100 cans of drums at the solidified waste storage
facility No.2 building were fallen due to the Chuetsu-oki Earthquake.

Not found(Event related to earthquake measures)

KashiwazakiKariwa

July 24, 2002

The leakage traces of tar-like solidified materials were found on the
drum of the solidified waste storage facility No. 2 building.

Not found(Event caused from solidified materials)

Fukushima
Daiichi

September 5,
1997

When lifting the drum for condensed waste liquid solidification prior
to a carry-out test, the bottom of the drum came out and the content
was spilled out from the drum.

Not found(Event caused from solidified materials)©International Research Institute for Nuclear Decommissioning48

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey results in Japan 2： INIS and others>
Facility

Date ofoccurrence

Summary of trouble cases

Primary screening

Oarai ResearchInstitute owned by theJapan Atomic Energy
Agency (JAEA)

January, 2019

【INIS】The leakage traces of asphalt were found out on the four cans of
the asphalt solidification filling drum.

Not found(Event caused from solidified
materials)

Waste disposal site,
Nuclear Science
Research Instituteowned by JAEA
(JAERI)

June 1, 2018
(Application forchanges)【Reference】 “Response to application for storage regulation
changes to confirm the integrity of the drum at the JAERI waste
disposal site: Response to long-term storage of radioactive waste,
JAEA”
Application for storage regulation changes to confirm the integrityof long-term storage drum at the underground pit and to conductrefiling work.

Not found•••(This is not a trouble case.
Radioactive isotope waste (RI
waste) should be screened.

The case of “To confirm the integrity of long-term storage drum at the underground pit of the JAERI” is not a trouble case but
the case was handled on a voluntary basis.
The waste surveyed was Radioactive Isotope (RI) waste but not included in from Level 1 waste (class C waste) to Level 3
equivalent waste.
The corrosion of the drum stored in the underground pit was caused by sea breeze and rain water at the temporary building,rain water infiltration when placing concrete slab, and directly adhering the waste to the drum since a polyethylene bag suchas waste cloth containing water is broken, etc. (Corrosion cases caused by water contents, etc. of the drum were similar to trouble cases of
Hamaoka and Shimane Nuclear Power Plants, therefore, this case was excluded in the detailed survey and there would not be a problem in extracting
issues.)【Other JAEA trouble cases 】••

There is no report of the waste storage from the press release about the accident and trouble cases provided by the JAEA
website JAEA (since 2005) .
There is no reference case regarding the waste storage from report of the JAEA Mutsu Office (Safety Monitoring Committee,
Aomori Prefecture)©International Research Institute for Nuclear Decommissioning49

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Primary screening in Japan>⚫ Most of trouble cases in Japan were caused by solidified materials after waste were
solidified and not caused by waste under storage.
⚫ A trouble case of the Shimane Nuclear Power Plant (NPP) was caused by corrosion of thedrum that were filled with the Low Level Waste (LLW). This case was selected to conduct a
detailed survey.

⚫ A trouble case of the Hamaoka NPP was caused by paint removal (corrosion) of the waste
container after waste were solidified, however storage of the waste was also one of thecauses and a similar trouble might occur during storage.
⚫ The following cases were selected to conduct the detailed survey.
1) Shimane NPP (June 13, 2011): Corrosion was found at the drum bottom that wasfilled with the LLW waste at the solid waste storage facility.
2) Hamaoka NPP (April 23, 2018) : Paint removal and water drop adhesion were foundat the bottom of the LLW drum carried out by Nuclear Fuel Limited.

⚫ No trouble case regarding the storage facility was found. Another survey of the Japan
standard for the storage facility was added in the detailed survey.
©International Research Institute for Nuclear Decommissioning50

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey methods in US>
On the basis of the below database, keyword searches were conducted by using keywords related to radioactive materials,
metal storage container (drum can and container) and storage facility (storage house, etc.) . Trouble cases of the containerand storage facility for solid waste, and during storing waste were surveyed by the keyword searches.
Name of database

Overview

Report of defects andincompatibility based on the
10 CFR Part 21

The 10 CFR Part 21, the U.S. regulation, requests to report the U.S. Nuclear
Regulatory Commission (NRC) when defects with safety concerns or incompatibility
related to the approval and authorization are found in nuclear and radioactive
facilities as well as the relevant significant facilities and equipment. The NRC reports
since 1995 are available on the NRC website.
（Duration of survey: From 1995 to August, 2021)

License Event Report (LER)

The 10 CFR Part 50.73. the U.S. regulation, requests providers of the nuclear facility
to create the License Event Report (LER) and submit the report within 60 days whenabnormal events and failures are found in the nuclear facilities. The LER should
include overviews of the events, clarifications of the causes, analysis of thesignificance and the correction measures, etc. The LERs since 1980 are available on
the NRC website.
（Duration of survey: From 1995 to August 2021)

Event Notification Report（ENR）

The NRC requests to submit the Event Notification Report (ENR) the report inaccordance with the 10 CFR 50.72 (emergency report case), the U.S. regulation,
within 1 hour up to 24 hours depending on the events of the emergency reportcases. The ENR is relatively simplified than the LER. The ENR since 1999 areavailable on the NRC website.
（Duration of survey: From 1999 to August 2021)©International Research Institute for Nuclear Decommissioning

51

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening< Survey results in U.S>
Database

Search key
words

No. ofsearchResults of primary screening

Reportbased
on the10CFRPart 21container

4

These cases are related totransport and the container ofbattery, etc. There is no survey
case.

package

23

These cases are related totransport and work packages,
etc. There is no survey case.

gtcc

0

Greaterthan-Class
C

0

Cask
(addition)

9

These cases are related tostorage casks for spent fuel.
There is no survey case.

drum(addition)

3

These cases are transport anddrum brakes. There is no survey
case.

Storehouse(addition)0―

Database

No. ofsearchResults of primary screening

container
AND waste

35

These cases are related totransport and the container ofbattery, etc. There is no survey
case.

package
AND waste

51

These cases are related totransport and work packages,
etc. There is no survey case.

―

gtcc

0

―

―

Greater-thanClass C

0

―

waste
container

64

These cases are related totransport, the medical facilities
and safety of criticality offission products. They are
exempt from survey case.

waste
package

3

These cases are related to the
medical facilities. There is no
survey case.

gtcc

8

Seven cases of the searched
result are related to spent fuel.
So, they are exempt fromsurvey case.

Greater-thanClass C

1This is a duplicated data of thesearch result of gtcc.

LER

ENR
(addition)

Search key
words⇒ Applicable case is only one case of the Event Notification Report (ENR).
©International Research Institute for Nuclear Decommissioning52

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Primary screening in U.S.>⚫ One following case was retrieved by the primary screening. However, this case was not reported as anabnormal event.
➢ Name of power plant: Zion
➢ Date of occurrence: February 3, 2015
➢ Subject: Functional loss of the safety system due to snow
➢ Event number：50787➢ Event (content of the report)✓ Ventilation parts of air flow inlets that are installed with the vertical type storage casks for spent
fuel, MAGNASTOR, manufactured by NAC International Inc. were clogged with snow and theclogging rate was more than 50%: 34 casks in total including 33 storage casks for spent fuel and1 storage cask for the GTCC waste.
✓ Exelon Corporation that is approved to use Zion, submitted the Event Notification Report (ENR)
to the U.S. Nuclear Regulatory Commission (NRC) within 24 hours after occurrence of the eventin accordance with the 10 CFR 72.75d(1)(i).
✓ On February 4, snow was removed from the ventilation parts and the clogging rate of these 34
casks were recovered to be less than 50%.
✓ After that, Exelon Corporation and NAC International Inc. reviewed the Safety Analysis Report(SAR). As a result of the review, the recovering measure to reduce clogging rate less than 50%
was completed within 10 hours, while the SAR requests to complete a recovering measure within
58 hours. Therefore, this case was not regarded as an emergency report case and the ENR-24
hour-report was withdrawn in accordance with the 10 CFR 72.75(1)(i).
⚫ The above event was not a trouble case that can refer to waste of the Fukushima Daiichi to be stored in thebuilding, so that a detail survey will not be conducted.
©International Research Institute for Nuclear Decommissioning

53

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey method in Europe>
On the basis of the below database, keyword searches were conducted by using keywords related to radioactive materials, metal storage
container (drum can, etc.) and storage facility (storage house, etc.) . Trouble cases of the container and storage facility for solid waste, andduring storing waste were surveyed by the keyword searches.
Nations
UK

FranceGermany

Database (range ofsurvey)Overview

Quarterly report of the
Office for Nuclear
Regulation (ONR)

The report of the nuclear facilities that is selected from among the ONR event reports provided by operators (responsible for thesite management) should be submitted to government ministries other than the ONR. These reports are posted on the ONR
website every quarter (The report that is not applied to the requirements of the ministry is not disclosed in public even though
the INES level exceeds level 1).
（Duration of survey: From last quarter in 2000 to first quarter in 2021)

Report for the ONRmunicipalities and stakeholder groups

The information on the nuclear facilities that is subjected to the ONR regulations such as status of the operation management,
regulatory inspection and experiences of the event, etc. is examined and the explanation is given to mainly administrative
authorities of municipalities with nuclear plants and stakeholders every quarter (or every half year). Occasionally, a report that is
not included in the quarterly event notification is provided.
（Duration of survey: From 2014 to first quarter of 2021)

Occasional event
notification of the
Nuclear Safety Authority(ASN)

The event reports of the nuclear facilities that are classified as the INES level 1 or more from among the ASN event reports
provided by operators (responsible for the site management) are extracted and disclosed on the ASN website as needed (Theevent reports are generally not available in English).
（Duration of survey: From 2009 to as of August in 2021)

ASN annual report

The information on trends of the ASN regulation activities, status of the nuclear facilities required regulations and others is
systematically organized and annually reported (The English report is also provided.). The information on the total number of thereport and details of the classification is mainly described. If there is a significant report that should be paid attention, theinformation on individual event is also described.
（Duration of survey: From fiscal year 2009 to 2020)

Monthly report for the
nuclear reactor event of
the Federal Office for
the Safety of Nuclear
Waste Management(BASE)

In case that an event of the nuclear facility occurs, an operator (responsible to the site management) reports the authority of
approval in the state and then the state authority of approval reports the Federal Office for the Safety of Nuclear Waste
Management (BASE). The monthly report for the nuclear reactor events includes events of nuclear power plants and research
reactors that are more than 50kW of thermal output and are summarized the BASE reports (including the INES level 0) provided
from each state authority every month (no English version).
（Duration of survey: From 2010 to June in 2021)

BASE annual report forevents of the nonnuclear reactor (fuelcycle facility and others)

The BASE annual report of non-nuclear reactor is summarized the BASE reports (including the INES level 0) of events of fuelcycle facilities and others provided by each state authority every year (no English version). Compared with the monthly report of
the nuclear reactor events, the annual report of non-nuclear reactor includes only limited information on each event.
（Duration of survey: From 2010 to 2020)©International Research Institute for Nuclear Decommissioning

54

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey results in U.K.>
NuclearPower Plant

Day ofoccurrence

Summary of trouble cases

Primary screening

Dounreayintermediatelevel
radioactive
waste disposal
facility

November 12,
2002(INES level 0)
A waste container in which spent batteries
were stored was put into a flask for waste
management; however, a confinement functionof the flask was not effective because zincsolution leaked from spent batteries. Eventually,
this case caused contamination and accidental
exposure.

None(Currently a battery is
not a subject to waste
management. Details of
the battery were notclarified and recognized
as small waste. )West Cumbria
high dose
waste
solidification
disposal facility

February 7,
2019

A lid of the container should have been welded
to close after high-level waste (HLW) was
solidified in the container; however cleaning
solution was sprayed for decontamination
without closing the lid.

None
(A trouble case related
to treatment after waste
solidification)

West Cumbria
THORP
reprocessing
facility

Discovered an
event during
April toSeptember
2019(INES level 1)
A welding part of a container that is designed
to enclose special nuclear fuel material had aquality defect (insufficient performance ofresistance pressure) caused frommanufacturing.

None(A trouble case related
to manufacturing of thecontainer)<Primary screening in U.K.>
Target event for detailed survey: None
©International Research Institute for Nuclear Decommissioning

55

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Survey in Japan and overseas, and primary screening<Survey results in France>
Nuclear powerplant

Day ofoccurrence

The La Hague
site

March 27,
2013

(INES level 1)
When the container (vinyl flexible container) thatwas packed with radioactive waste was loaded
into a transportation vehicle, the vehicle wasoverturned and the content of the vehicle was
exposed.

Tricastin(Pierrelatte)conversion plant(INB 105)

Discovered
on February
7, 2019(INES level 1)
As a result of investigating a storage area of metal
containers (drums) contained with materials andequipment that were no longer needed (including
radioactive materials), the result showed that seal
performance of the container was incompatible.

Tricastin(Pierrelatte)conversion plant(INB 105)

Discoveredin July 23
and August 6,
2018

Summary of trouble cases(INES level 1)
Two cans of the metal containers (drums)
contained with materials and equipment that wereno longer needed (including radioactive materials)
had lost sealing performance. Contamination had
spread outside of the container.

Primary screening

None(A trouble case of theflexible container)Yes( A trouble case of thewaste container was
surveyed as thesimilar site related
event)<Primary screening in France>
Target event for detailed survey: A trouble event of the Tricastin (Pierrelatte) conversion plant（INB 105）
©International Research Institute for Nuclear Decommissioning

56

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities< Survey results in Germany>

Survey in Japan and overseas, and primary screening

Nuclear powerplant

Day ofoccurrence

Summary of trouble cases

Primary screening

Isar Nuclear power
plant Unit 1

September 30,
2011(INES level 0) When waste was contained in 200 liter-drumand treated by drying before sealing it, the waste content
emitted smoke.

None(A trouble case of pre-storage
treatment)

Brunsbüttel Nuclear
power plant

August 13, 2014

(INES level 0) Corrosion of multiple drums was found, inwhich spent contamination removal filters were enclosed.
After that, survey was conducted and showed that there is a
deficiency in waste management such as handling, storage
and transfer of waste.

Yes(Corrosion event of thestorage container)

Gundremmingen
Nuclear power
plant

June 13, 2016

(INES level 0) When low-level waste (LLW) was enclosed in
180 liter-drum and treated by drying before sealing it, thewaste content emitted smoke.

None(A trouble case of pre-storage
treatment)

Brunsbüttel Nuclear
power plant

May 25, 2018

(INES level 0) The 200 liter-drum enclosed spentcontamination removal films became deteriorated and was
unable to continue to use. When it was opened and the waste
content was collected by using a vacuum machine, it caused
radioactivity leakage.

None(A trouble case of wastetransfer)

Brunsbüttel Nuclear
power plant

October 6, 2020

(INES level 0) The containers enclosed dismantled structures(waste metal) generated from the decommissioning weretransported to the melt-reprocessing facility. The waste
contents were not corresponding to data of the container
identification form that was registered in the dismantled
material tracking system.

None(A trouble case of datasystem management system)

Brunsbüttel Nuclear
power plantFebruary 9,
2021(INES level 0) Specifications of concrete filling material thatwas used when waste was enclosed into the metal container
for final treatment was not complied with designatedspecifications.

None(A trouble case related to
specifications of solidificationmaterials)<Primary screening in Germany> Target event for detailed survey: A trouble case of the Brunsbüttel Nuclear power plant in 2014.
©International Research Institute for Nuclear Decommissioning

57

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Detailed survey<Detailed survey in Japan:1>
Subject

Shimane Nuclear Power Station Verification of the corroded drums

Date of verificationof occurrence

June 13, 2011

Location of
occurrence

Solid waste storage facility Building B[Circumstances at the time of occurrence]
When drums stuffed with low-level radioactive waste were being moved, it was found that the base of 5 drums was corroded (through holes
were present). Thereafter, when external inspection of approx. 23,000 drums that were in storage was carried out, it was found that only 5 drums
had through holes in their bases due to corrosion.
[Overview of the drum]
Capacity: 200 liters[Cause investigation]
The corroded drums and their contents were checked for the following.
● Deterioration of the paint on the inner surface of the drums(Peeling off of the plaint due to the waste)● Checking for water and moisture from the contents of the drums

Corroded
portionApprox. 5mm▲Enlarged view of the corroded portion(through hole)▲Corroded drums Bases[Cause]As a result of checking the corroded drums and their contents it was presumed that water and moisture adhered to the portion of the innersurface of the drum where the paint had deteriorated because of which the portion got corroded and through holes developed.
[Recurrence prevention measures]
Revisions were made to incorporate the following into the internal procedures concerning management of solid radioactive waste.
● In order to prevent deterioration of paint and to prevent moisture adhesion, plastic containers with painted interiors (cylindrical inner container)
shall be used.
● Guidelines for re-use of drums and the methods for packaging waste shall be clearly specified.
● If drums in which corrosion has progressed and through holes are likely to develop are identified, the waste shall be transferred to drums thatare sound.

Issues of waste storage: Prevention of container damage and deterioration, and water adhesion©International Research Institute for Nuclear Decommissioning

58

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities<Detailed survey in Japan:2＞

Detailed survey

Subject Hamaoka Nuclear Power PlantPeeling off of the paint and adhesion of water droplets at the base of the drums containing low-level radioactive waste
Date of verification of
occurrence

April 23, 2018

Location of occurrence

JNFLLow-level radioactive waste burial center[Circumstances at the time of occurrence]
When preparations were being made for burying the 960 drums (waste bodies) that were handed over on March 25, it was found that the paint at the base of 1 of thewaste bodies had peeled off and water droplets had adhered to it. Thereafter, when external inspection of the remaining 959 waste bodies and of the 928 waste bodies
planned to be transferred in May the same year was carried out, it was found that the paint had peeled off and water droplets had adhered to the base of 1 waste body ineach of the lots (total 2 waste bodies).
Further, when external inspection of the waste bodies which had been temporarily stored in the burial center for over a year was carried out based on the internal
regulations of JNFL from February 25, 2019 to March 12, 2019, it was found that the paint at the base of 1 waste body had bulged and water droplets had adhered to it,
and the paint at the base of 1 waste body had bulged, peeled off and water droplets had adhered to it (hereinafter referred to as “peeling off, etc. of paint”
Peeling off, etc. of paint was found in a total of 5 waste bodies. (The radioactivity in the water droplets adhered to the bases of all the waste bodies was lower than thelower limit of detection.)[Cause investigation]
In the case of 1 of the waste bodies it was found that fine iron powder had adhered to the painted surface during the process of drying the outer surface when the drum
was manufactured, which got rusted over time and expanded due to which the paint on the outer surface bulged.
The following was found in the case of the remaining 4 waste bodies.
① Cracks on the inner surface of the base ② Gap between the mortar and the inner surface of the base of the drum ③ Chlorine (Cl) on the inner surface of the base
④ Corrosion from the inner surface of the base reaching up to the outer surface ⑤ Iron corrosion products and zinc corrosion products formed on the outer surface ofthe base under the presence of chloride ions (Cl-) ⑥ Calcium (Ca) which is a component of mortar was found on the outer surface of the base ⑦ Water (water vapor
close to the amount of saturated water vapor) is likely to be present on the inside (Document verification) ⑧ Corrosion from the inner surface of the base was found tohave reached the outer surface of the base of the waste body(One of the through holes was approx. 6mm)

[Causes]In the case of 4 waste bodies, cracks developed on the inner surface of the base of the drums when thewaste was being packaged or when the drum was being transported after packaging. Water accumulated in
the gap formed between the filled mortar and the inner surface of the base. The cracks served as thestarting point and corrosion occurred.
It progressed and led to peeling off of the paint on the outer surface.

③Bleeding water②Gap

Inner surface
MortarPaint
Zinc[Recurrence preventions measures]● (Crack reduction) The waste shall be placed by tilting the drum. In the case of plank-shaped waste the flat
surface shall be placed at the base.
● (Gap prevention) A vibration machine shall be used so that the mortar can be filled easily.
● (Personnel training) Personnel undertaking packaging work shall be well informed about this phenomenon
and shall undergo training on recurrence preventions measures.

①CrackBase ofthe drum④Progression of
corrosion
Base
Zinc
Paint

Base of the drum④’ Corrosion, peeling off, etc.

Issues of waste storage: Prevention of container damage and deterioration, and water adhesion©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Detailed survey<Relevant regulations in Japan: 1>
Requirements related to container storage of solid waste have been identified from the waste storage related articles in the Regulations forCommercial Nuclear Power Reactors⚫
Regulations related to installation, operation, etc. of commercial nuclear power reactors
Article 67 (Records)
In accordance to the provisions of Article 43-3-21, records pertaining to the items mentioned in upper section of the following table must be recorded foreach commercial nuclear power reactor in accordance with the topics mentioned in the middle section of the table respectively, and the records must be
retained for the periods mentioned in the lower section of the table respectively. (Here the table has been left out, and only the recorded items areindicated.)5
Radiation control recordsh
The quantity of nuclear fuel material transported outside the factory or plant, or items contaminated due to the nuclear fuel material, by type,
the type of container used for transporting these, and the date/time and route of transportation.
i
The type of radioactive waste disposed at the waste disposal facility or dumped in the ocean, quantity of radioactive substances contained insaid radioactive waste, quantity of containers in case said radioactive waste is enclosed in containers or in case it is integrally solidified withthe container, and its specific gravity, as also the date, location and method of its disposal or dumping
Article 90 (Disposal carried out at the factory or plant)
According to the provisions of Article 43-3-22 (1), with regards to the disposal of radioactive waste carried out at the factory or plant where the commercial
nuclear power reactor licensee has installed the commercial nuclear power reactor, the measures listed in each of the following items must be taken, andthe status of implementation of these measures must be verified before disposal.
11
If disposal is carried out by the method mentioned in Item 6 c, when radioactive waste is stored and disposed at the storage anddisposal facility with radiation hazard prevention effect, it must be carried out as follows.
a
If radioactive waste is enclosed in containers for storage and disposal, necessary measures must be taken to prevent the spread ofcontamination in case cracks develop in the container or the container gets damaged, such as wrapping said container withmaterial that can absorb the entire enclosed radioactive waste, or installing a pan that can hold the entire enclosed radioactive
waste, etc.
b
If considerable overheating is likely to occur due to decay heat, etc. of said stored and disposed radioactive waste, necessary coolingmeasures must be taken.
⚫
Regulations on technical standards for nuclear reactors for commercial power generation and their auxiliary facilities
Article 40 (Waste storage facility, etc.)
The radioactive waste storage facility must be installed in accordance with the following.
i) It must have the capacity to store radioactive waste generated during normal operation.
ii) The structure of the facility must be such that radioactive waste does not easily leak out from it.
iii) The facility must withstand decay heat or heat generated due to exposure to radiation, and must not get corroded considerably due tothe impact of chemicals contained in the radioactive waste or due to other load.
2 The commercial nuclear power reactor where the facility for storage of radioactive waste in solid form is installed, must be installed such that
contamination from radioactive waste does not spread.
©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Detailed survey<Relevant regulations in Japan: 2>
An example of “Technical Specifications for Radioactive Waste Management” specified at the nuclear power plant inaccordance with the rules (Technical Specifications of the Shimane Nuclear Power Plant) is given below.
Chapter 6 “Radioactive Waste Management”
Article 86 (Radioactive solid waste management)
3. Each Manager shall verify the following items, and if any abnormalities are found, the Manager shall take necessary
measures.
(1) In order to verify the storage conditions of radioactive waste at the storage facility, the Manager (Radiation Control)
shall patrol the storage facility once a week, and in addition, shall take stock of the stored items once in 3 months.
(2) The Manager (Operations) shall monitor the storage condition of the spent resin stored in the waste resin tank, etc. and thefilter sludge, and shall take stock of the stored items once in 3 months.
(3) In order to verify the storage conditions of the spent control rods, channel boxes, etc. irradiated in the reactor, which are
stored in the on-site bunker, the Manager (Fuel & Technology) shall patrol the on-site bunker once a month and in addition,
shall take stock of the stored items once in 3 months. Moreover, the Manager shall take stock once in 3 months of the spentcontrol rods, channel boxes, etc. irradiated in the reactor, which are stored inside the fuel pool.
6. The Manager (Radiation Control) shall ensure that at the time of transportation mentioned in the previous section, the dose
equivalent rate of the container, etc. does not exceed the value specified by law and the concentration of surface
contamination of the container, etc. does not exceed 1/10 th of the surface contamination limit specified by law. However, if
containers are transported from the zones specified in Article 92 (zones in radiation controlled areas) -1-1, the verification of
surface contamination concentration can be omitted.

Issues of waste storage: Inspection (considering a layout of the containers to observe the appearance ofthe storage container)©International Research Institute for Nuclear Decommissioning

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(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Detailed survey<Europe: Detailed survey 1>
France: Abnormality in the container for radioactive materials and non-conformance at the uranium conversion plant (INES Level 1)

Subject
Date of verificationof occurrence

2017

Location of occurrence

Tricastin (Pierrelatte) uranium conversion plant Shut down in 2008
(Facility registration number INB105)[Circumstances at the time of occurrence]
The operator, AREVA NC (formerly COMURHEX), started a series of inspection activities pertaining to metal containers (containers in which equipment andmaterial, etc. that are no longer required and which contain radioactive materials are placed) that are stored for several years at the storage buildings
exclusively for storing containers with radioactive materials. Essentially, the container was supposed to function as the primary containment barrier and thesurrounding structure (shaped like a vault, pit) in which the containers are placed was supposed to serve as the secondary containment barrier. However, as aresult of the inspections, non-conformances related to the containment barrier were found in several containers. Further, in spite of concerns about the sealing
performance of the containers being impaired, the service entrance of the surrounding structure was not closed. This was classified as Level 1 on theInternational Nuclear Event Scale (INES).
[Cause investigation]
The container with said storage zone issue had not undergone safety analysis.
Also, the container had not undergone periodic inspection contrary to the general operation rules of the facility and the perspective of the department
responsible for radiation protection on the protection policy.
Further, a part of the service entrance of the surrounding structure was not properly closed in reality but was mistakenly believed to have been properly
closed.
[Cause]● Inadequate container management● Failure to conduct periodic inspection[Recurrence prevention measures]● Improvement in container management● Implementation of safety analysis

Issues of waste storage: Pre-study of incompatibility management during storage©International Research Institute for Nuclear Decommissioning

62

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Detailed survey<Europe: Detailed survey 2>
Subject

France: Loss of sealing performance of the container with radioactive materials at the uranium conversion plant and resulting diffusion of contamination(INES Level 1)

Date of verification of
occurrence

July 23, 2018
August 6, 2018

Location of occurrence

Tricastin (Pierrelatte) uranium conversion plant Shut down in 2008
(Facility registration number INB105)[Circumstances at the time of occurrence]
The operator, Orano (formerly AREVA NC) discovered that the sealing performance of a total of 2 containers with radioactive materials (containers in which equipment
and materials, etc. that are no longer required and which contain radioactive materials are placed) was impaired. The measuring instrument of the radiological monitor
in the vicinity of the storage facility was showing an abnormal reading.
It was decided to collect and treat the contents from the containers in question and transfer them to different containers, and until then the containers were moved
inside a robust air-lock facility (surrounding structures with static and dynamic containment functions respectively) in which negative pressure was maintained.
However, of the 2 surrounding structures, 1 had not been properly closed, and contamination diffused to the outside from there.
When this event was discovered, maintenance work was not being performed and hence the event did not affect any personnel. Also, based on the level ofradioactivity detected outside the surrounding structure, it was concluded that there was no impact outside the site. However, since the surrounding structure notbeing closed properly and contamination being diffused to the outside is a violation of the general operation rules, and since the discovery was delayed, the event wasclassified as LEVEL 1 of the International Nuclear Event Scale (INES).
[Cause investigation]
As the temperatures are high in said storage zone (inside the surrounding structure) in the summer season, and since hydrofluoric acid was present in the containers
in gaseous form, the internal pressure of the container increased and sealing performance deteriorated leading to leakage from the head of the container. The
containers had not undergone safety analysis.
The alarm equipment that was supposed to be linked to the measuring instrument for the radiological monitor was not configured properly and as a result did notfunction, due to which the anomaly was overlooked.
[Cause]● Inadequate container management● Increase in internal pressure of the container due to high temperatures in storage zone (inside the surrounding structure) in the summer season, and presence ofhydrofluoric acid in gaseous forms in the containers● Inappropriate closing of the surrounding structures, and inappropriate configuration of the alarm equipment[Recurrence prevention measures]● Improvement in container management

Issues of waste storage: Pre-study of incompatibility management during storage, and study and measures of the cause
of inner pressure increase©International Research Institute for Nuclear Decommissioning

63

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage facilities

Detailed survey<Europe: Detailed survey 3>
Subject Germany: Inadequacies concerning the management of containers (drums) for solid radioactive waste at Brunsbüttel (INES LEVEL 0)
Date of verification ofAugust 13, 2014
occurrence

Location of occurrence

Brunsbüttel nuclear power plant (Decommissioning underway)[Circumstances at the time of occurrence]
Inadequacies were found in the management of containers (drums) for solid radioactive waste. The waste containers in question were generated during thedisposal of the decontamination filters in FY2012. One of the containers was significantly corroded (its entire contents were retrieved), and corrosion wasobserved in some of the other containers as well.
[Cause investigation]
The operator carried out a series of evaluations to check whether there are any issues in the handling, treatment and storage of solid radioactive waste. The
evaluations were carried out focusing on 6 solid waste storage facilities located in underground zones in the controlled areas. Containers in which low andmedium level radioactive waste (mainly, mixed waste containing filter resin, residue from the vaporization process, hazardous chemical substances)
generated during operation is placed, are stored in these zones.
As a result of evaluation, it was found that corrosion of the container was caused due to increase in moisture, inadequate corrosion prevention measures,the protective coating getting damaged during handling (coating material on the outside and inside), etc.
[Cause]● Inadequate container management● Increase in moisture in the storage facility● Inadequate corrosion prevention measures● Damage to the protective coating during handling[Recurrence prevention measures]● Improvement in container management● Enhancement of corrosion prevention● Prevention of damage to the protective coating and early detection

Issues of waste storage: Prevention of container damage and deterioration, water adhesion and early
detection of the container damage©International Research Institute for Nuclear Decommissioning64

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Study on issues and measures at the time of waste storage<Issues> On the basis of the primary screening and results of the detailed survey, requirements of thestorage containers and storage facilities, trouble cases that were needed the requirements and issues to bereflected into further study are indicated as below.
No.

1

2

3

Requirements

Trouble cases that were needed
the requirements

IssuesStorage managementof radioactive solid
waste

・Safety regulations
・France: Tricastin uranium
conversion facility
・Germany: Brunsbüttel Nuclear
power plant

・Early detection: Regular inspection 1
time/week (considering layout of thecontainer to observe the appearance ofthe storage container)
・Pre-study on incompatibility
management at the time of storage

Fall prevention at thetime of earthquake

・Japan: Chuetsu-Oki Earthquake(Kashiwazaki-Kariwa and Shika
Nuclear Power Plants (NPP))・Measures to prevent falling whenearthquake occurs.

Ensuring the integrityof the container suchas corrosion
management, etc.

・Japan: Shimane NPP, corrosion of
drum
・Japan: Hamaoka NPP, corrosion ofdrum・France: Tricastin uranium
conversion facility
・Germany: Brunsbüttel Nuclear
power plant・Prevention of container damage anddeterioration
・Prevention of water adhesion・Verification and treatment of thecontent (whether substances including
corrosive substance that should be
removed exist or not.
・Pre-assessment and measures forgeneration factors of inner pressure©International Research Institute for Nuclear Decommissioning65

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Study on issues and measures at the time of waste storage<Responding to issues>
Responding to issues are described as below.
No.

1

2

3

Requirements

Storage managementof radioactive solid
waste

Fall prevention at thetime of earthquake

Ensuring the integrityof the container suchas corrosion
management, etc.

Issues

Proposed measures

Regular inspection

Installation of effective inspection passage.
Specifically, work efficiency shall be enhanced considering risks ofthe container integrity; In case of high risks, the inspection passage
shall be arranged at a position that can be observed the appearance
and sampling inspection shall be conducted for lower risks.

Pre-study on incompatibility
management at the time ofstorage

Incompatible factors during storage shall be clarified and responding
to issues shall be considered in advance. Installing of alarm andconfinement system shall be considered if necessary.

Fall prevention measures

Fall prevention measures shall be implemented to prevent fromfalling with acceleration of assumed earthquake and analysis
assessment of the measures shall be carried out to make sure thatthe measures have no problem. When the containers are pilled up, aconnecting measure shall be implemented.

Prevention of container
damage and deterioration

A storage method to prevent from causing damage of the container
surface (i.e. Method for tilting the container, protruding objects,etc.)shall be carried out. Therefore, workers shall be trained.

Prevention of water adhesion

・A plastic container shall be used to store waste containing water.
・At the storage area, temperature and humidity shall be controlled.

Verification of the content(whether substances including
corrosive substance that should
be removed exist or not.

・Corrosion factors shall be verified: Chloride ion that is the cause of
metal container corrosion, existence of sulfate ions, dissimilar metals
that may cause galvanic corrosion resulting from contact with themetal. Then, neutralization and plastic containers shall be used.

Pre-assessment and measures
for generation factors of innerpressure

The inner pressure factors shall be considered. On the basis of theinner pressure assessment, ,measures (ventilation) shall be installed.

©International Research Institute for Nuclear Decommissioning66

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Survey of trouble cases of containers and storage
facilities

Study on issues and measures at the time of waste storage<Concrete measures 1>
Practical reference cases and requirements
are needed
to crystallize proposed plan of waste storage for
Examplesthatof concretemeasures
No.
Requirements
Issues
Proposed plan of concrete measures (draft)(reference
cases)
Fukushima Daiichi are indicated as below.

1

Storagemanagement ofradioactive solid
waste

Regularinspection

Pre-study on
incompatibility
management
at the time ofstorage2

Fall prevention
at the time ofearthquake

Fall
prevention
measures【Management of solid waste storage
houses in Fukushima Daiichi】
Stored drums were verified
abnormalities as much as possible bysame regular inspection as before theGreat East Japan Earthquake. The“Inspection flow of the drum” [1] covers
a specific method to verify the drums
that cannot be observed because of
their position.

Regular inspection for existing stored waste is being
studied. [2]When storage containers and their storage
condition are needed to revise in future, specificinspection and verification methods will becrystallized and methods for installing the storage
container will be considered.

【Performance evaluation of vent filter】(under this project)
Filtering events and measures that wereneeded to consider were studied.

On the basis of specifications for the storage facility,
the storage container and waste, incompatibility
factors will be clarified and necessity of specificmeasures will be considered.

【Fall analysis of low-level storage drumat the time of earthquake】
The Central Research Institute ofElectric Power Industry (CRIEPI)
verified the impact of the measures by
evaluating bearing force against
falling.[3]Existence of waste that may spread radioactive
materials due to the fall will be investigated, andradiation exposure assessment due to the spread ofradioactivity and fall prevention measures will be
conducted.

[1]：TEPCO Holdings, Inc. Fukushima Daiichi NPS, “Storage status of radioactive solid waste (drums, etc.) “ July 23, 2018, File No.4, Regulatory Councilof Radioactive Waste from Nuclear Facilities (7th),[2]：” Implementation Plan of Nuclear Facility, Fukushima Daiichi NPS”, Maintenance of Nuclear Ⅲ, Supplementary of Radioactive Waste Management 2”.
[3]：CRIEPI, “Analysis of Falling Low Level Storage Drum at the time of Earthquake”, 2011, the result of research for FY2020 issued by CRIEPI ResearchInstitute for Earth and Engineering, Environment and Science.
©International Research Institute for Nuclear Decommissioning67

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Study on issues and measures at the time of waste storage<Concrete measures 2>
Practical reference cases and requirements that are needed to crystallize proposed plan of waste storage for Fukushima Daiichi are
indicated as below.
No.

Requirements

Issues3

Proposed plan of concrete measures (draft)[Guideline of waste storage management
designated by utility companies]
Points to consider during storage aredescribed in the waste storage guideline.

At the stage of clarifying waste that will be stored in thecontainer, crystallization of storage methods will beconsidered (Ex. Protruding objects should not be stored
at the container bottom. Board-shaped waste should be
stored at the container bottom, etc.).

[Power plant and use of the plastic
container]
Use of polyethylene inner containers for
drums[Nine storage houses in Fukushima
Daiichi]Control of temperature and humidity

Water volume to be adhered will be understood andnecessity of water treatment will be considered.
In case of using inter containers, design and theapplication method of the container will be considered(When storing high dose waste in the container,
deterioration effects caused by radioactivity will be
evaluated. In addition, if there is a possibility to generatehydrogen gas, specifications of non-airtight container will
be considered).
Necessity of temperature and humidity control will be
studied when storing waste.

Verification of
the content(existence ofcorrosion
substances)[Guideline of waste storage management
designated by utility companies]
Clarified substances to be removed
including corrosive substances areprovided in the guideline of waste storage.

Investigation of corrosive substances in waste stored,clarification of substances to be removed and refection
into the storage management guideline.

Preassessmentand measures
for generation
factors of innerpressure[Study of storage methods for high-level
radioactive materials] (under this project)
Evaluation of hydrogen generation amountbased on investigation of waste properties
and study of ventilation deterioration
factors

Detailed investigation of waste properties, study on
factors of inner pressure increase and evaluation of theinner pressure (study on specifications of the storage
container and storage methods, and conducting of tests,
etc. )

Prevention ofcontainer
damage anddeterioration

Ensuring the
integrity of thecontainer suchas corrosion
management,
etc.

Examples of concrete measures(reference cases)

Prevention ofwater adhesion©International Research Institute for Nuclear Decommissioning

68

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage
facilities<Results of the survey>⚫ Trouble cases of waste storage facilities and metal containers classified from the L1 waste level (class C
waste) to the L3 level were searched for survey among the public reports posted on the website including
the regulation authority website in Japan and overseas.
⚫ The detailed survey regarding the waste storage was conducted while referencing the following trouble
cases.
➢

➢
➢

⚫

Japan
✓ Shimane Nuclear Power Plant: Corrosion of the drum bottom in which the low level radioactive waste (LLW) is
stored.
✓ Hamaoka Nuclear Power Plant: Coating removal and water adhesion of the drum bottom for LLW that was
carried out by Japan Nuclear Fuel Limited (JNFL).
FranceTricastin （Pierrelatte） Conversion Plant: Loss of airtight functions of the container enclosed with unnecessary
materials and equipment
Germany
Brunsbüttel Nuclear Power Plant: Corrosion of the filter installed drums

According to the results of the survey, issues were clarified in the aspect of storage management, fall
prevention at the time of earthquake and ensuring the integrity of the container. For responding to theissues, the following measures were studied and study items for reference cases and specific measures
are indicated as below.
➢➢➢

Storage management: Regular inspection and study on responding to incompatibility cases in advance. ⇒Crystallization of inspection and confirmation methods
Fall prevention at the time of earthquake: Analysis evaluation and connection installment. ⇒ Radiation dose
evaluation of scattering waste and fall prevention measures
Ensuring of the container: Prevention of damage and deterioration, confirmation of the content, pre-evaluation of
inner pressure and prevention of water adhesion ⇒ Crystallization of storage methods, removal of corrosive
substances, and study, evaluation and measures for factors of inner pressure increase©International Research Institute for Nuclear Decommissioning69

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities

Summary
Achievements until now✓

Trouble cases in Japan and overseas regarding the storage containers and the storage facilities for waste
equivalent from L1 to L3 class were searched. Then reference cases of waste storage for the Fukushima Daiichi
Nuclear Power Station were clarified by primary screening.

✓Among the clarified cases, surveys were conducted regarding the information on the container types, storage
waste, trouble occurring place and occurrence factors as well as measures for preventing re-occurrence. The
results of the survey were organized.

✓

On the basis of the organized results, issues when storing waste were considered from the aspect of storage
management, fall prevention when earthquake occurs and ensuring the integrity of the container. Reference
cases for the measures responding to the issues were studied and the study items were examined to clarify
specific measures.

Issues
✓According to the specifications of the container and waste properties, it is necessary to specify a storage methodthat would not cause damage to the container and a removal method for specific substances set by the guidelineof the storage management.

✓

Study of water management such as the use of inner container when storing waste and others is needed.

✓

Each waste is required to evaluate fall resistance at the time of earthquake, radiation dose when spreading thecontent in the container, inner pressure generation, etc. and also specific measures.

©International Research Institute for Nuclear Decommissioning70【Reference】(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Survey of trouble cases of containers and storage facilities<Example of the survey>

Example of domestic case study
The plant where the event occurred, thedate of occurrence, circumstances when
the trouble occurred, method ofinvestigation, causes pertaining to thecases that were studied has been
consolidated in a table.

Fig.1: Example of the case survey in Japan

Example of overseas case study
The plant where the event occurred, thedate of occurrence, overview of thetrouble were summarized for the casesthat were studied and consolidated in atable along with the reference source.

Fig.2: Example of the case survey in overseas©International Research Institute for Nuclear Decommissioning71

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters◼ Achievements up to FY2020


The information of the containers with filter ventilation that areused for responding to hydrogen gas generation in foreigncountries was organized.
Events of filters that are needed to consider and the measures
for the events were studied while referencing the process flow.
Tentative specifications of the filters were set.

◼ Goal

Functions of the filters when high level radioactive waste isstored in the container with the filters and evaluation of service
life of filters were examined.

◼ Details of implementation (See details on next pages)
High level of radioactive waste that is needed to be stored in
the container with filters was selected on the basis of propertiesof high level radioactive waste. The site environment in which
the storage container with filters is placed and conditions of thestorage facility were studied, and test methods and conditions
regarding functions and service life of the filters were also
considered. In addition, a verification method for maintaining
the filter functions during storage of the container with filters
was considered.

◼ Guideline to determine goal achievements


Evaluation method for the functions and service life of the filters
during storage of the container with filters are clarified.
（Target at the completion:ＴＲＬ：2 → 3）

Table 1: Measures of hydrogen gas generation conducted in foreign countries
In other countries, the moisture content (free water) is reduced to the extent possible by means of drying and cement
solidification, etc. and the generated hydrogen gas is passed through a filter, released from the vent, and managed in
the building.

US

UK

France・Ferric hydroxide sludge
Af ter cement solidif ication, the w astesurf ace is capped w ith grout andthen covered w ith a lid that isdesigned to dif f use hydrogen through・Moisture content is
the screw threads.

・At the WCS as well as the
WIPP disposal sites, amoisture content of 1vol％ or・Barium carbonate slurryless is accepted.
The process is almost similar to the(Documents are checked at the process f or the f erric hydroxide
sludge. A lid integrated w ith the f ilter
disposal sites)

decreased by drying(CSD-C: There is a track record
for moisture content of 5% orless in the hull and end piece
Hydrogen gas
is installed.
wastes compressed bodies)
generation
・Legacy m ixed waste
・The installation of a vent filter
Although the execution in case
3Stored temporarily in 3 m containers
is mandatory.
of DSC (sludge waste) is(w ith a double w all to mitigate the risk
countermeasures (Although TRU waste wasundecided, a drying processof sw elling, lined w ith concrete todisposed in sealed containers
wherein the moisture content
prevent corrosion, and w ith a f ilterin the past, there are instances installed on the lid to release the gas) will reduce to 5% or less is
f or several decades.
where the containers werebeing researched and・Plutonium contam inatedretrofitted with vent filtersdeveloped.
m aterial
thereafter)
・ Several super-compacted 200 liter
drums are stored in 500 liter drums
and stuf f ed w ith grout. A lid
integrated w ith the f ilter is installed.

Table 2: Events of filters that are needed to consider and the proposed measures
Work step

During transfer

During temporary (long-term) storage(3) Corrosion and deterioration of filter
・SUS316L or carbon composite materials that are corrosionresistant in atmospheric environment should be selected for thefilter media (sintered porous filter).
・Corrosion resistance should be confirmed w ith test pieces ofthe same material under the same conditions (or harsher
conditions).

Proposedcountermeasures*(1) Clogging of filter
・Multiple filters should be installed (multiplexing) so that even if one of thefilter gets clogged, hydrogen can be released through the remaining filters.
・The filter installation structure should be in the form of a labyrinth, etc. sothat the contents do not directly reach the filter.
・As the clogging of the filter is assumed to occur because the contents get
stirred up during transfer, a lid w ith attached filter to be used exclusively
during transfer should be used (w hen storing, that lid should be replaced
w ith a lid w ith attached filter to be used exclusively during storage).
・The impact on the storage container should be assessed assuming asituation w herein there is an explosion w ith the expected hydrogenconcentration (4 vol%).

(4) Performance degradation due to filter condensation
・The temperature and humidity should be controlled using
ventilation system of the storage facility to prevent condensation.
・If the ventilation system of the storage facility stops functioning
and condensation occurs, it is critical to restore the ventilation(2) Contamination due to dispersion of the contents
system and eliminate condensation before the hydrogen
・For fine particles, a particle collection efficiency equivalent to that of HEPA concentration in the storage container exceeds 4 vol%. It should(99.97% particle collection rate for 0.3 μm particles) is required.
be ensured that the hydrogen concentration inside the storage
・Gas component should be processed by means of a ventilation system.
container does not exceed 4 vol% w ithin the estimated time that
・Area classifications (red, yellow , green) should be developed based on is required for restoring the ventilation system. If it exceeds 4
the assumed dispersion from the filters to control contamination.
vol% (the time margin cannot be secured), it should be ensured
・Surface contamination of the storage container should be inspectedthat the ventilation system is running at all times by multiplexing(decontamination should be carried out if necessary) before transfer inside the ventilation system or by preparing an emergency pow er
the premises and before storage.
supply.

* Whether or not the vent filter can be used and the extent of measures depends on discussions with the authorities.

©International Research Institute for Nuclear Decommissioning72

(a) Investigation and evaluation of storage methods[2] Investigation of requirements of storage containers and facilities : Performance evaluation of vent filters<Study of performance evaluation of vent filters>
In accordance with the flow chart below, the deterioration of filters caused by long-term use was studied.
1. Organizing of input
conditions
2. Filter deterioration factors
and estimation of failure mode

Organizing of input conditions

Study of deterioration
factors
Clarification of filter
requirements
Estimation of failure mode

3. Filter functions andevaluation on service lifeof filter*

Study of performance
testing method

Study of integrity evaluation
methodProperties of waste and the conditions of storage
are examined.
The filter deterioration factors are clarified fromthe input condition.
The applicability of the filter requirements isclarified in consideration of properties of waste
and storage environment.
Failure mode including chronologicalenvironmental change and trouble errors is
estimated.
A performance test method that is needed toevaluate the effect on service life of filter isstudied.
A method for evaluating the integrity of filter isstudied.

* Evaluation of the filter functions aims to provide the guideline of performance evaluation and methods
for selection.
©International Research Institute for Nuclear Decommissioning73

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Procedures for collecting information on high level radioactive waste>
Waste list obtained fromthe integrated R&D
results (waste stream)

廃棄物リスト①解体廃棄物
S1 圧力容器

S2 格納容器金属

1

圧力容器

解体金属

2

蒸気乾燥器

配管類

3

気水分離器

機器類(ケーブル、盤等含む)

4

上部格子板

鉄筋

5

炉心シュラウド

ＰＣＶ本体

ＲＰＶペデスタルコンクリート（表面汚染） R/B高線量機器・配管（DHC系・AC系・FCS）

6

炉心支持板

ＰＣＶ内機器

トーラスコンクリート(表面汚染)

R/B高線量機器・配管（DHC系・AC系・RCW）

7

その他機器類

ＲＳＷ鋼板

ウェルプラグ(表面汚染)

R/B高線量機器・配管（AC系）

8

S3 格納容器コンクリート

ＲＳＷコンクリート（放射化）

ＢＳＷコンクリート（放射化）

燃料ｻﾎﾟｰﾄ

ＢＳＷ鉄筋

9ｼﾞｪｯﾄﾎﾟﾝﾌﾟ

ＲＰＶペデスタル鉄筋

制御棒案内管

11

CRDﾊｳｼﾞﾝｸﾞ

S5 建屋内コンクリート

S14 デブリ回収廃棄物

N1 4号機解体廃棄物

Dismantled waste
Rubbles, etc.

解体コンクリート

汚染粉塵（表面除染）

10

S4 建屋内金属

N2 ５・６号機解体廃棄物

N3 その他施設解体廃棄物

金属ガラ

解体コンクリート

DSPスロットプラグコンクリート

通常炉相当（S1～S5参照）

配管類

汚染粉塵（表面除染）

保温材

事故汚染物（S3～5参照）

鉄筋

T/B壁（内面）・床・柱等

PCVヘッド高温焼却設備

R/B機器

T/B壁（外面）

シールドプラグコンクリート

焼却工作建屋Rw/B壁（内面）・床・柱等

RPVヘッドサイトバンカ建屋

Rw/B壁（外面）RPV下鏡※

補助建屋

デブリ取出し関係機器類

共用プール建屋

通常炉相当（S1～S5参照）

震災前設置
集中環境施設プロセス建屋

追加遮蔽材

SPT建屋

タービン本体雑固体焼却設備

発電機本体
復水器

13

給水加熱器

14

その他機器・配管等

第二セシウム吸着装置（ＳＡＲＲＹ）

高線量配管（MS系）

第三セシウム吸着装置（ＳＡＲＲＹ-2）

16
17
18

置）

リー）

材）

ラム）

モバイル型ストロンチウム除去装置(KMPS)

炭酸塩スラリー

珪チタン酸塩系吸着材(FST,Sr-treat)

キレート樹脂系吸着材:水酸化鉄(Ly-RT)

珪チタン酸塩(ISM)

鉄共沈スラリー

活性炭系系吸着材(GAC,AGC,クラレコール)

4 銀含浸ハーシュライト(AGH)

22

Rw/B機器・配管・等

S9.2 吸着塔②（モバイルSr除去装 S10.1 多核種除去装置①（スラKURION

3 ハーシュライト(H,EH)

21

②瓦礫等

油分離装置

（KURION,SARRY）

2 界面活性剤改質ゼオライト(SMZ)

20

第2モバイル型ストロンチウム除去装置

S15 汚染土壌

使用済み保護衣

汚染土壌

8

紙・ウェス

礫・採石

プラスチック、ビニール類

浚渫土

木材

アスファルト

4 解体タンク類
5 ワイヤー類
6 空容器類

汚染粉塵（表面除染）
9 チャバサイト(IE-96)

10 珪チタン酸塩(IE-911)
11 珪チタン酸塩(EN-101,HS-726,ET-501)
12 ろ過砂+ゼオライト系:クリノプチロライト

7 工事資機材（鉄板、鉄パイプ等）
8 車両・重機

13 ろ過砂+活性炭系:アンスラサイト

9

14

10

15 ろ過砂+活性炭系:アンスラサイト

11

16 珪チタン酸塩(ET-501)

12

17

13
14
15
16

18
19
20
21
22

横置タンク
Srフィルタ(ナイロン樹脂製)
角型タンク
RO濃縮水処理設備

ろ過フィルタ（ポリプロピレン製）
固体廃棄物貯蔵庫

モバイル型Sr除去装置
スラリー安定化設備

減容処理設備SSフィルタ(ポリプロピレン製)

不燃物（ガラス，陶磁器等）
フェロシアン化コバルト(Cs-treat)

3号燃料取出用カバー
ウルトラフィルタ(ポリエーテルサルフォン製)
3号燃料取出装置
サブドレン他浄化設備

デブリ取出し関係設備
ろ過フィルタ（ポリプロピレン製）
デブリ一時保管庫・保管庫

建屋スラッジ（水抜きに伴い発生）

Csフィルタ(ナイロン樹脂製)
汚染水貯蔵タンク④その他廃棄物

Srフィルタ(ナイロン樹脂製)

S16.2 震災前廃棄物(サイトバン

物保管施設等）カ、燃料プール等）

除染装置スラッジ抜出し装置

X1 使用済み燃料

X2 デブリ

X3 環境影響物質（分別したもの）

造粒固化体（日立製濃縮廃液ペレット）

使用済み燃料デブリ

PCB

2 運転廃棄物（基準不適合or性状不明）

造粒固化体（東芝製濃縮廃液ペレット）

原子燃料（供用中）

デブリ（部材混合）除去金属類（鉛、アルミ等）

3

造粒固化体（東芝製廃樹脂ペレット）

原子燃料（使用前）

デブリ（部材付着）

アスベスト含有物184

タンク貯蔵廃樹脂類

損傷燃料

19

5

タンク貯蔵廃スラッジ類

6

タンク貯蔵廃液類

7

制御棒

8

チャンネルボックス

9

ヒューエルサポート10

中性子検出器11

使用済燃料プール

12

チャンネルボックス

13

ポイズンカーテン

14

その他

20
21
22

高性能ALPS

難燃物（塩ビ管等）
酸化チタン(GX194)

1 運転廃棄物

17

蒸発濃縮装置

脱塩用ＲＯユニット
ろ過フィルタ（ポリプロピレン製）
脱塩器

S13蒸発濃縮装置廃スラリー
蒸発濃縮装置廃スラリー

Csフィルタ(ナイロン樹脂製)
円筒フランジ型タンク

キレート樹脂系吸着材:水酸化鉄(Ly-RT)
N5 その他瓦礫等

什器類

S16.1 震災前廃棄物（固体廃棄逆浸透膜処理ユニット１
S12 フィルタ

Secondary waste
generated from water
treatment
機器類（ケーブル、電池等含む）

焼却灰

SARRY-2

除染装置スラッジ

ゼオライト系系吸着材:銀ゼオライト(AgZ）
左記に含まれない全ての瓦礫

伐採木・草

Yes

除染装置（ＡＲＥＶＡ）

S11 除染装置スラッジ

高分子系系吸着材:水酸化鉄系吸着材(CN4000)

S8 可燃物

7 ろ過砂(硅砂）コンクリートがら

3 鉄筋

S10.3 多核種除去装置③（処理カ

高分子系系吸着材:イノミジ酢酸型キレート樹脂(IRC748i)

S7 瓦礫コンクリート
6 珪チタン酸塩(TSG)

2 配管（金属）

コンクリートSARRY

S10.2 多核種除去装置②（吸着高分子系系吸着材:水酸化セリウム(ReadE2,Read-B)

珪チタン酸塩(IE-911,FST)(放水路浄化)

5 フェロシアン化コバルトハーシュライト(KH) 珪チタン酸塩(FST)(タンク浄化)

1 金属がらS6 瓦礫金属

Excluded from

検討対象から除外
consideration items

セシウム吸着装置（KURION）

S9.1 吸着塔①

1

19

No

震災後設置

湿分分離器

③水処理二次廃棄物

Excluded fuel, debris, solid waste
燃料及びデブリ、震災前固体廃棄物、
generated
before the earthquake
anddemolition waste generated
通常炉の解体廃棄物以外である。
from a normal reactor.

増設固体廃棄物焼却設備

12 CRD

15

Waste list
廃棄物リスト

鉛含有物

セシウム吸着塔一時保管施設

N4 その他水処理設備
高性能ALPS
珪チタン酸塩(HS-726,FST)
活性炭系系吸着材(K-MAC)
高分子系系吸着材(CRB05,CN4000,READ-B,WA30)

ゼオライト系吸着材(AgZ,KW-500G-7)

RO濃縮水処理設備

Objects
that have been stored in
既にベント孔/フィルタ付き容器でcontainers with vent holes and filters.

珪チタン酸塩(HS-72)
サブドレン他浄化設備

Yes

保管されているもの

ゼオライト系(A51-JHP)
珪チタン酸塩(HS-726)

Excluded from

検討対象から除外
consideration items

活性炭系(K-MAC)
高分子系(READ-B)
モバイル式処理装置
珪チタン酸塩(IE-901)(SFP浄化)
珪チタン酸塩(IE-901)(トレンチ浄化)
ゼオライト系(ET201)(トレンチ浄化)

高性能ALPS検証試験装置

No

珪チタン酸塩(HS-716)
高分子系(CRB05,WA30)
ゼオライト系(AgZ)
水酸化第二鉄系吸着材(As-CATCH2)

Other waste

その他

Yes

ゼオライト土嚢(ゼオライト)
ゼオライト土嚢(活性炭土嚢)No

デブリまたは汚染水とObjects
that have contacted withdebris and contaminated water.

15
16
17

Excluded from

検討対象から除外
consideration items

接触した履歴あり

18
19
20
21
22

Yes

•

Waste list
The gathered information is organized based on the waste list that was compiled by the integrated R&D(waste stream). On the other hand, waste objects are filtered considering the following items.
➢ Nuclear fuel, fuel debris, solid waste generated before the Great East Japan Earthquake andwaste generated from the normal reactor are not investigated under this project.
➢ Filters are assumed to be installed in the container that will be newly manufactured, therefore,
filters that have been already stored in the container with ventilation holes and filters are notinvestigated.
➢ If waste has contacted with fuel debris or contaminated water, the waste would be investigatedbecause of the possibility of α contamination. Other waste are not investigated.

No*1: Water spraying to cool the cut part is not included.

*1：切断部の冷却による散水は含まない

水中で保管されておりObjects thathave stored underwaterand handled
underwater cutting.
水中切断*1byや水中で取り扱うもの

Yes

No
Particulate
or fine particles would
粒状・または微細分
generate (including rust).
（錆を含む）が発生•

Waste objects that are needed to collect information
If waste is applicable to any one of the following items (major factors of deterioration) , the applicable
腐食性物質イオンを含む
Objects
that haveYes
contacted withsea water including corrosive ions.
waste are investigated.
No
海水との接触履歴がある➢ Existence of water➢ Existence of particle substances
Excluded from
検討対象から除外
consideration
items➢ Existence of corrosive ion•

Information acquisition of high level radioactive waste
The following information for the selected waste is compiled.
➢ Properties: Materials, concentration of radioactivity, water content, particle diameter andconcentration of corrosive ion➢ Storage condition: Storage environment, storage container, etc.

High level radioactive waste
フィルタ劣化の懸念を有する高線量廃棄物
havingconcerns filter deterioration
Fig.1: Flow chart of filtering waste objects
for investigation©International Research Institute for Nuclear Decommissioning74

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters
Waste listobjects selected from the waste list><WasteOrganizing of input conditions

1. Dem olition w aste
S1 RPV
RPV
Steam dryer
Steam separator

S2 PCV metal
Demolition metal
PipingEquipment (cables andS3 PCV concrete
Demolition concreteContaminated dust(surface
decontamination)
RSW concrete (activation)

S4 Metal in the building
Metal camera
Piping

S5 Concrete in the building
Demolition concreteContaminated dust(surface

S14 Debris collection waste
DSP slot plug concrete
Moisturizer

Reinforcing steel

decontamination)
T /B wall(inside),floor, column PCV head

BSW(activation)

R/B equipment

T /B wall(outside)

RPV pedestal

R/B high level radioactive

Rw/B wall(inside),floor,

concrete(surface
T orus concrete(surface

equipment
R/B high level radioactive

columndecontamination)
Well plug(surface

equipment
R/B high level radioactive

contamination)

Upper grid plate

plates)Reinforcing steel

Core shroud

PCV main partCore support plate

PCV inner equipment

Other equipment

RSW steel plate

Fuel support

BSW reinforcing steel

equipment, piping(ACAdditional shielding

Jet pump

RPV pedestal steel

T urbine main bodyControl rod guide pipe

Generator

CRD housing

Steam condenser

CRD

Moisture separator

Rw/B wall(outside)

N2 Unit 5 & 6 demolition

N3 Other facility demolition

Equivalent to waste

waste
Equivalent to waste

waste

generated from the normal

generated from the normal

reactorAccident contaminatedreactorN1 Unit 4 demolition waste

waste(refer to S3-5)

Installed before theearthquake
Intensive environment
facility process building
High-temperature incinerator

Shield plug concrete

Incineration building

RPV bottom plate*

Onsite bunker building

Equipment for retrieving fueldebrisAuxiliary building
Common pool building

Excluded objects because
they are equivalent to waste
generated from the normal
reactor.

Water supply humidifier

SPT building
Incinerator for miscellaneous
solid waste
Installed after theearthquake
Additional incinerator for
miscellaneous solid waste
Cesium adsorption
system(KURION)
Oil separator

cesium adsorption
Excluded objects because they have no contact Second
system(SARRY)
T hird cesium adsorption
with fuel debris and contaminated water.
system(SARRY-2)

Other equipment, piping
High level radioactive
piping(MS system)
RW/B equipment, piping

Decontamination

system(AREVA)
Reverse osmosis membrane
treatment Unit 1
Evaporation concentration

Excluded objects because they are
regarded as fuel debris.

systemDesalting RO unit
Desalting equipment
Cylindrical flange tank
Horizontal tank

2. Debris, etc.
S6 Debris, metal

Square tank
S7 Debris concrete

S8 Combustibles

S16 Contaminated soil

MetalsConcretes

Used protective clothing

Contaminated soil

Piping(metal)

Concrete

Paper

Rubble, collected stones

Plastic, vinyl

Dredged soil

Demolition tanks

LumberAsphalt

Wires

Cut trees, grass

Empty containers

Incinerated ash

Reinforcing steel

Contaminated dust(surface
contamination)

N5 Others, debris, etc.

Solid waste storage facility

Incombustible

Slurry stabilization

waste(polyvinyl chloride
Noncombustible

equipment

waste(glass,ceramic,ec.)
Equipment(cables,battery,etc.)
Appliances
Building sludge(generated bydraining)

Volume reduction equipment
Fuel removal cover for Unit 3
Fuel removal equipment for
Unit
3
Equipment
for fuel debris
retrieval
T emporary storage facility for

Construction

fuel debris
Contaminated water storage

machinery(steel plate, steel
Vehicles, heavy machinerytank
Decontamination equipment,
sludge removal equipment
T emporary storage facility forcesium adsorption tower©International Research Institute for Nuclear Decommissioning75

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Waste objects selected from the waste list>

3.Secondary w aste generated from w ater treatm ent

1
2

S9.1 Adsorption tower①

S9.2 Adsorption tower② (mobile

S10.1 Multi-nuclide(KURION,SARRY)

Sr.removal equipment)

remov al system① (slurry) system② (adsorbent)

Mobile type strontium

KURION
Surfactant modification

remov al
Silicide titanate(ISM)

Carbonate slurryzeolite(SMZ)
Second mobile type strontium
3 Harsh light(H,EH)
removal equipment
Silver impregnated harsh Silicide titanate(IE4light(KH)
911,FST)(water discharge
Ferrocyanide cobalt
Silicide titanate(FST)(tank clean-up)
5harsh light(KH)
6 Silicide titanate(TSG)

Iron coprecipitation slurry

7 Filtered sand(silica sand)8

S10.2 Multi-nuclide removal

S10.3 Multi-nuclide removal

S11 Decontamination

system③ (treatment column)

Silicide titanate adsorbent(FST,Sr-

Chelate resin typeequipment sludge
Decontamination

treat)
Activated carbon

adsorbent: ironequipment sludge

Filter(made from polypropylene)

adsorbent(GAC,AGC,kuraray coal)
Polymer system adsorbent: Ceriumhydroxides(Read E-2,Read-B)
Polymer system adsorbent:
iminodiacetic acid type chelate
Polymer system adsorbent: ironhydroxide adsorbent(CN4000)
Chelate resin type adsorbent: ironhydroxide(Ly:RT)
Zeolite type adsorbent: silver

SARRY

Ferrocyanide cobalt(cesium-treat)

Excluded objects because they have been
stored in the container with ventilation holes
and filters and are not planned to be
replaced.

501)

Filtered sand activated carbon:
anthracite

SARRY-2
Filtered sand+activated carbon:
anthracite

16 Silicide titanate(ET-501)

High-performance ALPS
Silicide titanate(HS726,FST)
Activated carbon typeadsorbent(K-MAC)
Polymer systemadsorbent(CRB05,CN4000,ReadZeolite adsorbent(AgZ,KW-5000G7)
RO concentration water

SS filter(made from polypropylene)

Ultra filter(polyether

equipment
Filter(made from polypropylene)

Silicide titanate(HS-72)
Sub drain and other filtering
equipment

Zeolite adsorbent(A51,JHP)
Silicide titanate(HS-726)
Activated carbon typeadsorbent(K-MAC)

Cs filter(made from nylon resin)
Sr filter(made from nylon resin)

Polymer system (ReadB)
Mobile type treatment
systemSilicide titanate(IE-901)(SFP cleanup)
Silicide titanate(IE-901)(trenchclean-up)
Zeolite system(ET201,trench

Polymer system (CRB05,WA30)

20

Zeolite system(AgZ)
Ferric hydroxide21adsorbent(As-CATCH2)

22

4.Other w aste

Others

S16.1 Waste generated

S16.2 Waste generated

before thebefore theearthquake(solid waste

earthquake(sitestorage facility,etc.)

banker,fuel pool,etc.)
Granulated

1 Operation waste

5

treatment system

Silicide titanate(HS716)

19

4

Filter(made from polypropylene)

equipment waste slurry

N4 Other water

verification system183treatment equipment

waste
slurryconcentration
Evaporation

clean-up)
High-performance ALPS

17

2

RO concentration water

sulfone)Sub drain and other filtering

12 Filtered sand+zeolite:clinoptilolite

15

Sr filter(made from nylon resin)concentration equipment

treatment equipment

Titanium oxide(GX194)

Silicide titanate(EN-101,HS-726,ET-

14

Cs filter(made from nylon resin)

Mobile type Sr removal equipment

10 Silicide titanate(IE-911)

13

High-performance ALPS

zeolite(AgZ)

9 Chabazite(IE-96)

11

S13 Ev aporation
S12 Filter

Operation waste(noncomformity or

solidification(concentrati
Granulated
solidification(concentrati
Granulated
solidification(waste resin

X3 Env ironment impact

X1 Spent fuelX2 Debris

Spent fuelDebrisPCB

Debris(mixed with

Removedcomponents)
Debris(adhered withmetals(lead,alminium,etc
Asbestos-containing

components)

material

Zeolite soil(zeolite)

substances(sorted)

Zeolite soil(activatedNuclear fuel(supplied)
Nuclear fuel(before use)

T ank storage waste regin Damaged fuelcarbon soil)

Lead-containing material

T ank storage waste
sludge6

T ank storage waste liquid

7

Control rod

8

Channel box

9

Fuel support

10

Neutron detector

11

Spent fuel pool

12

Channel box

13

Poison curtain

14

OttersExcluded objects because they areequivalent to waste generated fromthe normal reactor.

15©International Research Institute for Nuclear Decommissioning76

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Selection of waste objects based on their properties>
Waste to be studied
Large classification

Dismantling waste

Secondary waste
generated fromcontaminated water
treatment①
Presence of water②
Particulate matter③
Contains corrosive ions

Judgment

S1 Pressure vessel

〇

〇

〇

〇

S2 Containment vessel metal

ー

〇

〇

〇

S3 Containment vessel concrete

ー

〇

ー

〇

S14 Waste from debris collection

〇

〇

〇

〇

N3 Dismantling waste from other facilities (After theearthquake disaster)

〇

〇

〇

〇

S9.2 Adsorption vessel ② (Mobile Sr removal
equipment)

〇

〇

〇

〇

S10.1 Multi-nuclide removal system ① (slurry)

〇

〇

〇

〇

S10.2 Multi-nuclide removal system ② (adsorbent)

〇

〇

〇

〇

S10.3 Multi-nuclide removal system ③ (treatment
column)

〇

〇

〇

〇

S11 Sludge from decontamination equipment

〇

〇

〇

〇

N4 Other water treatment facilities

〇

〇

〇

〇

Medium classification➢Water existence・Underwater cutting is assumed if dismantled waste has high radiation dose.
・Waste generated from secondary water treatment is assumed to be stored in the condition
having water content because of preventing the spread of waste.
➢Having particulate substances
・Particles are assumed to generate from cutting dismantled waste.
・Particles are assumed to generate from crashing the secondary waste of water treatment.
➢Having corrosive ion
・It is assumed that both of dismantled waste and the secondary waste of water treatment haveexperience of contact with sea water.
©International Research Institute for Nuclear Decommissioning77

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters<Results of the investigation for dismantled waste> Organizing of input conditions
Waste classification
Large

Intermidiate

Demolition waste S1 RPVDemolition waste S2 PCVDemolition waste S3 PCV concrete

Demolition waste S14 Debris
collection waste

Demolition waste N3 Other demolition
waste

Waste properties
Small

Name of waste

content %
7
7
7
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

dehydration
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining

N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

N.A
N.A
N.A
N.A
N.A
N.A
12
3
7
12
7
7
7
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
Draining
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Carbon steel
Stainless steel
Concrete
Low alloy steel
Low alloy steel
Stainless steel
N.A.
]
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

RPV
Steam dryer
Steam separator
Demolition metal
PipingEquipment(cables,boards,etc.)
Reinforcing steel
PCV main body
PCV inner equipment
RSW copper plate
BSW reinforcing steel
RPV pedestal reinforcing steel

Carbon steel
Stainless steel
Stainless steel
Carbon steel
Carbon steel
Iron
Iron
Carbon steel
Iron
Iron
Iron

Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape
Block-shape

Demolition concreteContaminated dust(surface contamination)
RSW concrete(activation)
RPV pedestal concrete(surface
Torus concrete(surface contamination)
Well plug(surface contamination)
DSP slot plug concrete
Moisturizer
PCV head
Shield plug concrete
RPV head
RPV bottom head
Equipment for retrieving debrisRadioactive waste incineration facility
Cesium adsorption system(KURION)
Oil separator
Cesium adsorption system(SARRY)-2
Cesium adsorption system(SARRY-2)-3Decontamination system(AREVA)
Reverse osmosis membrane treatment
Evaporation concentration equipment
Desalting RO unit
Desalting apparatus
Cylindrical flange tank
Horizantal tank
Square tank
Solid waste storage facility
Slurry stabilization equipment
Volume reduction equipment
Fuel removal cover for Unit 3
Fuel removal equipment for Unit 3
Equipment for fuel debris retrieval
Temporary storage facility for fuel debris
Contaminated water storage tank
Decontamination equipment, sludge

Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Carbon steel
Stainless steel
Concrete
Low alloy steel
Low alloy steel
Stainless steel
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

Porous shape
Porous shape
Porous shape
Porous shape
Porous shape
Porous shape
Porous shape
Porous shape
Block-shape
Porous shape
Block-shape
Block-shape
Block-shape
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

N.A.
N.A.

Temporary storage facility for cesium
adsorption tower②
EstimatedShaperemoval equipment①
Total radiation

Water

Material

dose*(advanced
9.10E+13

1.27E+14
9.10E+13
1.27E+14
9.10E+13③
Particle④
Chloride

Other

diameter(μm)
ions(ppm)
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cuttingandchipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
Cutting and chipping
N.A
N.A
N.A
Cutting and chipping
N.A

corrosive
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

with sea
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇

N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
Cutting and chipping
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A
N.A

〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇
〇

N.A.

N.A

N.A

N.A

N.A

N.A

〇

N.A.

N.A

N.A

N.A

N.A

N.A

〇

Component
Carbon steel
Stainless steel
Stainless steel
Carbon steel
Carbon steel
Iron
Iron
Carbon steel
Iron
Iron
Iron

Occurrence factor

pH

Contacted

The inventory data is cited from back data of inventory estimation results [1] for each waste (generated from dismantlement, debris collection and water treatment)
estimated by Japan Atomic Energy Agency (JAEA/) Central Research Institute of Electric Power Industry (CRIEPI).

There are many uncertainties in non-material information.

[1] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka, Yoshikazu
Koma & Youko Takahatake (2021) Stochastic estimation of radionuclide
composition in wastes generated at Fukushima Daiichi nuclear power
station using Bayesian inference, Journal of Nuclear Science andTechnology, 58:4, 493-506,©International Research Institute for Nuclear Decommissioning78

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Compiled results of secondary waste generated from water treatment>
Waste classification

Waste properties

Intermidiate

Small

①

T otal radiation

Name of waste
Large

Material

Shapedose*(advanced
inventoryWater

S9.2 Adsorption

Mobile ty pe Sr.remov al

treatnebt

tower(mobile

equipment(KMPS)

secondary

Sr.removal

Mobile typewaste

equipment)

Sr.removal
equipment(KMPS)

S10.1Multi-nuclide-2

removal system ①

Silicide titanate(ISM)
Silicide titanate(IE911,FST )(discharge channel
Silicide titanate(FST )(tank cleanup)
Carbonate slurry
Iron cosedimentation slurry
Silicide titanate(slurry)
S10.2Multi-nuclide
removal system ②

adsorbent(FST ,Sr-treat)
Activate carbonate(adsorbent)

adsorbent(GAC,AGC,,kuraray
Polymer system adsorbent:

treatment system

Chelate resin
ZeoliteT itanium oxide
Ferrocyanide
Chelate resin
Barium sulfate

Silicide titanate(HS-726,FST )Carbonate type adsorbent(K-Silicide titanate

adsorbent(CN4000,READZeolite type adsorben(AgZ.KW-

water treatment
Sub drain and

Chelate resin

Decontamination system sludge

MAC)
Polymer system

RO concentration

500G-7)
Silicide titanate(HS-72)

Carbonate
Chelate resin
ZeoliteSilicide titanate

Zeolite system(A51-JHP)

ZeoliteSilicide titanate(HS-726)

Silicide titanate
Activateother filtering
equipment

Activate carbonate (K-MAC)
Polymer system(Read-B)
Mobile type
treatment system

Silicide titanate(IE-901)(SFP
purification)
Silicide titanate(IE-901)(trenchcarbonate
Chelate resin
Silicide titanate

Silicide titanate
clean-up)
Zeolite type(ET 201)(trench clean- Zeolite

High-performance
ALPS verification
systemmm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)

Chelate resin

iron hydroxide(Ly:RT )ALPS

mm diameter)
Granular(s ev eraliminodiacetic acid type chelate

Chelate resin type adsorbent:

High-performance

Carbonate

mm diameter)
Granular(s ev eral

mm diameter)

Ferrocyanide cobalt(Cs-treat)

N4 Other water

Silicide titanate

Granular(s ev eral

Granular(s ev eralzeolite(AgZ)
T itanium oxide(GX194)

s ludgecarbonate
Iron hydroxide

mm diameter)hydroxides

iron hydroxide(Ly:RT )Zeolite type adsorbent: silver

S11 D ec ontamination s y s tem

Silicide titanate
Calcium

Granular(s ev eral

Ceriumhydroxide adsorbent(CN4000)
Chelate resin type adsorbent:

removal system ③

Silicide titanate

mm diameter)

Cerium hydroxides(Read EPolymer system adsorbent:
resin(IRC748)
Polymer system adsorbent: ironS10.3Multi-nuclide

Silicide titanate

Granular(s ev eral

Silicide titanate(HS-716)

Silicide titanate

Polymer system(CRB05,WA30)

Chelate resin

Zeolite type(AgZ)

ZeoliteFerric hydroxide adsorbent(AsCAT CH2)

Ferric hydroxide

Zeolite soil(zeolite)

ZeoliteZeolite soil(activate carbon soil)

Activate carbon

Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)
Granular(s ev eral
mm diameter)

N.A.
N.A.
N.A.
7.72E+11
7.88E+12
5.53E+10
4.81E+10
N.A.
2.30E+10
N.A.
1.26E+11
4.81E+10
4.70E+10
4.65E+10
1.26E+11
8.17E+12
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.

Water
content %Approx.50%
Approx.50%
Approx.50%
Approx.60%
Approx.60
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.60%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%
Approx.50%②③

Estimated
dehydration

Component

treatment

Particlediameter(μm)④
Occurrence factor

Chlorideions(ppm)

Other
corrosive

ContactedpHwith seaions(ppm)

water

Draining

Silicide titanate

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

2300

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

2300

N.A.

N.A.

〇

Filter pressing

Calcium carbonate

N.A.

Pulverized powder

13000

N.A.

9.2-11.2〇

Filter pressing

Iron hydroxide

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Carbonate

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Cerium hydroxides

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

13000

N.A.

6-7

〇

Draining

ZeoliteN.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

T itanium oxide

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Ferrocyanide

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

13000

N.A.

N.A.

〇

Barium sulfate

N.A.

Pulverized powder

N.A.

N.A.

N.A.

〇

Silicide titanate

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

Carbonate

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

ZeoliteN.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

ZeoliteN.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Activate carbonate

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

ZeoliteN.A.

Pulverized powder

17000

N.A.

N.A.

〇

Draining

Silicide titanate

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

Chelate resin

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

ZeoliteN.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

Ferric hydroxide

N.A.

Pulverized powder

6350

N.A.

N.A.

〇

Draining

ZeoliteN.A.

Pulverized powder

N.A.

N.A.

N.A.

〇

Draining

Activate carbon

N.A.

Pulverized powder

N.A.

N.A.

N.A.

〇

Centrifugal
separation
Draining

The inventory data is cited from back data of inventory estimation results [1] for each waste (generated from dismantlement, debris collection and water treatment) estimated by JAEA/CRIEPI.
Water quality data was cited from a project report of a guideline of secondary waste generated from water treatment [2], FY 2015 subsidy project of safety technology measures for nuclear
power station.

••

Conservative evaluation of water quality (chloride ion, etc.) was conducted and the initial values calculated at beginningof treatment were extracted.
There are many uncertainties in non-material information.

[1] Daisuke Sugiyama, Ryo Nakabayashi, Shingo Tanaka, Yoshikazu Koma & Youko Takahatake (2021) Stochastic estimation of radionuclide composition in wastes generated at
Fukushima Daiichi nuclear power station using Bayesian inference, Journal of Nuclear Science and Technology, 58:4, 493-506,[2] Project report of a guideline of secondary waste generated from water treatment, FY 2015 subsidy project of safety technology measures for nuclear power station. Project
report Japan Atomic Energy Agency, March 2016
©International Research Institute for Nuclear Decommissioning79

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Clarification of storage conditions: Storage container>

Material: SS400
Inner surface: Polyethylene lining
External surface: Resin coating
Volume: Approx.5.1m3 (Estimating
5cm-thickness)

Vent<Image of storage container>
Vent

Iron coprecipitation

Approx. 1.5m
7 pieces

Approx. 2m
Inner container

Carbonate

Approx.2m

Fig.2: Illustration of storing secondary waste generated from water treatment *2

A storage case model of the storage container used forhydrogen generation evaluation in FY2020

Dehydrate processing for slurry by using a filter press

7 pieces

Dismantled waste

Inner container

Threaded portion(Installed on the lower surfaceof the HIC lid)

Taper threaded portion

Inside of HIC lid(Installed on the HIC housing)

Filter housing

Fig.1 Illustration of dismantled waste

・Dismantled waste is placed in the inner container and then stored into
the storage container.
・The inner container is assumed to be a grid-like rack to avoid
accumulating hydrogen.
・Materials of the storage and inner containers are not yet decided.
Carbon steel may be used same as for the secondary waste generated
from water treatment.
*1: Final report of subsidy project of Decommissioning and Contaminated Water Management in
FY2016, Upgrading of Fundamental Technology for Retravel of Fuel Debris and Reactor
Internals.
*2: Drawing the illustration while referencing the information provided by TEPCO Holdings.

Lateral gap (1.9mm)

Inside of the HIC
Filter lid (serving as face plate)
Retainer ring

Filtering material

Filter surface

Fig.3: Illustration of filtering part of the storage container

Filters should be installed in the lid of the storage container bytightening a screw after placing in the housing part with similar uses
as the HIC installation.

➢ The dismantled waste and secondary waste generated from water treatment are stored in dedicated containers installed ventilation.
©International Research Institute for Nuclear Decommissioning80

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers
and facilities: Performance evaluation of vent filters
Organizing of input conditions<Clarification of storage conditions: Storage container>
Using carbonate slurry that is assumed to be the highest dose, temperatures inside HIC
were measured when storing in HIC.*1
Table 1: Result of measuring temperatures of carbonate slurry
Location of measurement#

Temperature

Lower plate

Stored on: November 1, 2014
Measured on: April 10, 2015

Approx. 160mm

500mm (near the surface of water)
Approx. 500mm

Outdoor temperature: 13.9°C

Near the bottom surface
Carbonate slurry#: Distance from the lower plate*1: Resource from the TEPCO Holdings website (Verification of long-term integrity of HIC, 2018), p.7

Fig.1: Simplified drawing of High
Integrity Container (HIC)

Temperatures inside HIC are almost the same as outside temperatures.
➢Temperatures inside the container are assumed to be same value as
temperatures of the storage environments.

©International Research Institute for Nuclear Decommissioning81

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Organizing of input conditions<Clarification of storage conditions: Storage environments>

Possibility of changes after the stage of basic design

Fig.1: Dismantled waste treatment process*1
Fig.2: Processing of secondary waste generated fromwater treatment*2

Placing of waste in the storage building*1: Final report of subsidy project of Decommissioning and Contaminated Water Management in FY2016,
Upgrading of Fundamental Technology for Retravel of Fuel Debris and Reactor Internals.
*2: Drawing the illustration while referencing the information provided by TEPCO Holdings.

➢ The entire process from placing to storing waste is assumed to be conducted inside the building.
➢ The storage building is assumed to be equivalent to the No.9 storage house for solid waste that iscurrently under operation.
・Having shielding functions of the building and responding to measures for preventing the spread ofradioactive materials.
・Installing ventilation system (with air supply)
・Assuming that outside air temperatures are 0 – 40 ℃, though the temperature control system is notindicated.
©International Research Institute for Nuclear Decommissioning82

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Study of deterioration factors<Definition of deterioration mode>

Functions required the ventilation filter of the storage container are indicated below.
• Functions to release hydrogen that would generate in the process of water radiolytic ＝Hydrogen permeable function (hydrogen permeable rate)• Functions to confine particles including radioactive substances= Particle collecting function(collection efficiency)
Filtering deterioration
Hydrogen permeable Cross section of the filter
function

Hydrogen is not permeable.
＝A filter is clogged.
Particle collecting
function

Particles cannot be collected.
⇒ A filter is damaged.

HydrogenParticles
Fig.1: Conceptual image of filtering function

The deterioration mode defines filter clogging and damage.
The cause and event of filter clogging and damage were examined based on the storage
environments of solid waste.
©International Research Institute for Nuclear Decommissioning

83

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters
Study of deterioration factors<Study on the causes and events of filter deterioration (damage and clogging)>
・The storage environments of solid waste were classified into five environmental categories while referencing the safety
evaluation scenario creation method (FEP*1). (Radioactivity in the environment and thermal, stress, chemical environments)
・Assumed causes and events of the filter deterioration were compiled based on each environmental classification.
Environmental
classification

Radioactivity
in environment

CausesRadiolysis

Radioactivity

Events

Detailed events

Hydrogengeneration
Hydrogenembrittlement
Radiation
deterioration

A filter would be clogged by hydrogen generation due to radiolysis of solid waste and particle rise.
A filter would be damaged by filter clogging and increase of inner pressure, which is caused by hydrogen generationand particle rising due to radiolysis.
Hydrogen embrittlement would occur and damage by a long-term exposure of hydrogen that is generated fromradiolysis of water accompanied by waste.
Filter materials would deteriorate and be damaged by radiation of solid waste.

Waste temperature would increase due to decay heat. Heat conduction would increase filter temperature and the filter
would be damaged by thermal alteration.
Thermal alteration Decay heat would increase waste temperature and the head generated by increased temperature would deform the(container)
container. A fixed filter would be damaged by stress.
Thermal alteration

Thermalenvironment

Decay heat

Heat convection
Stressenvironment

Chemicalenvironment

Vibrations caused bytransportation andearthquake

Corrosion

ImpactDamageDamageDamageDamageWhen transporting filters or earthquake occurs, the container and filters would be deformed and damaged by impacts
of collision / fall, and the filter functions would be lost.

DamageDust production by
A filter would be clogged by flying particles which are caused by vibrations of transportation and earthquake.
vibration
Thinning
A filter would be clogged by adhesion of corrosive substances that generate in the thinning process.
Oxide film
A filter would be clogged by generating oxide film on the surface of a filter fiber due to corrosion and increasing a fiber
generation
diameter (decreasing a hole diameter).
A filter would be clogged by accumulating corrosive substances on the filter that is exposed with steam containing
Salt deposition
corrosive substances for a long-term.

UV deterioration

DamageClogging

Microbe such as mold would generate inside/outside a filter and a biofilter would be created, therefore, a filter would
Microbial film
be clogged and hydrogen cannot pass through the filter.
A water film would be created on the filter surface due to differences of temperatures inside/outside a filter, therefore,
Water film
a filter would be clogged and hydrogen cannot pass through the filter.
Differences between
A water film would be frozen when a filter is placed under circumstances below freezing point due to air conditioning
inside the container andFreezing
failure after a water film is created on the filter surface due to condensation. A filter would be clogged when inside aexternal environments
filter is filled with ice.
Aerosol deposition A filter is clogged by accumulating flying aerosol ,such as sea salt particles, on the external surface of a filter.
A filter would be clogged by flying particles due to the occurrence of the flow caused from differences of pressure
Spread of particles
inside/outside a filter.
UV in the environment

Clogging

A filter would be clogged by flying particles due to heat convection.

Electric corrosion A film would be clogged by corrosion products generated from a filter electric corrosion.

External
environment

Deterioration
mode

Filter materials would be deteriorated and damaged by UV in the environment.

Clogging
Clogging
Clogging
Clogging
Clogging
Clogging
Clogging
Clogging
Clogging
Clogging
Damage*1 A method for compiling scenario Feature, Events and Process are compiled to create a list.

18 deterioration events in total were extracted.
©International Research Institute for Nuclear Decommissioning

84

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements

Tentative specifications were provided under the previous project (FY2018) as below.*1➢Filtering materials → Corrosion resistance materials (SUS316L and carbon composites)➢Particle collection function → Equivalent to HEPA function➢Hydrogen permeable function → Hydrogen concentration in the container should notexceed more than 4%.
*1: Result of subsidy project of Decommissioning and Contaminated Water Management for FY2018 (Research on Treatment and Disposalof Solid Waste), report of FY2019 research meeting p.33

The following items were investigated and the filter
requirements were clarified.

In addition to the above tentative
specifications, specifications ofgeneral air filters were also
investigated.

A) Use applications of major
manufacturersB) Manufacturing methods of filtering
materials
C) Particle collection functionsD) Hydrogen permeable functions©International Research Institute for Nuclear Decommissioning85

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters
Clarification of filter requirements<Investigation of air filter specifications: Ex.1>
A) Use applications of major manufactures
Examples of use applications in nuclear facilities are indicated as below.
・U.S.: TRU waste container at the Waste Isolation Pilot Plant (WIPP) site.
・France: Hull end pieces and sludge waste container at the La Hague site
・U.K.: Waste containers such as for sludge and cladding tube
・Japan: High Integrity Container (HIC) for secondary waste generated fromcontaminated water processing

Major manufacturers are indicated as below.
・Nuclear Filter Technology（NFT)
・UltraTech
・Poral etc.
Fig. 1 Sintered metallic filter manufactured by Poral*1*1 Poral catalog （brochure-filtres-poral.pdf）

B) Manufacturing methods and filtering materials
The following materials were used in theterm of corrosion resistance.
・Sintered metal
・Carbon composite

C) Particle collection functions
In case of using the US WIPP waste container, specification values of theparticle collection functions are indicated for each type of the container.*2
The particle collection functions are indicated as an example of certification
product as below.
＞99.97%@0.3-0.5μmDOP(specified amount of flow: 210ml/min@330 Pa）
Equivalent to the HEPA functions.
Table 1: Example of collection functions of the filter for the waste
container at the US WIPP.
RH-TRAMPAC: that
CH-TRAMPAC:
TRUPACT-III:
the TRU waste
The TRU waste
TRAMPACTRU:
container that cancontainer that can
The waste transfer
be remotely
be directly handledcontainer for waste
handled.

CH-:Contact-Handled
RH-:Remote-Handled
TRAMPAC: Transuranic
Waste Authorized
Methods for Payload
Control
TRUPACT-III:Transuranic
Package Transporter III

The details of the manufacturing method are
not opened to the public. Generally, when
Particle collection
>99.9
>99.5
>99.9
using sintered metal, a filter is manufactured
function (%)
by connecting metal fibers and metal
Fig.2: Conceptual image D) Hydrogen permeable functions
particles that are processed in highof connection by highWhen using the filter for the waste container stored at the U.S. WIPP(Waste
temperatures.
temperature processing Isolation pilot Plant), Hydrogen diffusivity mol/sec/mol-fraction is defined as
the number of hydrogen mol that mol concentration gradient is set to passthrough a filter per one second. These values are measured by tests.*3＊2 DOE/WIPP 11-3384 “CBFO Approved Filter Vents” U.S. Department ofFilters with high-hydrogen permeability are included from x/2 to 5x and 25x
Energy Revision 2 March 2011 https://www.osti.gov/biblio/1041227
as a standard of NFT products and X（3.70×10-6m/s/mf).
＊3 Terry Wickland and John Schierloh “HYDROGEN DIFFUSIVITY
Hydrogen spreading factors of the Poral filters made in France are used as
THROUGH DRUM AND LINER VENTING FILTERS” WM Symposia 1998
an index. The index of the hydrogen spreading factors 3.66×10-6 (m2/s) is
SESSION 27 ,37
https://xcdsystem.com/wmsym/archives//1998/html/sess27/27-37/27-37.htm
included.
©International Research Institute for Nuclear Decommissioning86

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements<Investigation of air filter specifications: Ex. 2>
A) Use applications of major manufactures
Use applications
・Film forming, synthetic fiber spinning
・Synthetic fiber (Filtration of raw materials)
・Various air filters (High temperature gas, exhaust gas, etc.)
・Air equipment filter (Rectification, dispersion)
Etc.

Major manufacturers
・Fuji Filter Mfg. Co., Ltd.
・Kodan Kinzoku Co., Ltd.
・Nippon Seisen Co., Ltd.
・Nichidai Corporation
・Mossfil Co., Ltd.

Etc.

B) Manufacturing methods and filtering materials
Table 1

(ⅰ)

(ⅱ)

Type, manufacturing process, characteristics, filter material of the filter

Type(i) Sintered product inpowder form

(ⅲ)

(ⅳ)

Fig. 1
Cross-sectional structure of filter(i) Sintered product in powder form (ii) Long fiber sintered product(iii) Short fiber sintered product (iv) Multi-layered wire meshsintered product
Sintered metallic filter manufactured by Nippon Seisen Co., Ltd.* 1(ii) Long fiber sinteredproduct(iii) Short fiber sinteredproduct(iv) Multi-layered wire
mesh sintered product

Manufacturing process /Characteristics

Filter material

Sintered metal powderHigh strength

SUS316, SUS316L, SUS304, Al, Bronze, Cu,
Ti, Ni, 600 alloy, etc.

Sintered long metal fibers
Extensive filtration precision
Various shapes can be made

SUS316, SUS316L, SUS304

Sintered short metal fibers
Sub-micron filtration is possible
Suitable for gas and low viscosity
fluidsSUS316L

Sintered metal wire mesh
High strength

SUS316, SUS316L, SUS304, Cu*1 Nippon Seisen Co., Ltd. Website (https://www.n-seisen.co.jp/products/filter/metal/)C) Particle collection functions
Filtration precision differs depending on the manufacturer or type of filter, but with shortfiber sintered product of 0.1 to several hundred μm, sub-micron order particles (equivalent
to HEPA) can be collected as well.
Filtration precision is just the official name. It is a value that the manufacturer arbitrarily
decides based on the weight removal efficiency with respect to all particles that are of acertain size or larger. Its evaluation method has not been specified in JIS, etc.

D) Hydrogen permeable functions
Although it has not been clearly specified, since the pores of the sinteredproduct are to the order of several μm, and the porosity is 40 - 80%, thesize of the pores is 5μm, air flow rate is 170mL/min, and differential
pressure is 294Pa, it is presumed that its hydrogen permeation function
is comparable with vent filters used overseas.
(All values are in accordance with Nippon Seisen catalog values)©International Research Institute for Nuclear Decommissioning87

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements<Investigation of air filter specifications: Ex. 3>

A) Use applications of major manufactures

Use applications
・Deodorization
・Removal of foreign material
from high-temperature gas
・Gas separation
Etc.

Major manufacturers
・NGK Insulators, Ltd.
・IBIDEN Co., Ltd.
・Noritake Co., Ltd.
Etc.

C) Particle collection functions

Fig. 1 Ceramic filter manufactured by Nihon Gaishi*1

The collection efficiency is low for a particle size in the range of 100nm - 300nm, but
as the particles get collected and the pores get filled with them, collection efficiency isalmost 100% for all particle sizes. Hence it is presumed that collection efficiency
equivalent to HEPA is achieved by controlling the pores.

＊1 NGK Insulators, Ltd. Website(https://www.ngk.co.jp/product/dpf.html)

B) Manufacturing methods and filtering materials * 2, 3
・Weighing of raw material, mixing, kneading, extrusion, drying,finishing(sealing), firing, outer diameter polishing and coating

・The following materials are used.
Alumina (Al2O3): 0.004 - 15μm
Zeolite: nm level (molecular sieve)
Cordierite (2MgO, 2Al2O3, 5SiO2): 2 - 100μm
Silicon carbide (SiO2): 10μm
*2 “Systematization of industrial techniques pertaining toporous fine ceramics”
Systematization Report by the National Museum of Natureand Sciencevol 12 March 2008
*3 Transactions of The Japan Society of MechanicalEngineers Series B 72-721 (2006-9)

Fig. 2 Initial performance of DPF (Diesel
particulate filter)(If mock particles with an average particle size of
170nm, number density 6×106[cm-3] are supplied
at 10[NL/min])*2

Fig. 3 Correlation between the particle sizeand collection efficiency(The collection efficiency increases when theamount of particles collected increases) *2

D) Hydrogen permeable functions
Although it has not been clearly specified, it is presumed that the hydrogen
permeation function is sufficiently high since the resistance at the time of gas
permeation is low at 40-60Pa when exhaust gas is blown at 10NL/min.
©International Research Institute for Nuclear Decommissioning88

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements<Investigation of air filter specifications: Ex. 4>A) Use applications of major manufactures
Use applications
・Clean room
・Food processing plant・Hospital
・Nuclear facility

Major manufacturers
・Japan Air Filter Co., Ltd.
・Kondo Kogyo Co., Ltd.
(Cambridge Filter Corporation)
・Nitta Corporation
・Air Products Ltd.
・Oshitari Laboratory Inc.

C) Particle collection functions

Fig. 1
HEPA filter for air-conditioning facilities
manufactured by Cambridge Filter Corporation *1*1 Cambridge Filter Corporation website (https://cambridgefilter.com/pdf/)

B) Manufacturing methods and filtering materials
Fibers of diameter 1μm or less are spun by means of melt blowing*1or electrospinning*2, with a filling factor of approx. 10% (porosity 90%), and pore size of
about several μm.
The material used is mainly glass fiber, but depending on the use application,polymers (polypropylene, olefin, PTFE, etc.) are used.
#1 Melt blowing
A technique in which molten polymer is drawn into fibers using high speed
and high temperature gas, and the fibers are spun.
#2 Electro-spinning:
A technique in which high voltage is applied between the nozzle in front of acontainer of polymer solution and the target electrode, due to which theelectrically charged polymer solution turns into extremely thin fibers thatemerge from the nozzle, and are then spun.

In the case of HEPA filter, it is specified in JIS8901 thatfor particles of size 0.3μm, the collection efficiency is 99.97％ or more.
In the case of ULPA filter, it is specified in JIS Z 8122 thatfor particles of size 0.15μm, the collection efficiency is 99.9995% or
more.
Also, it is specified that the initial pressure drop is 245Pa or less respectively.
In the case of a HEPA filter, when it is used in a nuclear facility in particular,
it is specified in JIS Z 4812 thatfor particles of size 0.15μm, the collection efficiency is 99.97% or more.
Also it is specified that the initial pressure drop is 245Pa or less when there isstandard air volume (0.4 - 32m/s), whereas in the case of large air volume type
HEPA filter (28-56m/s), the initial pressure drop is 300Pa or less respectively.
This performance is not unequivocally decided for the filter material. It can be
specified depending on how the material is woven.

D) Hydrogen permeable functions
Although it has not been clearly specified, even when the gas flow rate isseveral tens of m/s, since the pressure drop is only of about several 100Pa,
and the porosity and the size of the pores are comparatively large, it ispresumed that hydrogen permeability function is sufficiently high.
©International Research Institute for Nuclear Decommissioning89

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements<Results of the investigation related to fine particle collection function>
Filters in the range of medium efficiency to ULPA having HEPA equivalent efficiency comparable to overseas vent filters, can be
manufactured using any of the metals, ceramics, glass fiber, polymers that are used in general industry.
Scope of investigation
Metals (Vent filters already used overseas)

Metals (Short fiber sintered product)

Ceramics

Glass fiber, polymers
Aerosol

Bacteria

Tobacco smoke

PollenCombustion nucleus

Virus

CementFly ashWheat flour

Carbon black
Soot

Electromagnetic wavelength
X-ray
Corresponding air
filterUltra-violet rays

Raindrops

Fog

Visible light

Infrared rays

ULPA filter
HEPA filter
Medium efficiency filter
Filter for coarse particles

Particle size (μm)

Fig. 1

Correlation between filters that can handle aerosols, etc. and the particle size *1*1 Ueshima, “Air filter unit efficiency”, Aerosol research, Vol. 4, No. 4 (1989)©International Research Institute for Nuclear Decommissioning90

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Clarification of filter requirements<Results of the investigation related to hydrogen permeation function>
The actual hydrogen permeation and diffusion coefficient for the filters used in the general industry has not been mentioned in
the catalogs, etc.
However, when compared with overseas vent filters, in principle, the efficiency of metal filters is the same, and that of filters ofother materials is presumed to be equivalent or more.

Metal
Ceramics

Glass fiber, polymers
Hydrogen permeation anddiffusion coefficient m2/s

10-6
Hydrogen permeation and diffusion
coefficient of overseas vent filters

10-5

10-4
Diffusion coefficient of hydrogen in air(6.27x10-5m2/s @20℃)( 3.66x10-6 m2/s)

Fig. 1

Range of hydrogen permeation and diffusion coefficient of the investigated filters

Hydrogen can be discharged easily©International Research Institute for Nuclear Decommissioning91

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
Hydrogenembrittlement

Radiation
deterioration

Thermalalteration

Heat
distortion(Container)

ImpactWallthinning

Electric
corrosion

UV deterioration

DamagedBlocking

Hydrogengeneration

Heat
convection

Dustformation
due tovibrations

Oxide film
formation

Salt
deposition

Dispersionof fineparticles

Aerosoldeposition

Microbial film

Water film

Freezing

Fig. 1 Study of events leading to filter deterioration (damage, blockage)
Based on the input conditions (No. 81-82) , since internal temperature = outdoor temperature (0-40°C),
・Thermal alteration/ heat distortion: No notable change in metal/ ceramics/ glass / polymer at said temperature.
・Oxide film: Oxide film of significant thickness is not formed at said temperature.
・Heat convection: Significant heat convection does not occur at said temperature.

According to the input conditions (No. 82), all processes from collection to storage are carried out indoors.
・UV deterioration: Does not occur as there is no exposure to UV because the processes are carried out
indoors.
・Aerosol deposition: Aerosol deposition does not occur due to the air-conditioning system (with air filter).
・Microbial film: Microbes do not get mixed up due to the air-conditioning system (with air filter).

18 deterioration events ⇒ Of which 7 were excluded leaving 11 events to be studied.

©International Research Institute for Nuclear Decommissioning

92

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
Table 1 Study of failure mode scenarios by filter degradation events
During collection

Failure mode scenarios
During transportation

Hydrogengeneration

Unlikely to get clogged due to hydrogengeneration in a short time

Unlikely to get clogged due to hydrogen generation in ashort time

Hydrogenembrittlement

Hydrogen embrittlement is unlikely in a shorttime.

Hydrogen embrittlement is unlikely in a short time.

Radiation
deterioration

Radiation deterioration is unlikely in a short
Radiation deterioration is unlikely in a short time.
time.
When the lid is being closed, the lid or filter
During transportation, the container or filter gets distorted /gets distorted / damaged due to the impact of adamaged due to the impact of a collision/fall, etc. in thecollision/fall, etc. in the event of an earthquake,
event of an earthquake, and the filter loses its functionality.
and the filter loses its functionality.

ImpactDust formation
Unlikely to get clogged because of dust
due to vibrations formation due to vibrations in a short time

Unlikely to get clogged because of dust formation due tovibrations in a short time

During storage
As hydrogen gas stream resulting from radiolysis of water is generated over along period of time, fine particles of solid waste get dispersed, the filter getsclogged and filtration functionality is lost.
The generated hydrogen gets absorbed, water is formed and theconcentration becomes such that embrittlement occurs during the storage
period, leading to damage and thereby loss of filtration functionality.

The radiation energy gets absorbed leading to deterioration, the filter material
gets damaged and filtration functionality is lost.
During transportation, the container or filter gets distorted / damaged due tothe impact of a collision/fall from a height of several m, etc. in the event of anearthquake, and the filter loses its functionality.
During transportation, if the container falls in the event of an earthquake, fineparticles scatter and adhere to the surface of the filter. The filter gets clogged
and the filtration functionality is lost.
Moisture adheres to the filter due to condensation or due to the water in thewaste, and along with it chloride ions or particles of waste adhere to the filter
as well. Pitting corrosion or crevice corrosion takes place, corrosion products
are formed that clog the filter due to which filtration functionality is lost.
Water contained in waste gushes towards the outside of the container due todifferential pressure or impact, it adheres to the filter and after drying salts get
deposited on the filter thereby clogging it, due to which filtration functionality is
lost.

Wall thinning

Wall thinning is unlikely to occur in a short time.Wall thinning is unlikely to occur in a short time.

Salt deposition

Salt deposition is unlikely to occur in a shorttime.

Electric corrosion

Electric corrosion is unlikely to occur in a short
Electric corrosion is unlikely to occur in a short time.
time.

The filter gets wet due to condensation or due to the water in the waste, andchloride ions adhere to it or the pH becomes alkaline (if the contents containconcrete), as a result of which electric corrosion occurs. The formed corrosion
products clog the filter and the filtration functionality is lost.

Water film

Condensation is unlikely to occur in a shorttime.

Condensation is unlikely to occur in a short time.

The filter gets clogged due to the water film formed because of condensation,and the filtration functionality is lost.

Freezing

Freezing is unlikely to occur in a short time.

Freezing is unlikely to occur in a short time.

During winter, if formation of water film due to condensation and failure of the
air conditioning system occur simultaneously, the filter gets clogged due tofreezing and the filtration functionality is lost.

Salt deposition is unlikely to occur in a short time.

During collection, the operation is carried out inDuring transportation, the operation is carried out in an
Due to the impact of the ventilation system of the storage facility, a current
an environment with uniform pressure. Since
Dispersion of fineenvironment with uniform pressure. Since the environment gets generated inside the storage container, due to which fine particles scatterthe environment is such that differential
particles
is such that differential pressure does not get generated, the and adhere to the surface of the filter. The filter gets clogged and the filtrationpressure does not get generated, the filter
filter does not get clogged.
functionality is lost.
does not get clogged.

: Scenarios wherein occurrence offailure is considered possible

Degradation events and failure mode scenarios were developed considering the input conditions and the processes
leading up to collection.
©International Research Institute for Nuclear Decommissioning

93

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
Table 1

Evaluation of occurrence of degradation by degradation event and filter material of the filter (determination of whether or
not the testing methods needs to be studied)

Typical filter materials were selected based on the results of systematic screening of filters
Metal
Stainless Nickel-based
steel
alloy

Ceramics

Glass

Polymer

Titaniumalloy

Alumina

Cordierite

ZeoliteE Glass

C Glass

S Glass

D Glass

Polyethylene Polypropylene

PTFE

Hydrogengeneration
Hydrogenembrittlement
Radiation
deterioration

－

－

－

－

－

－

－

－

－

－

－

－

－

－

－

x

－

x

x

－

－

－

－

－

x

－

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

x

x

x

Impact

x

x

x

x

x

x

x

x

x

x

x

x

x

Dust formation due
to vibrations

x

x

x

x

x

x

x

x

x

x

x

x

x

Wall thinning

x

x

x

x

x

x

x

x

x

x

x

x

x

Salt deposition

x

x

x

x

x

x

x

x

x

x

x

x

x

Electric corrosion

－

－

－

－

－

－

－

－

－

－

－

－

－

Water film

x

x

x

x

x

x

x

x

x

x

x

x

x

Freezing

x

x

x

x

x

x

x

x

x

x

x

x

x

Dispersion of fineparticles

－

－

－

－

－

－

－

－

－

－

－

－

－

11 degradation events remaining upon checking
against the input conditions

➢
➢
➢
➢

―

Testing methods do not need to be studied as there is no possibility of this occurring✓

Testing methods do not need to be studied as even though the event occurs, the extent of progressof generation can be estimated based on the investigation results(# In this case, it is necessary to evaluate filter performance over time using theoretical formulas)x

Testing methods need to be studied

Hydrogen generation, electric corrosion, dispersion of fine particles are - with no impact on filter degradation (study not required)Hydrogen embrittlement is - for all except titanium alloy, cordierite, zeolite, and polypropylene, with no impact on filter degradation (study not required)
Radiation deterioration is ✓ for all except polymer material. Testing methods do not need to be studied as the extent of progress of generation can be estimatedTesting methods need to be studied for the above-mentioned events that are marked as x other than those that are marked as ✓ or -Details pertaining to the reasons that led to such an evaluation are provided in the following slides.
©International Research Institute for Nuclear Decommissioning

94

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on the filter〇 Hydrogen generation
・As hydrogen gas stream resulting from radiolysis of water is generated overa long period of time, fine particles of solid waste get dispersed, the filter gets
blocked and the filtration functionality is lost.
・The impact at the time of storage of secondary waste generated fromcontaminated water treatment which has a higher proportion of fine particles
than dismantled waste was considered*.

Storagecontainer

FilterSecondary waste
generated fromcontaminated water
treatment(Carbonate slurry,iron co-precipitation.
etc.)* If the system of secondary waste generated from contaminated water
treatment does not have any impact, it is believed that there will be no impact
from dismantled waste either.
(An event that occurs inside the storage container)・Hydrogen is generated from the entire filled layer of secondary waste
generated from contaminated water treatment, and that increases thegaps between the waste.

Height1m

Hydrogengas stream

Internal dimensions 1.3m × 1.3m[1]Fig.1 Schematic diagram of thegeneration of hydrogen stream

Flow increase・If the hydrogen gas stream is large, fine particles start floating within
the layer.
・The speed at which the particles start floating is the same as the speed
at which the particles settle, and can be calculated using the Stokes’
Law given below.

Estimation formula[1]： vs = dp2(ρp-ρf) g /18μ
dp: Average particle size (m), ρp: Particle density (kg/m 3), ρf: Fluid density (kg/m 3),
g: Gravitational acceleration (m/s2), μ: Fluid viscosity (Pa・s)

Dispersion
board
Fluid
In the form ofa fixed bed

Incipient
fluidization(Particles rise)

Particles
circulateFig. 2 Schematic diagram of the rising of particles [2]Incipient fluidization speed is calculated and it is studied whether or not fluidization canhappen due to hydrogen gas stream[1]: Assumed using information related to secondary waste produced during contaminated water treatment as reference[2]: Kohei Ogawa, “Series <New Chemical Engineering> 1, Analysis of the movement of fluids” First Edition, Asakura Publishing Co., Ltd., (FY2011), p51, 58©International Research Institute for Nuclear Decommissioning95

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode

<Computational conditions>
・Particle size (0.1μm - 10μm)
The size of small particles that can easily get dispersed is set at 0.1μm which is inclusive of
the standard particle size of 0.3, 0.15μm with respect to which the performance of HEPA andULPA was evaluated.
・Particle density (1-8g/cm3)
Set up in the range that is inclusive of carbonate slurry (approx. 3g/cm 3) that is assumed to
be stored, concrete (2-3g/cm3), stainless steel (approx. 8g/cm3)
・Fluid viscosity and density: The values at 20°C are used
⇒ The terminal velocity when fine particles float is in the range of 10-7 - 10-2(m/s)

Speed at which particles start floating (m/s)■ Investigation of the impact that each degradation factor has on the filter〇 Hydrogen generation(Terminal velocity when fine particles float)
Calculated for the range of ordinary waste density and particle size

Flow velocity whenfine particles float
8g/cm3More than 3
digits smaller
Flow velocity resulting
from hydrogen generation(Flow velocity resulting from hydrogen generation)
・The hydrogen generation rate is considered to be 1.6E-6(L/s) which is assumed while evaluating
the amount of hydrogen generated.
・The hydrogen generation flow rate is calculated using the following formula
Flow rate (m/s) resulting from hydrogen generation= Hydrogen generation rate/1000 (m 3/s) / Cross-sectional area of passage (m2)・The secondary waste generated from contaminated water treatment (□1.9m×1.9m)[1],
dismantling waste (□1.3m×1.3m)[2] are used for the cross-sectional area of the passage
・The cross-section of the container for dismantling waste which has a higher flow rate resulting
from hydrogen generation (smaller cross-sectional area) is used for calculation⇒ The flow rate resulting from hydrogen generation is 9.5E-10(m/s), which is sufficiently
lower than the speed at which particles start floating

5g/cm33g/cm32g/cm31g/cm3

Particle size (μm)

Fig. 1 Correlation between the flow rate at
which fine particles start floating and the flowrate resulting from hydrogen generation

It was determined that the filter degradation effect caused by dispersion of fine particles due to hydrogen generation can be ignored.
[1]: Assumed using information related to secondary waste produced during contaminated water treatment as reference[2]: Subsidy Project of Decommissioning and Contaminated Water Management in the FY2016 Supplementary Budget“Upgrading of fundamental technology for retrieval of fuel debris and internal structures” Final Report©International Research Institute for Nuclear Decommissioning96

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode

The partial pressure of hydrogen inside the container is calculated using PV=nRT based on
the hydrogen generation rate in Table 1.
Assuming that the temperature is 30°C, the hydrogen pressure is
0.013Pa in the case of waste (concrete) generated while collecting the pressure vessel andfuel debris, and 0.0006Pa in the case of secondary waste generated from contaminatedwater treatment. The hydrogen pressure at 300°C becomes approx. double.
Thus, the hydrogen pressure inside the container at 30°C is 0.063Pa in the case of waste(concrete) generated while collecting the pressure vessel and fuel debris, and 0.0006Pa in
the case of secondary waste generated from contaminated water treatment.

Hydrogengeneration rate
(L/sec)Hydrogen generation rate

Pressure
vessel(Metal waste)

Waste generated
while collecting
fuel debris(Concrete)

Secondary waste
generated fromcontaminatedwater treatment(Carbonate slurry)

1.6E-6

1.59E-6

7.36E-8

In the case of stainless steel and nickel-based alloys
・Considering the hydrogen pressure inside the container, based on Table 1, the hydrogen solubility of SUS304L islower than 5ppm even at 300°C*. As the hydrogen solubility of austenitic stainless steel is almost the same fordifferent kinds of steel, it is believed that the solubility is the same in the case of SUS316L as well.
Hydrogen pressure (√P (√atm))
・Hydrogen embrittlement does not occur** even if SUS316L absorbs several hundred ppm at room temperature or
Fig. 1 Correlation between hydrogen solubility ofseveral dozen ppm at 85°C.
・The solubility of nickel-based alloys differs depending on the composition, but at 400°C the volume of Ni is 85%, SUS304L and hydrogen pressure*
SSRT in hydrogen
which is about 5 times that of Fe.
gas at room
Hydrogen absorption
temperature→ SSRT in
atmosphere
It has been reported*** that hydrogen solubility of Fe does not change at all even in the case of SUS. Based on this, it
SSRT with cathodic charge
is estimated that the hydrogen solubility at 300°C when the volume of Ni is 85% is 30ppm or less.
SSRT inAt this concentration hydrogen embrittlement is very unlikely to occur.
atmosphereAs hydrogen solubility tends to decrease as temperature reduces, it is believed that hydrogen embrittlement does
not occur even at room temperature.

*Nomura et al, Bulletin of the Japan Institute of Metals and Materials, Volume 15, Issue 9, p.563 - 570 (1976)**Omura, Nakamura, Materials and Environment, 60, 241-247 (2011)*** Yoshida et al, Bulletin of the Japan Institute of Metals and Materials, Volume 11, Issue 7, p.533 - 548 (1972)

Contraction ratio in hydrogenand in atmosphere (%)◼

Table 1

Hydrogen solubility (ppm)◼
Investigation of the impact that each degradation factor has on the filter〇 Hydrogen embrittlement◼
Evaluation of the partial pressure of hydrogen in the container
The pressure inside and outside the container is assumed to be the same.
The hydrogen pressure is 0.05Pa since the hydrogen concentration in air is 0.5ppm.

Surface hydrogen concentration (ppm)Fig. 2 Impact of surface concentration of hydrogen on hydrogen embrittlement characteristics **©International Research Institute for Nuclear Decommissioning

97

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters◼ Ti
Based on Table 1, hydrogen solubility of Ti is 10ppm at room
temperature, and that of Ti for industrial use (containing Fe:0.3%) is 100 mass ppm. Considering that hydrogen
embrittlement occurs at a concentration of 100 - 300 ppm,
hydrogen embrittlement is likely to occur at room temperaturedepending on the type of Ti.*
* Fujii et al, Materials and Environment, 60, 248 253 (2011)

Temperature / °C

Estimation of failure mode

α + Residual β

α + γ (hydride) + Residual β

Table 1 Amount of hydrogen content in puretitanium ASTM Gr. 3 for industrial use whichcontains 0.3% Fe, and its phase structure*◼ In the case of ceramics
Alumina which is a ceramic is capable of reducing the amount of hydrogen
permeation. Hence it is being considered for use as a film for suppressinghydrogen embrittlement. An example of a report on a film with hydrogenbarrier function is shown in Table 1**.
It has been reported that alumina does not get embrittled in a hydrogenenvironment***
There were no reports on hydrogen embrittlement of cordierite and zeolite.

Table 1 An example of a report on a film withhydrogen barrier function
Type ofmembraneRate of reduction of amount ofhydrogen permeation** NEDO
Fundamental Research Project on Advanced Hydrogen Science Interim Evaluation Report (2008)*** JST / Strategic Funds for the Promotion of Science and Technology, “Research on the function of hydrogen in
environmental embrittlement of structural material” , 2001 – 2002: Term 2©International Research Institute for Nuclear Decommissioning

98

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode

◼ Glass fiber
There were no reports on the degradation behavior of glass fiber due to hydrogen. It has been reported thatcrystalline quartz glass fiber does not get embrittled in a hydrogen environment** JST / Strategic Funds for the Promotion of Science and Technology, “Research on the function of hydrogen in environmental embrittlementof structural material”, 2001 - 2002: Term 2.

Fluid◼ Polymer
The High Pressure Gas Safety Institute of Japan has reported thatpolyethylene can be used for hydrogen gas pipes.**
There are no reports on polypropylene.
PTFE is generally recognized for use in hydrogen gas environment.***** Tabata, Hydrogen Energy System, Vol 35, No. 4, (2010)*** Source: Website of the Chunichi Giken Products Co., Ltd.

Di-isobutylene
Diethylphthalate
Carbon tetrachloride
Dioctyl phthalate
Cyclohexanone
Cyclohexane
Dimethylaniline
Dimethyl phthalate
Potassium dichromate
Oxalic acid
Bromine water
Nitric acid 10%
Nitric acid 70%
Ferrous nitrate
Ferric nitrate
Potassium nitrate
Calcium nitrate
Silver nitrate
Sodium nitrate
Zinc acetate
Zinc chloride
VinegarMercury
Magnesium hydroxide
Hydrogen gas
Steam

PTFE
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●
●

CS
✕
●
●
✕
✕
✕
✕
✕
✕
●
△
●
●
✕
✕
●
●
●
●
△

Joint material
SUS
SUS
304
316
●
●
●
●
△
△
●
●
●
●
●
●
●
●
△
●
✕
✕
△
△
△
△
●
●
●
●
△
●
●
●
●
●
△
△
●
●
△
●
△
●
●
●
●
●
●
●
●
●
△
-

Brass
●
●
△
●
ー
●
●
●
✕
✕
✕
✕
●
△
△
●
✕
✕
✕
●
△
△

Penetration

C
A●: Good △: Can be used for a limited age of use ×: Not recommended ー: Test data not available

Fig. 1 Usability of PTFE in various environments**©International Research Institute for Nuclear Decommissioning99

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter
 Radiation deterioration
Radiation generated from waste⚫ The nuclides selected during the hydrogen generation evaluation are considered as radiation emitting nuclides.
・Alpha rays: Energy is high at 5 to 7MeV, but its impact can be ignored as its range is short (several cm in air).
・Beta rays and gamma rays: Its range is long, and hence it is necessary to consider that radiation is received even from waste that is at a
distance from the filter.
Radiation impact on various materials(1) Metal, ceramic material (stainless steel, nickel-based alloy, alumina, etc.)⚫ Stainless steel and nickel-based alloys are extensively used inside the reactor. Hence it is believed that there are no issues.
⚫ Upon conducting a three-point bending test after subjecting alumina to electron beam irradiation of 300MGy, a slight change in strength
was observed.
The change was about 15% at most (Fig. 1), and hence it is believed that this issue can be avoided by selecting material with some extra
margin [1].
(2) Glass fiber⚫ When glass is subjected to radiation exposure, its mechanical strength changes.
Depending on the elements that glass contains, the mechanical strength either
decreases or increases [2].

However, as glass material is not ductile, it is believed that degradation due to radiation
exposure is not an issue.
Fig. 1 Results of the test on irradiated alumina material [1][1] Michikazu Kincho and others “ Irradiation Effects on Alumina Ceramics for Vacuum Duct” 41st Session of the Vacuum Society of Japan Proceedings,
Vacuum,P131. Vol. 44, No. 3, 2001
[2] Norihide Takeyama “The Integrated Design Approach of the Lens-Based Space Optical Systems” Optics, Vol 40, Issue 7, P. 331 (2011)©International Research Institute for Nuclear Decommissioning100

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter
 Radiation deterioration

γ rays (PTFE)
Soft X rays (PTFE)γ rays (PE)

Dose (k Gy)
Dose (k Gy)

Fig. 1 Results of material testing of the irradiated polyethylene [5]X rays (PE)Maximum length (relative value)

Breaking
elongation (%)

UnirradiatedMaximum tensile strength (relative value)

Radiation impact on various materials(3) Organic material (Polyethylene, PTFE, etc.)⚫ When organic material is subject to radiation exposure, its mechanical strength changes [3].
⚫ When the HIC lining is simulated, polyethylene is exposed to gamma rays at 60°C, and the environment is changed (air, water, ironhydroxide, calcium carbonate, magnesium carbonate), breaking elongation decreases [4]⚫ In the case of PTFE, radiation resistance is even lower than polyethylene, and mechanical strength reduces with a lower radiation dose [5].
Summary
Since the mechanical characteristics of organic material change due to radiation, the usage conditions need to be carefully studied.

X rays (PTFE)
Soft X rays (PTFE)γ rays (PE)X rays (PE)

Dose (k Gy)

Fig. 13 Changes in the maximum tensile strength (figure on the left) and the maximum length (figure on theright) of PE and PTFE when exposed to various types of radiation

Fig. 2 Results of material testing of the irradiated polyethylene PTFE [5][3] Seiichi Saiki and others “Special Issue New Radiation Chemistry (Basic edition) 9 Effects of ionizing radiation for polymers” RADIOISOTOPES, 66, 469–477 (2017)[4] Results of tests related to radiation deterioration of polyethylene National Institutes for Quantum and Radiological Science and Technology (February 19, 2018)(https://www2.nsr.go.jp/data/000223970.pdf)[5] Masayuki Sekiguchi and others, “Evaluation of the deterioration of polymer material due to X-ray irradiation and the absorbed dose” Tokyo Metropolitan Industrial
Technology Research Institute Research Report No. 7 (2004)©International Research Institute for Nuclear Decommissioning101

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter Impact
Events that have an impact➢ During transportation, the likelihood of the container or filter getting distorted / damaged due to the impact of a collision/fall, etc.
in the event of an earthquake, and the filter losing its functionality, are studied.
Waste storage container
Specifications are assumed considering only the metal part
The weight of the storage container is calculated based on the above-mentioned specifications. The study is conducted assuming 4
ton and 7 ton.

Table 1

Material, external dimensions, etc. of
the waste storage container (assumed)
For dismantling waste

Material

Vent
Vent

For secondary waste
generated from contaminatedwater treatment

Carbon steel (density
7.8g/cm3)

Carbon steel (density
7.8g/cm3)

External
dimensions

1.42m×1.42m×1.0m[1]2.0m×2.0m×1.5m[1]Thickness

0.06m

0.05m

Vent partΦ100mm×1

Φ100mm×2

Approx.
1.5m

Approx. 2mApprox.
2m

For dismantlingwaste*1

For secondary waste
generated fromcontaminated water
treatment*2
Fig. 1 Waste storage container (assumed)*1: Subsidy Project of Decommissioning and Contaminated Water Management in the FY2016 Supplementary Budget “Upgrading offundamental technology for retrieval of fuel debris and internal structures” Final Report*2: Storage container is assumed based on information related to secondary waste produced during contaminated water treatment©International Research Institute for Nuclear Decommissioning102

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities:
Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter Impact
作業
１．廃棄物を内容器に詰め込み・セル内移送

3. Placing the inner container in the storage
３．内容器を保管容器へ収納・蓋締め
containerand closing the lid

２．各種計測Site with high surface
radiation / contamination:

Waste container
handling cellWork that is believed to have lead to an impact✓Fall after the storage container is lifted up(including earthquakes) 作業フロー✓ Lateral collision of the storage container due toswinging while the storage container is beingtransported or shaking during an earthquake

Surface dose
measurement

Site with medium
surface radiation /contamination:

Retrieval cell

Storage container
lid closing
equipment

Storage container
handling cell
(Maintenance cell)

Lid

Measuringequipment

Filter vent
Storagecontainer
Trolley

Storage

StorageTransport cask

container・各種計測（詳細検討中）
withcontainer・内容器（セル内移送手段）
filter vent ・内容器に収納した状態で測定を行う
・遮蔽は無し（内容器は常に遠隔操作で取り扱う）
transportation
・測定の候補・内容器は、取り出し装置との取り合い、複数遮蔽厚の保管容器
equipment
・重量への収納高率等を踏まえて合理的な形状を検討
・表面線量率 waste[1]Dismantling・概算必要個数（上取り出しルートＢ工法1号機あたり）①内容器□500 mm×300 mmの場合→10016個②内容器□335 mm×272 mmの場合→27216個③内容器□450 mm×272 mmの場合→14612個④内容器□535 mm×272 mmの場合→10180個

・推定の候補・放射線量（γ）
・放射能量（α・β）
・核物質量
・発熱量
・水素発生速度・保管容器取り扱いセルの汚染防止
Secondary waste generated from・取り出しセルと保管容器取り扱いセル間のダブルドアシステム相当
品により、保管容器取り扱いセルの汚染を抑制
contaminated water treatment[2]

・保管
・保
品

・保管容器の仕様
・遮へい機能を有する（内容物に応じて複数の遮蔽厚：0, 60, 140, 180mm）・フィルタベント機能を有する
・遠隔操作により移送容器へ収納可能
・遠隔操作により蓋締め可能
・保管時の腐食を考慮する
・保管庫内での無人フォークリフトによる運搬を考慮し、容器収納後の重量が現行の無人フォークリフトの制限重量を超過しないように寸法を設定する。（暫定外径：1.4 m×1.4 m×1.0 m）
・一部大型撤去物を保管容器サイズまで切断し、保管容器に直接収納する可能性がある。
・保管容器取り扱いセル内の廃棄物容器の移動は、クレーン、台車、またはベルトコンベアを用いる
・必要に応じて保管容器の各種測定を行う
例）表面線量測定、表面汚染
・サブプロセス例）蓋締め不具合が生じた場合→（3.a蓋締め直し）
例）表面汚染が保管庫の受け入れ線量を超過した場合→（3.b除染）

・移送
・遮
・閉
・遠
・遠
・移
・水
・繰

Fig. 1 Example of the work undertaken during the flow of collection, transfer and storageof waste (assumed)

・ベルトコンベア（セル内移送手段の例）

・スループット向上・遠隔操作の簡素化の為、取り出しセル内のImpact on the filter
・サンプリングの実施（詳細検討中）
移動にはベルトコンベア等を用いる
仕様
・必要に応じて廃棄物のサンプリングを行い、分析施設へ移送する✓When the storage container (4 ton, ・取り出し時にカメラにて寸法・外観、表面状態、材質の確認を行う。
7 ton) falls from a height of several m, the filter is likely to get distorted/damaged and lose its
・サブプロセス
functionality due to the impact of the・一部大型撤去物に関しては、内容器に収納されずPCV内から払い出
fall or due to distortion of the lid 例）測定値が基準値を上回った場合
of the storage container. Further, the condition of the distortion /→2.a 保管容器あたりの内容器収納数変更
される可能性がある。
damage differs depending on the height
from
which
the
fall
occurred,thespeed of movement, the portion that comes in contact with
→2.b
撤去物の詰め直し
・一部大型撤去物に関しては、内容器サイズまで切断され、内容器に収納される可能性がある。
the ground (storage container falls with its bottom facing down, gets overturned, etc.), the way in which the filter is fixed, etc.

✓ If there is a collision, the filter may get distorted / damaged and is likely to lose its functionality. Further, the condition of the distortion /
装置・設備
装置・設備
装置・設備
damage differs depending on the thespeed of movement
when the collisionoccurs, theportion that collides,
the way in which
the filter
安全機能
安全機能
安全機能
取出しセル / 保守セル
保守セル
保守セル / 搬出入セル
内容器
内容器
保管容器
is fixed, etc.
・閉じ込め機能を確保する。
・閉じ込め機能を確保する
・閉じ込め機能を確保する
汚染拡大
・フィルタベントによる微粒子の✓ It needs to be considered that
dependingon thespecificationsof the閉じ込め
inner container,
the ・HEPAフィルタ等で排気をspecifications閉じ込めof theplace
where thefilter is
閉じ込め
・HEPAフィルタ等で排気を・HEPAフィルタ等で排気を/内部被ばく
閉じ込め
管理
管理
管理
installed (shape of the installation location, whether or not a baffle plate is installed, its specifications, etc.), the waste inside thestorage container could move up and down due to which the filter could get distorted / damaged or blocked.
重大

外部被ばく

遮へい

-

・遮へい厚により遮へい機能を遮へい-

・遮へい厚により遮へい機能を遮へい

・内容物に応じて複数の

・搬出
ベル

・一部
る。

安全機

閉じ込

-

遮へ

有する
有する
事象 on the “Subsidy Project of Decommissioning and Contaminated
遮蔽厚[1] Quoted from FY2020 Final Report (March 2021)
Water Management in the FY2018 Supplementary
Budget(R&D for Treatment and Disposal of Solid Wastes)”, and edited.
[2] Quoted from FY2020 Research Report (Final Report) on the “Subsidy Project of Decommissioning and Contaminated Water Management in the FY2018 Supplementary Budget (R&D
for Treatment and Disposal of Solid Waste
・フィルタベント機能を
・換気する
・換気する
・換気する
火災
水素対策
・上部は常に解放
水素対策
・上部は常に解放
水素対策
水素対
and edited.
有する©International Research Institute for Nuclear Decommissioning103

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on
the filter〇 Dust formation due to vibrations•

During transportation, due to vibrations in the event of anearthquake, or if the container falls, fine particles scatter andadhere to the surface of the filter. The filter gets blocked andits filtration functionality is lost.

(Method of evaluating dust formation due to vibrations)

Highest reaching point

Buoyance Fb

The highest reaching point of the fine particles was evaluatedusing a physical model in which it is assumed that the fine Fluid resistance Rfparticles gain upward initial velocity due to the vibrations,leading to an upward motion while being subject to fluid
Initial velocity
resistance and buoyance in the gravitational field. (Figure 1)
resulting from

Gravity

vibrations v0(Initial velocity resulting from vibrations during transportation orin the event of an earthquake)
Seismic ground motion: Vibration acceleration during the Great
East Japan Earthquake 2933 (gal=cm/s2)[1]Fine
powder
Surface of waste

Fig. 2 Vibration acceleration during the Great East
Japan Earthquake [1](1.5×0.6)0.5/0.6 =1.5(m/s)

Fig. 1 Physical model

Period of vibration 0.3(s) ⇒ Initial velocity: approx. 4.5m/s
Vibrations during transportation [2]：Vibration acceleration during transportation 1.5 (m/s2)
Period of vibration 0.6 to 20(Hz) ⇒ Initial velocity： 0.27 to
1.6m/s⇒ Conservatively considered to be 1.6m/s
Fig. 3 Sweep acceleration during vibration testing at
the time of transportation [2][1] Quoted from the National Research Institute for Earth Science and Disaster Prevention, K-NET & KiK-NET, Earthquake Alert, Maximum acceleration distribution map[2]: Kohei Ogawa, “Series <New Chemical Engineering> 1, Analysis of the movement of fluids” First Edition, Asakura Publishing Co., Ltd., (FY2011), p51©International Research Institute for Nuclear Decommissioning104

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on the filter〇 Dust formation due to vibrations
・ The more the density and larger the particle size the higher is the reach. A maximum distance of 9.23cm is reached due to vibrations
during transportation and a maximum distance of 46.7cm is reached due to seismic motion.
・Based on the distance between the waste and filter, the seismic motion is likely to have an impact on clogging of the filter.
Seismic motion(4.4m/s)

Reach [cm]Reach [cm]Vibrations duringtransportation (1.5m/s)

Fig. 1
Correlation between
particle size and distance thatthe particle reaches in the eventof an earthquake

Particle size [µm]Particle size [µm]Storage container for
dismantling waste(assumed) *1

Distance between waste and filter is assumed to be 1.00 – 0.06 = 0.4m

Vent

Distance between the waste and filter is assumed to be 1.5 x 1/6 = 0.25m
Gap 1m3

Storage container forsecondary waste
generated fromcontaminated water
treatment (assumed)*2

Vent

Approx.
1.5m

7 pieces*1: Subsidy Project of Decommissioning and Contaminated Water Management in the
FY2016 Supplementary Budgets
Upgrading of fundamental technology for retrieval of fuel debris and internal
structures Final Report*2: Figure drawn using information presented by Tokyo Electric Power Company
Holdings Inc.

Water approx. 4t
Filter(dehydration
product) 2.65t(Converted from
Bq to W)

Approx.
2m

Inner container

Approx. 6m3
in its entirety

Approx. 2mFig. 2Estimation of the distance between waste and the filter©International Research Institute for Nuclear Decommissioning105

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on the filter〇 Wall thinning(1) Metal filter
Since crevice corrosion takes place at a potential lower than pitting corrosion, wall thinning of the metal filter was evaluated based on crevice corrosion.
According to Figure 1, under atmospheric environment the chloride ion concentration that gives rise to crevices in SUS316L is 500ppm or more.
As nickel-based alloys and Ti alloys are highly corrosion resistant, the concentration at which crevice corrosion occurs in these is presumed to be higher, but
its occurrence cannot be denied if it is assumed that particles originating from waste or corrosive ions, water adhere to them under storage environment.

Under alkaline environment, if stainless steel becomes a strong alkali as shown in Figure 3, its corrosion resistance decreases. Nickel-based alloys and Ti
alloys are highly corrosion resistant to alkalies, but their specific amount of corrosion cannot be estimated.

April 2013

Hiroshi Sato et al, Piping
engineering, 24, 66 (1982)

Proceedings of the Japan Concrete Institute,Vol.30, No.1 (2008)
Chloride ion concentration threshold value
for occurrence of corrosion

Crevice corrosion potential (V (Saturated
calomel electrode reference))

CRIEPI Report Q12001

316 stainless steel

No crevice corrosion
Titanium

Corrosion potential
under open
atmosphere [2]Corrosion potential
under open
atmosphere (SUS403)

Occurrence of
crevice corrosion

Plain rebars

Chloride ion concentration
Temperature

Fig. 1 Correlation between crevice
corrosion potential and chloride ion
concentration (50°C)

Fig. 2 Crevice corrosion occurrence threshold potential
for Ti and various types of stainless steel in a saturated
aqueous sodium chloride solution environment

Table 1 Corrosion resistance of ceramics with respect
to acids and bases 7

amphoteric, More basic than Al2O3

MgAl2O4 More basic than Al2O3
AmphotericWeakly acidic

Large ← Resistanceto bases

Weakly basic
BeO

Acid resistance →Large(2) In the case of ceramics
Alumina (AL203) ・・ Corrosion resistant in acid solution but has poor corrosion resistance in an
alkaline solution.
Higher the extent of purity better is the corrosion resistanceCordierite （2MgO・2Al2O3・5SiO2) ・・ Has poorer corrosion resistance than alumina, but has beenreported to have stronger corrosion resistance to alkaliies.
Zeolite (SiO4 and AlO4 basically with a tetrahedral structure) ・・ There are no reports on corrosion
resistance.
* Yoshio et al, Materials and Environment, 44, 405 - 415 (1995)

Fig. 3 Correlation between salinity threshold and pH

Fig. 4 Corrosion resistance of ceramics
with respect to acids and bases＊©International Research Institute for Nuclear Decommissioning106

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode(3) In the case of glass fiber
E (for general use), C (acid resistant), S (high strength), D (low dielectric constant) are
the types of glass fiber filters, but only C is manufactured considering its corrosion
resistance.
Glass fiber is not very alkali resistant. This is believed to be because the Si-O bonds arebroken.
Moreover, there are no reports on corrosion in neutral diluted sea-water, under water,
etc.
(4) Polymer
Polyethylene, polypropylene, PTFE indicate good corrosion
resistance to saline water and alkali solution.
Type of plastic material

Vinylidene
chloride

Polyvinyl
alcohol

Polyvinyl
acetatePolystyrene

ABS

Polyethylene

Polypropylene

Polyisobutylene

Polyamide(Nylon)Polyacetal(Delrin /Duracon)Methacryl(Acryl)Polycarbonate

Fluorine(Teflon)

Cellulose
acetateFig. 1 Alkali resistance of glass fiber

Soft

Hard
PVC

PVdC

PVA

PVAc

PS

ABS

PE

PP

PIB

PA

PCM

PMMA

PC

PTFE

CA

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

Calcium chloride

◎

◎

◎

◎

◎

◎

◎

◎

Marcurio chloride

〇

◎

◎

◎

◎

◎

Stannic chloride

〇

◎

◎

◎

Ferric chloride

◎

◎

◎

◎

◎

Copper chloride

〇

◎

◎

◎

◎

Potassium chloride

Na OH 60°C treatment

Treatment time (time)
Vinyl chloride resin

PVC

Oil, solvent, chemicals (according to the order of Japanesesyllabary) [mass concentration % and temperature °C]

Tensile strength retention rate (%)■ Investigation of the impact that each degradation factor has on the filter〇 Wall thinning

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

〇

◎

◎

x

〇

◎

〇

◎

〇

◎

〇

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

Thionyl chloride

x

x

x

x

x

x

x

Nickel chloride

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

Barium chloride

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

Benzyl chloride

x

△

x

x

Magnesium chloride

◎

◎

◎

◎

◎
x

x
◎

◎

◎

◎

x

x

◎

◎

◎

◎
◎

◎

◎

◎
x

Methyl chloride

x

x

x

x

x

x

△

x

x

x

x

◎

Hydrochloric acid [10･RT]

〇

◎

◎

◎

◎

◎

◎

◎

〇

〇

◎

◎

Hydrochloric acid [20･RT]

〇

◎

◎

◎

〇

◎

◎

◎

△

〇

〇

◎

Hydrochloric acid [20･SO]

x

△

△

△

x

x

△

△

〇

△

△

△

◎

Hydrochloric acid [38･RT]

△

◎

◎

◎

△

◎

◎

◎

x

〇

△

◎

Salt water

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

◎

Chloride gas (dry)

△

〇

〇

◎

x

△

△

x

x

x

〇

x

Chloride gas (wet)

△

〇

〇

x

△

△

x

x

x

△

x

〇

Chlorinated solvents

x

x

x

x

x

x

x

x

x

x

x

x

x

x

◎〇

x

Aqua regia

x

△

x

x

x

x

x

x

△

x

x

x

x

x

◎

x

Octyl alcohol

x

〇

〇

〇

〇

◎

〇

△

△

△

〇

◎

Ozone

〇

〇

〇

△

△

〇

x

△

Olive oil

△

〇

〇

◎

〇

〇

x

◎

〇

◎

◎

◎

x

x

x

△

◎

◎

△

〇

x

〇×

◎

〇

〇

〇

〇

△

〇

〇

◎

Hydrogen peroxide [5･RT]

〇

◎

◎

△

◎

◎

◎

〇

◎

Hydrogen peroxide [5･50]

〇

◎

◎

△

〇

◎

△

〇

◎

Hydrogen peroxide [30･RT]

〇

◎

◎

x

〇

◎

〇

△

◎

Sodium peroxide

〇

〇

〇

◎

◎

◎

x

x

x

◎

Sodium peroxide [10･RT]

〇

◎

◎

◎

◎

◎

◎

◎

◎

◎

〇

△

◎

Sodium peroxide [30･RT]

〇

◎

◎

◎

〇

◎

◎

◎

◎

◎

△

△

◎

Sodium peroxide [30･70]

x

△

△

△

△

〇△

〇

◎

◎

x

△

x

◎

◎

〇

◎

Perchloric acid

◎

◎

x
◎

〇

Oleic acid◎

Kasuga, Journal of the Japan Research Association
for Textile end-uses, Vol. 6, No 4 (1965）

Table 1 Corrosion resistance of polyethylene,
polypropylene, PTFE is good and is indicated by ◎Excerpts from Kayo Corporation
http://www.kayo-corp.co.jp/©International Research Institute for Nuclear Decommissioning107

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on the filter〇 Salt deposition

Storagecontainer

FilterDispersion of
micro droplets

Hydrogen bubbles

・ An event wherein water contained in waste gushes towards the outside of the container, Water
it adheres to the filter and after drying salts get deposited on the filter thereby blocking it,
containing salts
due to which its filtration functionality is lost.
⇒ Water containing salts coming in contact with the filter is a necessary condition(Dispersibility of droplets of water containing salts)
・Droplets get dispersed due to hydrogen generation and due to seismic motion.

Waste

Maximum watercontent 50wt%[1]Fig. 1 Conceptual diagram ofthe dispersion of droplets [1]① Dispersion due to hydrogen generation
・ Hydrogen bubbles in the water contained in waste rise and reach the water surface.
When the bubbles burst, micro droplets are formed.
・Micro droplets are generally called film drop, and their size is about 5-10μm[2]・ As the density of water is 1g/cm 3, based on the results of evaluating the floating of fineparticles due to hydrogen generation, even if micro droplets are formed they do notreach the filter.

No. of pieces

Surface area

Weight② Dispersion due to seismic motion
・ Water containing salts is highly likely to come in contact with the filter due to vibrations
caused by seismic motion.
・ The size of salt particles that are formed when water containing salts dries is assumed
to be 200-800μm[3]・ HEPA filter is mainly made up of around 1μm glass fibers, and its porosity is about
90%[4]. Hence the size of the pores is believed to be roughly about 1μm／(1-0.9)=10μm

Particle size

Fig. 2 Grain size of the crystals formed
when saturated salt water vaporizes [3]⇒ Water containing salts is likely to come in contact with the filter due toseismic motion, and the salt formed when this water dries is likely toblock the filter.
[1]: The conceptual diagram has been drawn assuming the storage container based on information related to secondary waste
produced during contaminated water treatment[2] Satoshi Kadowaki, “Sea-salt Particles in the Atmosphere”, Aerosol research, Volume 6, Issue 2 (1991)[3] Akira Yokochi, “Recent Tendency of High Efficiency Particulate Air (HEPA) filters”, Health Physics, 19, 249258 (1984)[4] Li Bao “Influence of Fiber Size Distribution on the Collection. Performance of HEPA Filter”, Aerosol research, 12(1), 38 - 44 (1997)

Fig. 3 Example of the structure of
HEPA filter[4]©International Research Institute for Nuclear Decommissioning108

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
■ Investigation of the impact that each degradation factor has on the filter〇 Electric corrosion
The filter gets wet due to condensation or due to the water in the waste,

Threaded portion

Taper threaded portion

Filter housing

and chloride ions adhere to it as a result of which electric corrosion occurs.
The formed corrosion products block the filter.
● Following are hypothetical conditions
The storage container is made of carbon steel
Structure of filter is according to Figure 1
The housing of the filter is made of stainless steel or nickel-based
alloys● Since ceramics, glass and polymers are insulators, electric corrosion
does not occur.
Hence only metal filters were evaluated.

Filter lid

FilterRetainer ring
Filter surface

Fig. 1 Example of the structure of the filter● If the filter is made of SUS, nickel-based alloys or Ti alloys, there is almost no potential and hence electric corrosion does not occur.
Table 1 Material of the contact area and occurrence of electric corrosion

Positive

Corrosion potential

Negative

Platinum

Storagecontainermaterial

Carbonsteel

Filterhousing

Made of SUS

Filter material

Occurrence ofelectric corrosionof the filter

SUS316L

AbsentNickel-basedalloy

AbsentTi alloy

AbsentTitanium

Silver brazing
Lead
BronzeAluminum brass
Soldering
CopperTin
Brass
Low carbon steel
Mild steel, cast iron
Aluminum. Al alloy
Zinc

Fig. 2 Potential difference in metals under sea water (temperature 25°C)** Urban Steel Research Group “Steel frame landscape material guide” July 1997
©International Research Institute for Nuclear Decommissioning109

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter Water film
Formation of water film
⚫ A water film is likely to be formed on the filter surface due to condensation. Let us assume that the temperature inside thecontainer falls from 30°C to 10°C.
Assuming that the amount of saturated water vapor: 30.4g/m 3＠30°C, 9.4g/m3＠10°C, capacity of the waste container isapprox. 1.5m3, area of the lid is approx. 1.69m2, the amount of condensed water is 31.5g. If the condensed water adheres
only to the surface of the lid, the amount of water adhered to the surface of the filter is 31.5/1.69＝18.6g/m2.
If the specific gravity of water is assumed to be 1g/cm 3, the water film will be 18.6μm which is extremely thin, but if it piles
up, the filter element will get blocked by water.
⚫ As the formation of water film due to condensation does not depend on the material of the filter, this will be discussed for all
types of filter material.
If the filter gets blocked due to moisture⚫ Condensation can be reduced by appropriately controlling the temperature and humidity, but blockage for a short timecannot be denied.
In that case, whether or not the increase in internal pressure due to hydrogen generation can push out the moisture frominside the filter element becomes the performance indicator.
⚫ After the filter is blocked due to condensation when the pressure inside the container is 1 atm and hydrogen is 0vol%, when
the hydrogen concentration becomes 4 vol% due to hydrogen generation, the overall pressure becomes approx. 1.04 atm.
Hence, it is believed that a filter that has gotten wet can be used as long as the pressure is lower than this and air starts
flowing. Meanwhile, if air cannot flow, hydrogen is likely to get accumulated inside the container.
Summary
When the filter element is covered with a water film due to condensation, whether or not ventilation can be brought about before
the hydrogen concentration reaches the lower concentration limit for combustion, needs to be verified by means of a test.

©International Research Institute for Nuclear Decommissioning

110

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode
◼ Investigation of the impact that each degradation factor has on the filter
 Freezing
Occurrence of freezing⚫ When a water film is formed on the surface of the filter due to condensation and the air conditioning equipment fails, thefilter gets exposed to below freezing conditions resulting in freezing. If the entire filter is covered with ice, the filter element
gets blocked. Also, in the case of organic material, the mechanical characteristics are likely to change due to exposure to
low temperature.
⚫ As blockage due to freezing does not depend on the material of the filter, this will be discussed for all types of filter material.
If the filter gets frozen⚫ Under appropriately controlled temperature and humidity, condensation is very unlikely, and as long as the temperature is
0°C or more freezing does not occur.
However, if the filter is completely blocked due to freezing, hydrogen is likely to get accumulated.
⚫ In that case, it can be evaluated whether or not the hydrogen concentration will become 4% based on the hydrogengeneration rate and the time for which freezing continues.
⚫ If organic material will be used, it is necessary to investigate its mechanical strength at low temperatures.
Summary
It needs to be verified by means of tests whether or not the prescribed mechanical strength is met if the filter element getsfrozen.
Whether or not the ice melts and ventilation can be brought about before the hydrogen concentration reaches the lower
concentration limit for combustion, does not need to be verified through tests, but can be evaluated based on the hydrogengeneration rate and the time for which freezing continues.

©International Research Institute for Nuclear Decommissioning111

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode

(Conditions in which the impact becomes apparent)
・ When fine particles disperse from the surface of waste due to the current
generated as a result of ventilation, there is an impact.
・ As the impact of ventilation does not depend on the filter material,
this will be discussed for all types of filter material.

Distance from the surface■ Investigation of the impact that each degradation factor has on the filter〇 Dispersion of fine particles
・ Due to the impact of the ventilation system of the storage facility, a current gets generated inside the storage container.
As fine particles scatter and adhere to the surface of the filter, the filter gets blocked and the filtration functionality is lost.

Gas stream

Particle bed

Solid surface

Fig. 1 Dispersion of fine particles from the solid(Ventilation conditions)surface Conceptual diagram[1]・ Passages through which gas streams flow are not present inside the storage container. The differential pressure insideand outside the container gives rise to gas streams inside the container.
・ However, the ventilation system inside the building is supposed to perform air intake as well as exhaust, and hence there
is no difference in the pressure inside and outside the container.
⇒ It is believed that the flow rate conditions that would lead to dispersion of particles would not arise.

Excluded from the failure mode assuming that the ventilation system does not lead to dispersion of fine particles.
[1] Kuniaki Goto, “A Study on Particle Dispersion Technique and Powder Adhesiveness.”, Pulverization No.59 (2016), p51-59©International Research Institute for Nuclear Decommissioning112

(a) Investigation and evaluation of storage methods [2] Investigation of requirements of storage containers and facilities: Performance evaluation of vent filters

Estimation of failure mode

◼ Investigation of the impact that each degradation factor has on the filter
 Summary⚫ The impact of each degradation event on filter degradation (blockage, damage) was evaluated.
⚫ The following 8 events were selected as degradation events that are likely to have an impact on degradation of the filter
during the process of collecting solid waste.
Note that, the process in which the impact becomes apparent is mentioned in [ ] and the filter material that is affected is
mentioned in ( ).
・ Hydrogen embrittlement
・ Radiation deterioration
・ Impact
・ Dust formation due to vibrations
・ Wall thinning
・ Salt deposition
・ Water film・ Freezing[During storage] (Polymer)[During storage] (Polymer)[During collection / during transportation / during storage] (All materials)[During storage] (All materials)[During storage] (All materials)[During storage] (All materials)[During storage] (All materials)[During storage] (All materials)⚫ Hereinafter, degradation simulation and performance assessment methods will be investigated for the 8 events that havean impact using the performance test method.

©International Research Institute for Nuclear Decommissioning113

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of performance testing methods◼ Methods of testing filter materials
 Hydrogen embrittlement test⚫

The amount of hydrogen generation is calculated based on the duration of storage of solid waste and the rate of hydrogengeneration from waste. Based on that, the maximum, intermediate and minimum amount of hydrogen generation are
selected, and each type of filter material is charged with hydrogen. The filter material charged with hydrogen is mounted on
the filter performance evaluation equipment and the test is carried out. Hydrogen can be charged either by means of theelectrolytic method or the high pressure charging method. The method that is suitable for the material is used.

⚫

The charged filter is checked for the amount of hydrogen charged. This is carried out using the thermal desorption method.
(Refer to Fig. 1)
Thermocouple

Thermocouple

Vacuum chamber
Quadrupole mass
spectrometer
SampleTransparent quartz plateStage for quartz material
Thermocouple(For stage temperature)
Quartz pillar for
introducing infrared rays

Infrared rays

Fig. 1 Example of thermal desorption analyzer©International Research Institute for Nuclear Decommissioning114

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of performance testing methods◼ Methods of testing filter materials
 Radiation deterioration test⚫

The material to be tested is exposed to radiation with the expected radiation dose. The mechanical characteristics before and after
irradiation are investigated.
It is investigated whether or not the above-mentioned characteristics satisfy the stress expected when the material is used as a filter.

⚫

Exposure dose
It is difficult to estimate the exposure dose of the filter as it depends on the amount of radioactivity present in the contents of thecontainer, the distance, moisture content, etc. Hence it is tentatively set at 30mSv/h as shown in Fig. 1 [1]. The cumulative dose over a
period of 50 years is
30 x 10-3 (Sv/h) x 24 (h) x 365 (day) x 50 (year) = 1.3 x 104(Sv)
It is believed that the material should be irradiated up to approx. 130000 Sv to evaluate the correlation between the extent ofdeterioration and the radiation dose.

⚫

Testing the effect
The effect of radiation on the mechanical strength can be evaluated by simulating deterioration during use by exposing the filter toradiation beforehand, and then subjecting it to an impact test.
Filled in the stabilization(dehydration) storage
container

Slurry extraction

Supply tank・Verification of thesurface dose of the
HIC to be treated.

・Recording of the surface dose
and transfer amount after
homogenization

Storage container
・Management of the surface dose of
the dehydration product container(30mSv/h or less)Fig.1 Example of waste container dose rate [1][1] Set using the container surface dose rate mentioned in “Design for slurry stabilization treatment” Special Nuclear Facility Monitoringand Evaluation Committee (88th Session) Document 1-2-1, 2021/1/25, as reference.
©International Research Institute for Nuclear Decommissioning

115

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of performance testing methods◼ Methods of testing filter materials
 Impact test
The filter is subjected to impacts expected to arise during the work undertaken as part of the collection,transfer and storage flow. The integrity of the filter after being subjected to impacts is checked.
<Testing the effect>• The filter is fixed on to the lid part of the structure simulating the storage container (hereinafter, simulated
storage container).
• The simulated storage container is lifted up to the designated height by means of the hoisting accessory.
• The simulated storage container is detached from the hoisting accessory and allowed to fall freely.
• The integrity of the filter is checked after the fall （External appearance, filter performance test, etc.).

Hoisting
accessory
Simulatedstoragecontainer

Vertical fall

Inclined fall

Inclined fall
(90°)

Vertical fall
(Falling of storage
container on astorage container)

Fig. 1 Example of the direction of the fall of the simulated storage container while testing the effect➢ Note that the results are likely to differ depending on the method of attaching the filter to the lid part, the shape anddimensions of the storage container, the method of attaching the lid part to the storage container, the shape of theinner container, etc., and hence it is desirable to conduct the test after the detailed specifications are finalized.
©International Research Institute for Nuclear Decommissioning

116

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of performance testing methods■Methods of testing filter materials〇 Test on dust formation due to vibrations (Test on dust formation due to seismic ground motion)
A vibration test is conducted in which random vibrations simulating actual seismic ground motion are imparted, andfine particles that get airborne as a result are made to adhere to the filter to simulate filter deterioration.
Housing(For attaching the filter)

Filter(Testing method)・ A vibration table where seismic ground motion can be simulated is arranged for.
Simulated fineparticles
・ Simulated fine particles are selected considering the types of waste(density, particle size)
・ The simulated fine particles are placed on the vibration table,
and the filter is arranged at a height equal to the distance expectedVibration tablebetween the waste and the filter inside the storage container.
Fig. 1 Schematic diagram of seismic ground(If the specific structure of the storage container has been finalized,
motion simulation test
a test piece simulating that structure is used.)・ Seismic ground motion is simulated to conduct the vibration test.

・ After imparting vibrations, the amount of matter adhered and the adhesion status is checked based on the weight of
the filter and by means of a spectral analysis, and the performance is evaluated separately by means of aperformance assessment test.
# Note that the results are likely to differ depending on the method of attaching the filter to the lid part, the shape anddimensions of the storage container, the method of attaching the lid part to the storage container, the shape of theinner container, etc., and hence it is desirable to conduct the test after the detailed specifications are finalized.

©International Research Institute for Nuclear Decommissioning

117

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters[Study of performance testing methods]◼ Methods of testing filter materials
 Methods of testing wall thinning⚫

⚫

⚫

Cyclic corrosion test (JIS K 5600_7_9): A method used for evaluating the paint by means of a corrosion test in which thecycle of salt spray, drying and humidification is repeated. This method is used to evaluate wall thinning and blockage ofvarious types of filters.
Using the above-mentioned testing method, various filters are exposed to corrosive environment to develop wall thinningand blockage conditions. The filters in which the extent of deterioration changes are placed on the filter performanceassessment equipment and tested. Thus the extent of deterioration under which performance can be maintained becomes
clear.
Meanwhile, corrosion tests are conducted on various types of filters by simulating expected environmental conditions, andthe occurrence of deterioration is studied through relative comparison with the extent of deterioration under corrosive
environment, and in the expected environment.

The filter is placed inside the equipment illustrated in Fig.
1, combined cycles are conducted and the filter is subject
to corrosion.
The tower in the center serves as the spray tower for
spraying water, saline water, etc. A humidifier is installed at
the bottom, and water in the form of mist blows in throughthe steam pipe. The temperature is adjusted by means of
direct steam heating.

Test piece

Fig. 1 Example of the cyclic corrosion test apparatus©International Research Institute for Nuclear Decommissioning118

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of performance testing methods■Methods of testing filter materials〇 Simulation test on deterioration due to salt deposition
Deterioration of filters due to salt deposition is simulated by impregnating
and drying the filter with saline water.
(Testing method)・ A simulated aqueous saline solution is prepared (Type of salt and its
concentration is simulated).
・ The ratio of the solution to the waste (simulated) is adjusted, and thesimulated waste is prepared in the container.

Sealedcontainer

Filter・ The filter is installed on top of the container.
(The height is adjusted according to the distance between the waste andfilter)・ The lid is closed and a vibration test is conducted by simulating seismic
ground motion.

Simulated waste

Vibration table

・ After imparting vibrations the filter is removed. After drying it, the amountof salt adhered to it and the form of adhesion are evaluated.
Fig. 1 Schematic diagram of the simulation test
・ The performance is evaluated by means of a separate performanceon salt depositionassessment test.
* Note that the results are likely to differ depending on type of filter, amountof waste filled in the container, moisture content, type and concentration ofsalt in the solution, its pH, dimensions and shape of the container, etc., andhence it is desirable to conduct the test after the detailed specifications are
finalized.

©International Research Institute for Nuclear Decommissioning

119

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters[Study of performance testing methods]◼ Methods of testing filter materials
 Water film test
Air permeability test of filter that is blocked due to droplets⚫

If the filter is not completely blocked・ Frazier Air Permeability Test (JIS L 1096, JIS L 1913)［1］The difference in pressure inside and outside the test piece is kept constant, aeration is brought about using a fan and the amount of air(cm3/cm2･s） is measured.
If partial blockage is considered equal to reduction in the area of the filter membrane, the margin in the filter area can be evaluated throughthis test.

⚫

If the filter is completely blocked

Vertical barometer*2

・ Proposed air permeability test of a blocked filter[2]The method of testing water resistance of fibers is used for

Vertical barometer*1

evaluating the pressure at which water is discharged from
Test piece

the filter.

Support plate

Sample

The filter is attached, pressure is applied using water and the pressure
when water is discharged from the filter is measured.
The pressure required for discharging gas when condensed water isformed can be evaluated with this test.
Air hole*1 Indicates the differential pressure
between the atmospheric pressure
above the test piece and ①.
*2 Indicates differential pressure
between ① and ②.

Air intake fan

Fig. 1 Conceptual diagram of
Frazier Type Air Permeability Test [1]Hydraulic pressure

Fig. 2 Conceptual diagram of the airpermeability test of a blocked filter(Created based on [2])[1] “Air permeability test of textiles” Aichi Center for Industry and Science Technology News December 2018(http://www.aichi-inst.jp/other/up_docs/no.201_03.pdf)[2] Document from W. L. GORE and Associates（https://www.gore.co.jp/sites/g/files/ypyipe116/files/2016-04/PTV-TechInfo-WEP-Testing-US.pdf）©International Research Institute for Nuclear Decommissioning120

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters[Study of performance testing methods]◼ Method of testing the effect of freezing of filter materials○ Testing the effect of freezing⚫ The material to be tested is cooled and frozen while it contains water. The mechanical characteristics
when the material is frozen and after it is thawed are investigated.
It is investigated whether or not the above-mentioned characteristics can handle the stress expected whenthe material is used as a filter.
⚫ Freezing conditions
The cooling temperature and cooling period are determined depending on the use conditions of the filter.
⚫ Testing the effect
The effect on mechanical strength of the filter material can be evaluated by simulating deterioration of thefilter material due to freezing and then subjecting it to an impact test.

©International Research Institute for Nuclear Decommissioning121

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of integrity evaluation methods◼

◼

◼

Methods of evaluating the particle collection function• There are mainly 2 types of air filter testing methods namely the weighing method and the number counting method. The number counting method isused for HEPA and ULPA.
• In the number counting method, the minimum collection rate (reference number) and the pressure loss when an air current containing aerosol ispassed at a rated flow specific to the filter, are evaluated.
Testing method[1]• The testing environment is prepared with temperature 25±10℃, humidity 55±15℃, flow rate from the fan motor 0.45m/s±20% at the filter portion• Aerosol is fed for testing purposes on the upstream side of the filter (1 type (DOP) of the 2 particles for test purposes or an equivalent aerosol isused[2])• The concentration of the aerosol for test purposes is adjusted (The particle measurement instrument on the upstream side indicating 0.3μm or morenumber of particles at 5×106 particles/min or more).
• The particle measurement instrument[3] on the downstream side is adjusted (Measured in clean air and let to stand until the reading becomes stable).
• Aerosol for test purposes is fed to the filter, the numbers of particles on the upstream and downstream side are continuously measured at fixed timeintervals, and the collection rate is calculated.
• The static differential pressure on the upstream and downstream side of the filter is measured using a manometer and the pressure loss is calculated.
Determination conditions[4]• Minimum collection rate (Reference number): 99.97% or more for particle size 0.3μm in the case of HEPA, and 99.9995% or more for particle size
0.15μm in the case of ULPA
• Pressure loss: Initial pressure loss 245Pa or less for both HEPA and ULPA
Pressure loss
measurement equipment

Downstream duct

For HEPA and ULPA, if the filter performanceafter the deterioration test satisfies thedetermination conditions, deterioration effect isconsidered to be absent.

Filter for testing
Filter for air intake

Upstream particle concentration
measurement location

Air volume control
damper
Upstream duct

Detection port

[1] JIS K 3801 Test Method of HEPA Filters for Microbiological
BlowerAerosol generator for testing purposes

Light scattering airborne
particle counter
Anemometer

Use[2] JIS Z 8901 Test Powders and Test Particles[3] JIS B 9921 Light Scattering Airborne Particle CounterFig. 1 Performance assessment test Equipment configuration (Example) [1][4] JIS Z 8122 2000 Contamination Control - Terminology©International Research Institute for Nuclear Decommissioning122

(a) Study and evaluation of storage and management methods [2] Study of the requirements for containers and storage
facilities : Performance evaluation of vent filters

Study of integrity evaluation methods◼
•

Method of evaluating the hydrogen permeation function
The method for evaluating the hydrogen permeation function of filters used in nuclear facilities in the US is described in
References [1]. Hydrogen of designated concentration (4%) is sealed inside the container with attached filter and it is let tostand. The hydrogen concentration inside the container is measured 7 times every 5 minutes. The rate of hydrogen
permeation (hydrogen diffusivity D) is obtained using the formula listed below.

𝐷=

𝑃𝑉
𝐻0
𝐿𝑛
𝑡𝑅𝑇
𝐻𝑡
Nitrogen gas

D: Hydrogen permeation rate (Mol/sec/molFraction),
P: Pressure (atm), V: Container capacity (L), t: Time (seconds),
R: Ideal gas constant (0.082Latm/molK),
T: Temperature (K), H0: Initial hydrogen concentration,、
Ht: Hydrogen concentration at time ｔ• If being installed inside the container, the required
number of designated filters are installed depending on
1ml
Sample loop
the hydrogen generation rate inside the container.

Hydrogen 4% +Nitrogen 96%
FilterArgon gas

Container

Purge
valve

Gas chromatographFig. 1 Hydrogen permeation function evaluation test Equipment configuration (Example)[1] Terry Wickland and John Schierloh "HYDROGEN DIFFUSIVITY THROUGH DRUM AND LINER VENTING FILTERS" WM98
(http://archive.wmsym.org/1998/html/sess27/27-37/27-37.htm)©International Research Institute for Nuclear Decommissioning123

(a) Study and evaluation of storage and management methods [2] Study of the requirements forcontainers and storage facilities : Performance evaluation of vent filters◼ Results of implementation
 Organizing of input conditions✓ Target waste was selected from the Waste List and the waste characteristics.
✓ Storage conditions and storage environment were organized.
 Study of deterioration factors✓ Deterioration modes were defined as blockage and damage of the filter.
✓ The causes and events leading to blockage of filters were organized based on the storage environment ofsolid waste, and 18 deterioration events were identified in all.
 Clarification of filter requirements✓ Besides filters used in overseas nuclear facilities, ① manufacturing methods, filter material, ② application,
main manufacturers, ③ fine particles collection function, ④ hydrogen permeation function of the filters forgeneral industrial use (metal, ceramic, glass fiber, polymer) were investigated and the results were organized.
✓ Filters for general industrial use that satisfy the fine particle performance specifics (equivalent to HEPA) of
overseas vent filters can be manufactured. Although the actual value pertaining to the hydrogen permeation
function has not been specified in the catalog, etc., in principle, it is presumed to be equivalent to overseas
vent filters.
 Estimation of failure mode
✓ 11 deterioration factors that are likely to have an impact on filter deterioration were identified from 18
deterioration factors based on input conditions.
✓ Failure mode scenarios were developed for each process for the 11 deterioration factors.
✓ Occurrence of deterioration, extent of progression of deterioration when each type of filter is used were
estimated through literature survey, and 8 affecting events were identified.
 Study of performance testing method✓ The performance testing methods required for evaluating the impact of the 8 failure modes on the life of thefilter were studied and organized.
 Study of integrity evaluation method✓ The filter performance testing methods for evaluating the impact of each failure mode on the integrity of thefilter were studied.

©International Research Institute for Nuclear Decommissioning

124

(a) Study and evaluation of storage and management methods [2] Study of the requirements forcontainers and storage facilities : Performance evaluation of vent filters

Summary
Results so far✓ Characteristics that are common to each type of waste were identified based on the organization of
the types of solid waste, and in addition, based on the expected treatment processes, theenvironments in which the filters will be placed were assumed as input conditions and organized.
✓ The function of letting hydrogen permeate and the function of collecting fine particles were defined asfunctions required of a filter, and environmental deterioration factors that are likely to affect thesefunctions were identified. Also, general filter specifications were investigated, and prospects of being
able to select filter material depending on environmental requirements were seen.
✓ Scenarios likely to impair filter performance were studied based on filter functions and solid waste
treatment processes, respective failure mode scenarios were investigated and organized, and in
addition, proposed testing methods for evaluating the impact that the failure modes have on filter life
were studied.

Issues, etc.
✓ As types of solid waste, treatment methods, storage requirements, environmental criteria, facility
constraints, etc. serve as input information for determining filter specifications, items pertaining towhich conditions become concrete as production level work progresses need to be compared with
the input conditions assumed in this research, and the necessity to reselect the deterioration
simulation test conditions and failure modes needs to be studied once again.

©International Research Institute for Nuclear Decommissioning125

(b) Technological development of contamination evaluation for sorting solid waste◼ Achievements up to FY2020
 α camera was manufactured and a function verification test of the system was conducted.
 A mock-up test of the α contaminated equipment (flange tank piece) was conducted at the site to assess
performance.
It was possible to measure the 220Bq, 740Bq, 2960Bq α ray sources (Am-241)
affixed on the flange tank piece, but the noise generated was from equivalent or
higher concentrations.
◼ Goal The factors due to which the flange tank piece generated noise will be
specified and noise countermeasures will be taken.
 Upon undertaking countermeasures α contamination detection performance
will be evaluated once again.

740Bq
220Bq
2960BqRange up towhich noise
has animpact◼ Details of implementation (See details on next pages)
Fig. 1 FY2020 mock-up test results
 Issues in the items that affect measurement will be consolidated based on site investigation.
 The following will be implemented in response to the consolidated issues.
α camera: Noise countermeasures will be taken for the flange tank piece.
Distance measurement system: The range within which the TOF conditions are satisfied will be indicated basedon the results of site investigation.
Means for access: A conceptual drawing of the crawler on which the α camera is mounted will be provided. (Thescope of this Government R&D project is up to consolidation of issues)◼ Indicators for the achievements of goal Presentation of issues based on the mock-up test results for using the α camera at the site, and corresponding
countermeasures.
(Target at the completion: TRL: 3→4)

©International Research Institute for Nuclear Decommissioning126

(b) Technological development of contamination evaluation for sorting solid waste

Current status of α-contaminated waste management and issues thereof⚫
Current status of α-contaminated waste management
When α-nuclide contamination with surface contamination density of 0.4 Bq/cm2 or more is detected by means ofsmear measurement, the waste is sorted and placed in storage containers. Adopting smear measurement forascertaining the status of contamination in the future when full-fledged decommissioning work will be carried out, is
not practical considering the time required for measurement and from the viewpoint of radiation dose reduction.
⚫

Requirements related to the technology for sorting waste➢ Being able to ascertain surface contamination distribution of target objects from a distance.
➢ Being able to ascertain contamination distribution of targets over a wide area in a short time.

⚫ Purpose of this project
To develop a technology that can comprehensively and in a short time measure the surface of the object (or area)
to be dismantled. Also, to clearly specify the applicable process depending on performance.

α contaminated area

α surface contamination
density distribution on thesurface of equipment, etc.
(metal, etc.)

α surface contamination
density distribution on
the surface of building
concrete
Image of distribution measurement
Surface contamination density(Bq/cm2) distribution

Contamination distribution
measurement

Under the lower detection limit

Fig. 1 Image of distribution measurement using the α camera©International Research Institute for Nuclear Decommissioning127

(b) Technological development of contamination evaluation for sorting solid waste

Principle of an α camera and its advantages
Measurement principle①
Reaction takes place with nitrogen within α radiation range (about
several cm) to produce several 100 photonic ultraviolet rays.
②
These ultraviolet rays are focused and imaged with a lens, andthe distribution of α rays is measured from the distribution of thenumber of photons reaching the light detector.
③
PMT (Photomultiplier Tube) is shielded from γ rays as it is likely tobe affected by Cherenkov light or others when γ rays enter PMT.

Detector(PMT)

Visible-light filter
MirrorShield for γ rays

Lens

Ultraviolet rays

α raysLight emission

Advantages
・Can be measured at a distance from the source of α contamination
・A wide area can be measured at once.

Examples of main presentations made in connection with α camera
1. IEEE2013 NSS、 Remote Detector of Alpha-Ray Using Ultraviolet Ray
Emitted by Nitrogen in Air
2. ICONE-23 REMOTE DETECTION OF ALPHA RADIATION USING UVPHOTONS EMITTED BY NITROGEN
3. Spring meeting of Atomic Energy Society of Japan in FY2013, Technology
for remote measurement of α radiation
4. Autumn meeting of Atomic Energy Society of Japan in FY2014, Technology
for remote measurement of α radiation - Application to lighting environment -Uranium, etc.

Fig. 1 Schematic diagram illustrating the principle

Light emission in theultraviolet range

Light emission spectrum

Fig. 2 Nitrogen light emission spectrum
Cited from J.Sand. Remote Optical Detection of Alfa Radiation. IAEA－CN-184/23.
©International Research Institute for Nuclear Decommissioning128

(b) Technological development of contamination evaluation for sorting solid waste

Development history
Table 1
Items

Development history

FY2018FY2019FY2020

Environmental
condition setting

Environmental condition setting

Designing & Manufacturing

FY2021

FY2022

Identification of factors from measurement principle,
setting of required specifications
Sensitivity improvement, weight reduction of moving
elements, combination with distance measurement system,
verification of basic performance

Testing

α camera alone

Measurement
systemSystematization(With functions of
pan tilt,
raising/lowering and
3D distribution
preparation)

Improvement

Testing

Designing

Testing

・Evaluation of performance as α
camera・Additional verification of thefunction for mapping measurement
results

Improvement andcountermeasures
Measures in
response to theissues found as aresult of the mockup test

Anti-ghosting*0 measures,shielding mechanism
improvement
・Evaluation of sensitivity to material (surface
condition)
・Mock-up tests at 1F site

Manufacturing Performance verification test

Site investigation
Identification of issues
・Identification of the issues through site investigation
・Measures in response to issues (α camera, distancemeasurement system)

Countermeasures*Peak other than the radiation source location©International Research Institute for Nuclear Decommissioning129

(b) Technological development of contamination evaluation for sorting solid waste

This year’s report
The following was implemented in response to the unanticipated noise generation found as aresult of the FY2020 1F mock-up test.
(1) Common items• Obstructive factors concerning the α camera, distance measurement system, and means
for access were identified based on the site photographs.
• Issues in each element were consolidated based on the obstructive factors.
(2) Measures in response to issues with the α camera• Method of implementing measures in response to the site investigation results• The α rays, β rays emission processes were organized quantitatively (This year the focus
was on nitrogen light emission)
A prediction model was developed based on emission tests and simulations, and wasevaluated.
• Countermeasures were comprehensively studied based on the characteristics of noise
emission, and their effectiveness was evaluated.
(3) Measures in response to issues with the distance measurement system• Clear specification of the range in which distance measurement is feasible• Improvement in the distance combination algorithm(4) Means of access• Conceptual drawing assuming that an existing decontamination crawler is used
• Issues until application to actual equipment©International Research Institute for Nuclear Decommissioning

130

(b) Technological development of contamination evaluation for sorting solid waste

Consolidation of issues
Looking back at the FY2020 mock-up test (1)- Details of the test location The test was conducted at the building for large component inspection where the flange tank piece decontamination work
was being carried out.
β γ contamination was spread over a wide area. Hence a flange tank piece with as high a dose rate as possible wasselected from that area for measurement.
•
Measurement target: 1 flange tank piece meeting the test conditions was selected from flange tank pieces prior todecontamination work.
•
Measurement location: Contaminated area (Y zone). Contamination was mainly Sr/Y-90 nuclides emitting β rays.

External appearance of large
component decontamination equipment

Fig. 1 External appearance of the location of the large components inspection building
Cited from TEPCO HD “Status of installation of large components decontaminationequipment April 16, 2018”

Fig. 2 Flange tank piece
Photographed on October 30, 2020

©International Research Institute for Nuclear Decommissioning131

(b) Technological development of contamination evaluation for sorting solid waste

Consolidation of issues Looking back at the FY2020 mock-up test (2)Table 1 Test items by evaluation items
Test details

Test method

Results

Evaluation of correlation
between β γ surface
contamination density andnoise (Test ①)

β γ surface contamination density (2 - 5k Bq/cm2) at different sites was measured bychanging the tilt angle to 10°, 40°, 60°, 90°, 120°, to evaluate the correlation
between β γ surface contamination density and amount of noise.

Correlation between β γ surface contamination density
and noise was obtained. The amount of noise
increased 20 times than the functional test.

Evaluation of correlation ofamount of signals with respect
to sources of α rays (Test ②)

Standard sources of α rays of varied strengths (Am-241, 220, 740, 2960Bq) werearranged in the angle of view when the tilt angle was 90°to obtain the correlation
between α surface contamination density and the amount of signals *1.

The sensitivity was 0.31±0.16[s-1/Bq/cm2] in theperformance verification test and it was 0.47[s-1/Bq/cm2]
in the laboratory test. Thus it was consistent with amargin of error.

Evaluation of dependency on
the measured distance (Test ③)

Standard source of α rays (Am-241, 2960Bq) was set up at locations with varied
measured distance to evaluate the dependency of α rays sensitivity on the distance.

It was found that just as in the functional verification test,
the sensitivity was inversely proportional to the square of
the distance. However, real-time correction was difficult
due to the impact of noise.

Evaluation of the accuracy inascertaining contamination
distribution (Test ④)

Based on data from multiple angles (10°, 40°, 60°, 90°, 120°) three dimensional
contamination distribution was developed from the difference with themeasurement results obtained when there was no radiation source, to evaluate theaccuracy of identifying locations with α surface contamination

The accuracy worsened than the position resolution when
the test was conducted using standard radiation source
due to the impact of noise, but it was possible toidentify the location of radiation source at a position
resolution of approx. 6 cm which is lower than thetarget 10 cm.

*1 Due to the impact of noise, the difference between the value of pixels with radiation source and the value of pixels without radiation source was calculated as the signal originating from the radiation source.

(Counting rate = a /(measured distance)2+b)

Test ②740Bq

Test ④2960BqTest ④ Test ②
Difference *2

Laboratory test

α surface contamination density

Counting rate per unit
contamination density

Counting rate

220Bq

Range up towhich noise
has animpact*2 Difference in the data whenradiation source is present andabsent as explained in *1

Fig. 1 Correlation between α surface contamination
density and counting rate (Test ②)

Measured distance [m]Fig. 2 Dependency of α rays sensitivity on
the measured distance (Test ③)

Fig. 3 Distribution after subtracting noise from themeasurement result (Test ④)©International Research Institute for Nuclear Decommissioning132

(b) Technological development of contamination evaluation for sorting solid waste

Identification of issues: Investigation of site conditions
Example 1) Unit 3 -1F- western passageway

Example 2) Unit 2 -5F- Eastern service entrance

Table 1 Issues and factors identified in Unit 3

Table 2 Issues and factors identified in Unit 2

Items

Identified issues and factors

Items

Identified issues and factors

Measurementof α rays

Wall

Paint

Wall

Paint

Floor

Paint (water accumulated)

Measurement of
α rays

Floor

Paint

CablesBlack

Cables

None

Sheet

None

Sheet

Pink covering sheet

Light
shieldingproperty

No light emitting
substance

Light shielding
property

External light present

Distancemeasurement

No impact

Accessibility

Unevenness of floor surface

Distancemeasurement

Impact of equipment likely

Accessibility

Unevenness of floor surface, dispersion ofequipment and material©International Research Institute for Nuclear Decommissioning133

(b) Technological development of contamination evaluation for sorting solid waste

Mapping of site investigation results
Covering sheet on residual material

Covering sheet on shelf

Light leakage from service entrance
Equipment has covering sheet Up to □5m
1m x 2m x 2m (H)Up to 2m
Several cableson the stud bolt
tensioner

Covering sheet on the
RPV head support

Up to 5m
Up to 1.1m

Fig. 1 Unit 2 site investigation results
Covering sheet on
the stud bolt tensioner
and shelf

Translucent plate

Fig. 2 Unit 3 site investigation results

Accumulated water

Several cables

High β γ
contamination density

Light leakage (external light present)

Covering sheet©International Research Institute for Nuclear Decommissioning134

(b) Technological development of contamination evaluation for sorting solid waste

Consolidation of issues: Items that are likely to have an impact on measurement
Items that are likely to have an impact on measurement are identified from the perspective of measuring thecontamination distribution of the building before dismantling.
Table 1
Target system
α camera

Emission noisedue to the impactof β rays, γ rays

Impact on
amount of α raysignals

Limiting
conditions during
measurement
Distance measurement

Access cart

Items that must be evaluated during site investigation
Items

Event

Site conditions that must be evaluatedScintillation lightLight emission of nitrogen (air)－Light emission besides nitrogen

Paint, type of sheets, etc.

Cherenkov light

Cerenkov light emission frommatter having a refractive index

Paint, type of sheets, etc.

Material to be measured

Change in α ray sensitivity due toroughness of the surface of thematerial

Presence of rubble(More the roughness higher the sensitivity)

Wetness of the surface

Presence of water accumulation

Light shielding property

Impact of lighting

Presence of emergency lighting
Light leakage from outside the building

Impact of matter in thevicinity

Shooting angle

Direction of access

Infrared transmission

Materials

Infrared absorption

Materials

The cart cannot be carried in

Presence of steps, width of thepassageway, etc.

Route for carrying-in©International Research Institute for Nuclear Decommissioning

135

(b) Technological development of contamination evaluation for sorting solid waste

Issue consolidation drill down: Issues pertaining to the α contamination measurement
systemGoal

Implementation items

Occurred phenomenon

Influencing factor

To indicate theapplicability of the αcontamination
measurement system

Clear specification of
the lower detection
limit of the α ray
measurement system

Impact on α raysignals

Probability of themeasurement method
Change in α ray sensitivity
due to roughness of thesurface of the material
Wetness of thesurface

Noise due to lightemission from β rays,
γ raysScintillation light

Environmental / material factors

Countermeasure policy

Shot blasting concrete・It has been confirmed thatthe event is on the highly
sensitive side (FY2020)

Water accumulation, etc.

Material in the nitrogenlight emission band
Material outside thenitrogen emission band
Water

Cherenkov light・In principle, α measurement is
not possible.
Avoid by removing in advance.
High priority items for whichmeasures involving α camera
must be undertaken

Water
Paint, sheets
Optical system

Light shielding
property

Lighting, external light

Gap with the wall, etc.
Emergency lighting, etc.

・Additional verificationof the mapping
function

Clear specification of
the range in which
distance can bemeasured・Shielding from external light(operation)
・Partial light shielding
mechanism
・Verified that mapping ispossible when the flange tankpiece is measured (FY2020)

Measurement notpossible

Shooting distance andangle

Roughness of surface
Reflectivity(Permeation/absorption)

Development or verification complete
Development underway or requires development

Direction of access

Material

・Present the range within
which distance can bemeasured with shooting angle,
material as parameters.

Fig. 1 Drilling down into issues pertaining to the αcontamination measurement system©International Research Institute for Nuclear Decommissioning

136

(b Technological development of contamination evaluation for sorting solid waste

Initiatives for addressing α camera issues
Noise due to lightemission from βrays, γ rays

Material in the nitrogenlight emission band

Scintillation lightMaterial outside thenitrogen emission band

Water
Water

Cherenkov light

Paint, sheets
Optical system<Matters that must be considered with respect to issues>① Amount of light emitted by various material and thesignals detected by α camera need to be closely
examined② Actions need to be taken so as to be able to deal withunanticipated light emitting sources at the site
③ Measures must be taken in response to items for whichmeasures can be taken in advanceFig. 1 Excerpts from the process of drilling down issues①, ②: Makes it possible to quantitatively ascertain each light emission and provide estimates based on simulation.
③: The possible countermeasures are comprehensively studied based on the characteristics of light emission.

Table 1 Light emission sources and their characteristics
Light emission
phenomenon

Light emission source

Characteristics of light emission

Scintillation light

Nitrogen in air

・Light emission by nitrogen due to α rays as well as β rays has the same
wavelength distribution
・The condition of spatial light emission is likely to differ depending on the differencein the range of α rays and β rays
・Duration of light emission depending on the length of the range

Other than nitrogen (Paint,
covering sheet)

・The amount of light emitted differs depending on the material
・The wavelength distribution differs depending on the material
・The duration of light emission is in the ns order

Paint
Sheet
Water
Optical system

・The wavelength distribution is ∝λ-2, which is different from the light emission bynitrogen due to α rays
・Amount of light emission depends on the refractive index n.
・The direction of light emission is conical at an angle of cosθ=c/nv (v: velocity in a
medium) with respect to the incident direction of charged particles.
・The duration of light emission is in the ps order

Cherenkov light©International Research Institute for Nuclear Decommissioning137

(b) Technological development of contamination evaluation for sorting solid waste

Course of action for FY2021

740Bq

220Bq

Status in FY2020 and preconditions for evaluating noise• Surface dose rate of β rays (β) is about 20mSv/h• The noise signal count rate is about 500cps
Overview of the FY2021 policy
1. Quantitative ascertaining of the light emission process

2960Bq1-1. Light emission test using sources of β rays
1-2. Development of simulations
1-3. Quantitative ascertaining of the light emission
phenomenon
Range up towhich noise
has an impact

Fig. 1 FY2020 mock-up test results

Implemented up to quantitative ascertaining of lightemission by nitrogen due to α rays and β rays

2. Noise countermeasures
1-1. Verification of the light emission measurement
characteristics of α camera and comprehensive study basedon the characteristics of each light emission
1-2. Evaluation of the effectiveness of noise countermeasures
Effectiveness of various measures evaluated©International Research Institute for Nuclear Decommissioning138

(b) Technological development of contamination evaluation for sorting solid waste

Overview of the method of quantitative organization of the light emission process- Process of light emission by nitrogen due to α rays Event

Evaluation method

Evaluation items

Bibliography

EnergyRange of α rays

Analysis

In air range

Energy transfer

Analysis

Energy imparted

Light emission

Bibliography

Amount of lightemission

Spread of photons

Test

Solid angle

α contamination
sourceIncident to the opticalsystemNumber ofphotons

Concentrationof light

Reflectivity

Incident to thephotoelectric surface

Quantum
efficiency

Electronmultiplication

Designvalue

Signal detection

Designvalue

Signalprocessing

Bibliography and analysis
・Bibliographic survey
・Estimation by means of on-paper calculations
based on literature data, etc.

Test / Geant4・Acquisition of basic data through α rays lightemission test
(Previously reported: Implemented in FY2020)・Development of simulation model using Geant4・Comparison between the test results and thesimulation, improvement in simulation accuracy

Design value
・Re-verification of design information of the αcamera detection system

Fig. 1 Flowchart of light emission bynitrogen due to α rays
©International Research Institute for Nuclear Decommissioning139

(b) Technological development of contamination evaluation for sorting solid waste

Overview of the method of quantitative organization of the light emission process- Process of light emission by nitrogen due to β rays Event

Evaluation method

Evaluation items

Source of βcontamination

Bibliography

EnergyRange of β rays

Bibliography

In air range

Energy transfer

Analysis

Energy imparted

Light emission

Bibliography

Amount oflight emission

Test / Geant4

Solid angle

・Acquisition of basic data through β rays lightemission test
・Development of simulation model using
Geant4・Comparison between the test results and thesimulation, improvement in simulation accuracy

Spreadof photons

Test

Incident to the opticalsystemNumber ofphotons

Concentrationof light

Reflectivity

Incident to thephotoelectric surface

Quantum
efficiency

Electronmultiplication

Designvalue

Signal detection

Designvalue

Signalprocessing

Bibliography and analysis
・Bibliographic survey
・Estimation by means of on-paper calculations
based on literature data, etc.

Design value
・Re-verification of design information of the αcamera detection system

Fig. 1 Flowchart of light emission bynitrogen due to β rays
©International Research Institute for Nuclear Decommissioning140

(b) Technological development of contamination evaluation for sorting solid waste

Purpose of the β ray light emission test and means of implementation(1) Ascertaining the amount of light emission under high β radiation dose: Straight line system
Source ofβ raysCameraobscuraβ raysLightemission
Blackout
sheet

Sampleα camera・Source of β rays is equivalent to 1F mock-up
environment (Approx. 15mSv/h)・Only scintillation light (nitrogen light emission)
when samples are not provided.
・Scintillation light as well as Cherenkov lightwhen samples are provided.

Fig. 1 Schematic diagram of the set-up for testing lightemission from high dose β rays

(2) Separating the scintillation light and Cherenkov light to ascertain the amount of light emission:
Right-angle system
Light emission (Cherenkov)
SampleCameraobscuraLight emission (Scintillation)

β raysα camera

Source of β rays

・Considering the direction of emission of Cherenkov
light, the source of β rays - sample - α camera areplaced at right angles.
・Light emission is measured in varied conditions ofoccurrence of Cherenkov light by using high energy
as well as low energy* radiation sources.
*Radiation sources that have very few β rays thatexceed critical energy, are selected.

Fig. 2 Schematic diagram of the test set-up in which lightemission phenomena are separated©International Research Institute for Nuclear Decommissioning141

(b) Technological development of contamination evaluation for sorting solid waste

Test set-up for the β ray light emission test
Table 1 List of sources of β rays
Cameraobscuraα camera

SamplePowersupplySource ofβ rays

Dose rate*1

Distance*2

Sr-90/
Y-90

15.6
mSv/h
(98mSv/h)

50cm(20cm)

Nitrogen lightemission, scintillation
light, Cherenkov light

Pm-147

0.21
mSv/h

20cm

Scintillation light(excluding
Cherenkov light)

Evaluation
targetExternal
appearance

β rays*1 Dose rate at the surface where the sample is placed*2 Distance from the source of β rays to the sample

Fig. 1 External appearance of a straight line system

α camera

Source ofβ raysRadiation
sourceCameraobscuraPowersupply・Sr-90/Y-90 are measured at a distance where
the dose rate is equivalent to the dose ratearound the 1F flange tank piece.
・Scintillation light and Cherenkov light aremeasured when nitrogen light emission +sample are placed.
・As a maximum of approx. 220keV of β rays
are produced by Pm-147, the amount ofCherenkov light emitted is so small that theamount detected by α camera from thedirection of occurrence can be ignored.

Fig. 2 External appearance of a right angled system©International Research Institute for Nuclear Decommissioning142

(b) Technological development of contamination evaluation for sorting solid waste

Energy distribution of the β ray source

SOURCEMean
Std Dev
Entries

0.12

100
0.78490.5498

0.08

a.u.

a.u.

0.1

0.06

0.04

0.02

0
0

0.2

0.4

0.6

0.8

1

1.2

1.4

16

1.8

2

Energy (MeV)
Fig. 1 β ray energy distribution of Sr-90/Y-90

2.2

1.30E-01
Critical energy
1.20E-01
1.10E-01
257keV (Refractive index
1.00E-01
1.33 Water)
9.00E-02
8.00E-02
7.00E-02
6.00E-02
184keV (Refractive index
5.00E-02
1.5 Acrylic)
4.00E-02
3.00E-02
2.00E-02
1.00E-02
0.00E+00
0.000 0.030 0.060 0.090 0.120 0.150 0.180 0.210 0.240
Energy (MeV)

Fig. 2 β ray energy distribution of Pm-147©International Research Institute for Nuclear Decommissioning143

(b) Technological development of contamination evaluation for sorting solid waste

Straight line system for the β ray light emission test
Organization of light emission phenomenon
Cameraobscura wall

Nitrogen light emission ②

SampleFrame (Aluminum frame)
SampleBlack-out sheet
β raysPermeated β rays

α camera
Sr-90/Y-90

Nitrogen light emission ①

75mm⑤ Scintillation light + ⑥ Cherenkov light

Fig. 2 External appearance of sample
Pink covering sheet

Fig. 1 Light emission phenomena in a Sr-90/Y-90 straight line system

Table 1 List of light emission phenomena in a Sr-90/Y-90 straight line system
No.

Light emission/phenomena

Memo

①

Nitrogen light emission before the sample

Nitrogen light emission in the 50cm between the radiation source andthe sample②

Nitrogen light emission after the sample

Nitrogen light emission from attenuated electrons③

Shielding of ① due to the sample

Permeability differs depending on the sample④

Shielding of ① due to the frame

Shielding effect is assumed to be the same regardless of the conditions⑤

Light emission from the sample Scintillation light

Phenomenon to be clarified

1⑥

Light emission from the sample Cherenkov light

Phenomenon to be clarified

2⑦

Light emission inside the α camera

Generated by the bremsstrahlung or the electrons that have reached theoptical system©International Research Institute for Nuclear Decommissioning144

(b) Technological development of contamination evaluation for sorting solid waste

β ray light emission test : Test conditions of the straight line system and affecting factors
Table 1 Sr-90/Y-90 straight line system: Affecting factors in each condition
No.

SampleSupplementary
information①
Pre N2②
Post N2③
Shieldedbysample④
Shieldedby frame⑤
Scintillati
on light⑥
Cherenkov light⑦
Light
emission
insidethecameraLegend〇: There is an impact△: Although small there is an
impact―: There is no impact

〇

〇

－

－

－

－

〇

t0.3mm

△

〇

〇

〇

〇

〇

〇

White paintt0.5mm

△

〇

〇

〇

〇

〇

〇

4

White paintt1.0mm

△

〇

〇

〇

〇

〇

〇

5

Gray paintt1.0mm

△

〇

〇

〇

〇

〇

〇

6

Black paintt1.0.mm

△

〇

〇

〇

〇

〇

〇

7

Pink covering sheet

〇

〇

△

〇

〇

〇

〇

8

Transparent covering
sheet

〇

〇

△

〇

〇

〇

〇

9

α camera lightshielding

－

－

－

－

－

－

〇

Processing of detection results

10

Frame only

〇

〇

－

〇

－

－

〇

11

Frame + shielding

△

〇

〇

〇

－

－

〇

12

Transparent covering
sheet + shielding

△

〇

〇

〇

－

－

〇

・Measurement is started after
sample to be measured is set up.
・After completing measurement,
the radiation source is removed
and BG is measured.
・The difference between BG andthe sample measurement for eachch is obtained and that value isused as the evaluated test result.

1

None (Nitrogen)2

White paint

3

1.3

2.3

2.7

3.1

3.2

2.7

2.1

1.0

Analysis of test results

1.8

3.5

4.9

5.5

5.1

4.3

3.2

1.5

2.4

5.2

7.4

9.9

9.2

6.2

4.2

2.1

① Sum total of the data in the 2×2 cells in the center→ Size of view equal to that of the sample

2.6

5.9

9.5 37.9 32.1

8.2

4.6

2.2

2.6

5.8

9.7 37.2 32.1

8.3

4.6

2.2

2.1

4.8

6.9

8.8

8.7

6.3

4.1

2.1

1.8

3.2

4.3

4.9

4.5

4.0

2.9

1.6

1.3

2.2

2.7

2.9

2.9

2.6

2.1

1.1

Other test conditions
・Measurement time 10 min
・BG measurement 5min・Signal is the difference with BG
・Room temperature 21.3℃ to 21.9℃
・Atmospheric pressure 1005.8hPa to1011.2hPa・Humidity 26.7% to 41.7%② Sum total of the data in the 4×4 cells in the center→ Light from outside the frame is considered as well
③ Sum total of the data in all the 8×8 cells in the center→ Entire light emission is considered

Fig. 1 Example of test results (No. 7) Unit: cps©International Research Institute for Nuclear Decommissioning145

(b) Technological development of contamination evaluation for sorting solid waste

β ray light emission test: Straight line system test results

Unit: cps

0

20

40
0.8

1.1

1.1

1.2

1.3

1.2

1.1

0.7

0.9

1.0

1.1

1.1

1.1

1.1

0.9

0.8

1.0

1.2

1.2

1.1

1.1

1.0

1.1

0.9

1.2

1.2

1.0

1.1

1.0

1.0

1.0

0.9

1.6

1.0

1.2

1.2

1.2

1.1

1.0

1.0

0.9

3.5

1.6

1.1

1.2

1.3

1.1

1.2

1.1

1.2

1.0

3.2

2.5

1.2

0.9

0.9

1.2

1.0

1.1

1.0

1.0

0.8

2.1

1.9

1.5

0.7

0.7

1.0

1.3

1.3

1.2

1.1

1.1

0.6

1.9

2.1

1.8

1.4

0.6

1.4

2.0

2.7

3.2

2.9

2.8

2.1

1.1

3.7

3.5

3.1

2.3

1.0

1.8

3.5

4.6

5.1

4.7

4.0

3.0

1.4

5.8

6.0

4.3

3.0

1.4

2.4

5.1

6.8

8.2

7.7

5.9

3.9

1.9

6.0 11.0 10.3

5.2

3.2

1.4

2.8

5.7

8.1 15.5 14.2

6.8

4.2

2.2

5.8 10.7 10.7

5.2

3.3

1.6

2.7

5.5

8.2 14.7 14.1

6.6

4.1

2.1

3.3

4.5

5.5

5.5

4.1

3.0

1.4

2.3

4.5

6.3

7.4

7.3

5.6

4.0

2.0

1.2

2.6

3.1

3.2

3.3

2.8

2.2

1.2

1.8

3.3

4.0

4.4

4.3

3.5

3.0

1.6

0.7

1.4

1.8

1.8

1.8

1.7

1.4

0.6

1.1

2.1

2.5

2.9

2.8

2.6

2.0

1.0

1.3

2.3

2.7

3.1

3.2

2.7

2.1

1.0

1.0

2.1

2.5

2.8

2.5

2.5

2.0

0.9

1.8

3.5

4.9

5.5

5.1

4.3

3.2

1.5

2.4

5.2

7.4

9.9

9.2

6.2

4.2

2.1

2.6

5.9

9.5 37.9 32.1

8.2

4.6

2.2

1.6

2.6

5.8

9.7 37.2 32.1

8.3

4.6

2.2

2.4

5.0

7.5

9.9 10.5

6.7

4.0

2.2

3.2

1.7

2.1

4.8

6.9

8.8

8.7

6.3

4.1

2.1

2.0

4.2

5.7

6.5

6.7

5.3

3.6

2.0

3.1

2.5

1.3

1.8

3.2

4.3

4.9

4.5

4.0

2.9

1.6

1.7

3.2

3.9

4.2

4.1

3.5

2.8

1.6

2.2

2.0

1.5

0.7

1.3

2.2

2.7

2.9

2.9

2.6

2.1

1.1

1.1

1.9

2.4

2.5

2.6

2.4

1.9

1.0

2.1

2.3

1.9

1.4

0.7

1.2

2.4

2.7

3.0

2.6

2.7

2.2

1.1

3.5

3.5

3.0

2.5

1.0

1.8

3.8

4.6

5.2

5.0

4.1

3.3

1.6

4.8

5.9

5.6

4.4

3.1

1.4

2.5

4.9

7.1

8.6

8.1

5.8

4.1

2.1

5.5

8.9

9.1

5.4

3.3

1.6

2.7

5.8

8.6 20.2 17.9

7.1

4.2

2.2

3.8

5.8

7.6

8.5

5.6

3.2

1.5

2.5

5.7

8.5 19.3 17.4

7.2

4.2

2.1

1.5

3.2

4.4

5.1

5.2

4.1

3.0

1.5

2.1

4.8

6.4

7.8

7.6

5.8

3.9

2.1

2.1

4.3

6.0

6.8

6.8

5.3

3.7

1.9

1.0

2.3

2.9

3.1

3.2

2.9

2.4

1.1

1.8

3.4

4.2

4.4

4.4

3.9

3.0

1.7

1.7

3.2

3.9

4.2

4.1

3.5

2.8

1.5

0.7

1.5

1.8

1.8

1.8

1.8

1.4

0.7

1.2

2.1

2.7

2.8

2.8

2.6

2.0

1.1

1.1

1.9

2.5

2.6

2.5

2.3

1.8

0.9

1.7

3.1

3.9

4.1

3.9

3.5

2.9

1.7

2.5

4.8

5.9

6.8

6.4

5.4

4.1

2.1

3.3

6.4

9.1 11.3 10.5

7.6

5.3

2.7

3.8

7.4 11.1 20.2 19.7

8.8

5.4

2.7

3.5

7.4 11.1 20.1 19.6

9.0

5.5

3.2

6.0

8.6 10.7 10.2

7.3

2.4

4.6

5.5

6.3

5.9

1.9

3.1

3.6

3.8

0.8

1.5

1.9

1.1

2.9

3.6

1.7

3.9

5.3

2.1

4.5

1.8

4.4

6.3

1.6

3.6

1.1

0.7

1.4

1.9

2.1

2.1

2.1

1.5

0.7

1.1

2.5

3.2

3.8

3.9

3.4

2.6

1.1

1.5

3.2

4.3

6.0

6.5

5.0

3.7

1.6

1.7

3.8

4.9

9.1 11.1

6.3

3.6

1.8

2.9

1.7

3.6

5.4

9.0 11.4

6.4

3.6

4.9

2.7

1.6

3.2

4.5

5.6

6.1

4.9

5.1

4.0

2.2

1.2

2.5

3.3

3.4

3.6

3.8

3.5

2.8

1.5

0.7

1.5

1.8

2.1

2.2

2.1

1.9

1.5

0.5

0.8

1.5

1.9

3.9

3.8

3.3

2.4

1.2

1.2

2.6

3.4

6.7

6.3

4.8

3.3

1.6

1.5

3.6

4.8

6.5 10.1

9.7

5.7

3.4

1.6

1.9

4.2

8.6

8.9

5.5

3.4

1.4

1.6

3.9

4.9

5.5

5.4

4.3

3.1

1.5

1.5

2.6

3.1

3.3

3.3

2.8

2.3

1.2

0.9

1.5

1.9

1.9

1.9

1.7

1.5

0.7

0.9

1.6

2.0

2.3

2.3

2.1

1.6

0.6

1.3

2.9

3.6

4.1

4.1

3.4

2.6

1.2

1.8

3.8

5.2

6.6

6.5

5.1

3.7

1.6

1.9

4.5

5.9 10.6 11.2

6.2

3.7

1.7

1.7

4.1

6.0

9.1 10.1

6.0

3.5

1.7

3.4

4.7

5.6

5.9

4.8

1.2

2.4

3.2

3.5

3.6

0.7

1.6

2.0

2.1

0.7

1.5

2.0

1.0

2.6

3.2

1.5

3.6

1.8

4.0

1.7

No. 1
No sample(Nitrogen)

No. 2
White paintt0.3mmNo. 3
White paintt0.5mmNo. 4
White paintt1.0mmNo. 5
Gray paintt1.0mmNo. 6
Black paintt1.0mmNo.9
α camera light
Shielding

No. 10
Frame only

No. 7
No.11
1.7 3.3 4.2 4.8 4.6 3.9 2.9 1.6
Pink
2.2 4.6 6.2 7.5 7.1 5.3 3.8 1.8 Frame +Covering sheet 2.4 5.1 7.2 10.2 10.6 6.6 3.9 2.1 shielding

1.1 1.9 2.4 2.7 2.7 2.4 1.7 0.7
No.8
No. 12
1.7 3.4 4.2 4.8 4.3 3.9 2.7 1.4
Transparent
Transparent
2.2 4.7 6.4 7.9 7.6 5.4 3.6 1.8
Covering sheet 2.6 5.4 7.7 16.3 15.4 6.7 4.0 1.9 covering sheet
2.4 5.3 7.9 15.9 15.0 6.6 3.9 2.1+ shielding

Fig. 1 Test results under all conditions©International Research Institute for Nuclear Decommissioning146

Legend Results ofobtaining the differential〇: There is an impact△: Although small there is an
impact―: There is no impact(b) Technological development of contamination evaluation for sorting solid waste

β ray light emission test: Analysis of straight line system test results
Table 1 Results of difference processing and remaining effect④
Shieldedby frame⑤
Scintillati
on light⑥
Cherenkov light⑦
Light
emission
insidethecameraSampleSupplementar
y information①
Pre N2②
Post N2③
Shieldedbysample2-11

White paintt0.3mm

－

△

△

－

〇

〇

－

3-11

White paintt0.5mm

－

△

△

－

〇

〇

－

4-11

White paintt1.0mm

－

△

△

－

〇

〇

－

5-11

Gray paintt1.0mm

－

△

△

－

〇

〇

－

6-11

Black paintt1.0mm

－

△

△

－

〇

〇

－

8-11

Transparent covering
sheet

△

△

△

－

〇

〇

－

8-12

Transparent covering
sheet

－

△

△

－

〇

〇

－

7-11

Pink covering sheet

－

△

△

－

〇

〇

－

1-9

No sample

〇

〇

－

－

－

－

－

Differential
No.*1

Nitrogen lightemission

2×2*2

4×4*2

8×8*2

-3.9

-15.2

-41.1

-0.1

-10.0

-31.4

-7.1

-23.7

-56.1

-0.6

-12.9

-39.9

1.6

-14.2

-46.3

33.7

43.8

55.7

12.2

19.7

32.4

98.2

119.2

134.1

75.3

176.9

322.3

Difference (cps)*1 Difference with the test results on P.146, *2 Refer to Fig. 1 on P.145

200
150

Covering sheet: Lot of pink light emission

100

Transparent covering sheet

Paint: Little light emission

50

Pink covering sheet

0
-50
-100

2-11

3-11

4-11

5-11

6-11

8-11

8-12

7-11

Corresponding differential no.

Fig. 1 Results of difference processing and remaining effect (including statistical error)©International Research Institute for Nuclear Decommissioning147

(b) Technological development of contamination evaluation for sorting solid waste

Right-angled system for the β ray light emission test
Organization of light emission phenomena
Black-out sheet

Cameraobscura wall
① Nitrogen lightemission

Camera obscura wall

⑤ Cherenkov light

β raysβ rays

Sr-90/Y-90① Nitrogen lightemission⑤ Cherenkov
light

Pm147
④ Scintillation light④ Scintillation light

Scattered β rays

Scattered β rays
α camera

α cameraFig. 1 Light emission phenomenon in a Sr-90/Y90 right-angled system

Fig. 2 Light emission phenomenon in a Pm-147
right-angled system

Table 1 List of light emission phenomena in a right-angled system
No.

Light emission/phenomena

Memo

①

Nitrogen light emission before the sample

Nitrogen light emission in the 20cm between the radiation source andthe sample②

Nitrogen light emission after the sample

Nitrogen light emission from scattered electrons③

Reflection of ① by the sample

Reflectivity differs depending on the sample④

Light emission from the sample Scintillation light

Phenomenon to be clarified

1⑤

Light emission from the sample Cherenkov light

Phenomenon to be clarifiedof Pm-147

2, level that can be ignored in the case⑥

Light emission inside the α camera

Generated by the bremsstrahlung or the electrons that have reached theoptical system©International Research Institute for Nuclear Decommissioning148

(b) Technological development of contamination evaluation for sorting solid waste

β ray light emission test: Test conditions of the right-angled system and affecting factors
Table 1 PM-147 straight line system : Affecting factors in each condition
No.

SampleSupplementary
information①
Pre N2②
Post N2③
Reflected
by sample④
Scintillation
light⑤
Cherenkov
light

⑥
Light
emission
inside thecameraNone (Nitrogen)

〇

－

－

－

－

－

2

Pink covering sheet

〇

－

△

〇

－

－

3

Transparent covering
sheet

〇

－

△

〇

－

－

4

White

Paint

〇

－

△

〇

－

－

5

Light gray
(Gray)

Paint (Same as theflange tank piece)

〇

－

△

〇

－

－

1

Table 2 Sr-90/Y-90 straight line system : Affecting factors in each condition①
Pre N2

②
Post
N2

③
Reflected
by sample④
Scintillation
light⑤
Cherenkov
light

⑥
Light
emission
inside thecameraNone (Nitrogen)

〇

－

－

－

－

〇

7

Pink covering sheet

〇

〇

△

〇

－

〇

8

Transparent covering
sheet

〇

〇

△

〇

－

〇

9

White

Paint

〇

〇

△

〇

－

〇

10

Light gray (Gray)

Paint (Same as theflange tank piece)

〇

〇

△

〇

－

〇

11

α camera lightshielding

－

－

－

－

－

〇

No.

Sample6

Unit: cps

Supplementary
information

0.1

0.0

0.1

0.0

0.1

0.0 -0.1

0.0

0.2

0.3

0.2

0.2

0.2

0.0

0.0

0.0

0.4

1.2

1.1

0.6

0.2

0.1

0.1

0.1

0.6

7.1

5.7

1.9

0.2

0.1

0.0

0.1

0.5 10.0

8.9

3.5

0.2

0.1

0.0

0.0

0.3

1.8

1.7

0.8

0.2

0.0

0.0

0.0

0.1

0.4

0.3

0.2

0.1

0.1

0.0 -0.1

0.0

0.1

0.2

0.2

0.1

0.1

0.0

0.0

Fig. 1 Data processing position in a rightangled system

Legend〇: There is an impact△: Although small there is an impact―: There is no impact
Other test conditions

・Measurement time 10 min
・BG measurement 5min・Signal is the difference with BG
・Room temperature 20.0℃ to 20.4℃・Pressure 1003.8hPa - 1022.3hPa・Humidity 31.2% to 43.0%

Processing of detection results
・Measurement is started after
sample to be measured is set up.
・After completing measurement,
the radiation source is removed
and BG is measured.
・The difference between BG andthe sample measurement for eachch is obtained and that value isused as the evaluated test result.

① Sum total of the data in the 2×2 cells in the center→ Size of view equal to that of the sample② Sum total of the data in the 4×4 cells in the center→ Light from outside the frame is considered as well
③ Sum total of the data in all the 8×8 cells in the center→ Entire light emission is considered©International Research Institute for Nuclear Decommissioning149

(b) Technological development of contamination evaluation for sorting solid waste

β ray light emission test : Right-angled system test results
Results of Pm-147

Results of Sr-90/Y-90 (Reference)
No. 1
Pm Nitrogen

No. 5
Pm Gray paint

2

4

6

8

8

9

7

3

3

8

13

17

18

19

17

7

6

20

35

40

34

27

22

10

8

79 118

65

37

28

22

11

0.1

8 101 150

70

36

28

21

11

0.0

0.0

6

26

41

35

26

21

18

9

0.1

0.0

0.0

3

10

14

15

16

14

13

6

0.0

0.0

0.0

2

5

7

7

8

8

6

3

2

4

6

7

8

8

7

3

3

8

12

16

18

18

16

7

6

19

32

38

32

25

19

8

7

76

82

47

27

20

16

8

0.0

7

98

92

44

26

19

14

8

0.0

6

25

39

32

24

18

14

7

0.1

0.0 -0.1

3

9

13

15

16

14

12

6

0.1

0.1

0.0

0.0

3

5

7

7

8

7

6

3

0.1

0.0

0.0

0.0

0.0

2

4

6

7

8

9

8

4

2

4

6

8

8

9

7

4

0.1

0.0

0.0

0.0

0.1

3

8

12

16

19

20

18

8

3

8

13

16

19

19

17

7

5

17

31

39

36

31

25

12

5

18

32

39

34

27

21

9

7

75

79

54

42

33

26

13

7

75

81

49

32

24

19

9

7

97

87

52

40

32

25

13

7

95

90

45

31

23

17

9

6

23

37

33

29

25

20

10

6

24

39

34

27

21

16

8

0.1

0.0

0.0

0.1

0.0

0.0

0.0 -0.1

0.0

0.1

0.2

0.0

0.0

0.0

0.0 -0.1

0.3

1.0

0.8

0.3

0.1

0.0

0.0 -0.1

0.5

6.3

2.9

1.3

0.1

0.1

0.0 -0.1

0.0

0.6

8.7

4.1

2.3

0.0

0.0

0.0

0.0

0.0

0.3

1.5

1.4

0.6

0.1

0.0

0.0

0.0

0.0

0.1

0.3

0.4

0.2

0.1

0.0

0.0

0.0

0.0

0.2

0.2

0.1

0.2

0.1

0.0

0.1

0.0 -0.1

0.0

0.2

0.2

0.0

0.0

0.0

1.1

0.6

0.2

0.1

0.1

0.1

5.7

1.9

0.2

0.1

0.0

0.1

0.5 10.0

8.9

3.5

0.2

0.1

0.0

0.3

1.8

1.7

0.8

0.2

0.0

0.0

0.1

0.4

0.3

0.2

0.1

0.0

0.1

0.2

0.2

0.1

0.1

0.2

0.1

0.2

0.2

0.0 -0.1

0.1

0.0

0.0 -0.1

0.0

0.0

0.0

0.2

0.2

0.0

0.1

0.1

0.0

0.1

0.2

0.9

0.8

0.2

0.1

0.0

0.1

0.1

0.5

6.2

2.2

0.7

0.1

0.1

0.1

0.0

0.5

8.9

3.1

1.2

0.2

0.1

0.0

0.3

1.5

1.2

0.4

0.2

0.1

0.1

0.3

0.3

0.1

0.1

0.1

0.1

0.1

0.1

0.1

0.0

0.1

0.2

0.3

0.2

0.4

1.2

0.6

7.1

No. 2
Pm Pink
covering sheet

0

5

0

No. 3
Pm
0.4 1.0 1.0 0.4 0.1 0.1 0.0 0.1
Transparent
0.4 6.9 4.2 1.8 0.2 0.1 0.1 0.0
0.5 9.7 6.2 2.6 0.1 0.1 0.1 0.0 covering
0.4 1.6 1.5 0.5 0.1 0.0 0.1 0.0 sheet

10 Unit: cps

200 Unit: cps

100

No. 6
Sr Nitrogen

0.2

0.4

0.4

0.1

0.1

0.0

0.0

0.0

3

9

12

15

16

15

14

7

3

9

13

15

16

15

13

6

0.1

0.1

0.1

0.1

0.0

0.0

0.1

0.0

3

4

6

7

7

8

6

3

3

5

7

8

8

8

7

3

0.0

0.0

0.1

0.1

0.1

0.0

0.0

0.0

2

4

6

8

9

9

7

4

0.1

0.0

0.1

0.1

0.0

0.1

0.0

0.2

0.2

0.1

0.1

0.0

0.0

0.0

3

9

13

17

19

19

17

7

0.1

0.1

0.1

0.1

0.1

0.1

0.1

0.1

0.3

1.1

0.9

0.3

0.0

0.0

0.0 -0.1

6

21

36

42

35

27

22

0.5

7.1

3.0

1.5

0.0

0.0

0.1

0.0

9

82 136

72

37

27

22

No. 7
Sr Pink
10
covering sheet
11

0.1

0.1

0.6 10.2

4.3

2.6

0.1

0.1

0.0

0.0

9 105 183

83

36

26

20

11

0.0

0.1

0.1

0.0

0.1

0.1

0.1

0.1

0.3

1.6

1.4

0.6

0.2 -0.1

0.0

0.0

7

28

44

37

27

21

17

9

0.1

0.1

0.0

0.0

0.0

0.1

0.1

0.0

0.1

0.3

0.3

0.2

0.1

0.0

0.0

0.0

3

10

14

16

16

15

13

6

0.1

0.0

0.1

0.0

0.0

0.1

0.1

0.0

0.1

0.1

0.1

0.1

0.1

0.1

0.0

0.0

3

5

7

8

8

8

7

3

0.2

0.1

0.2

0.1

0.1

0.1

0.1

0.1

No. 4
Pm White
paint

No.8
Sr Transparent
covering sheet

No.9
Sr White paintNo. 10Sr Gray paintNo.11
Sr Camera
0.2 0.0 0.0 0.1 0.1 0.1 0.1 0.1
0.1 0.1 0.1 0.1 0.1 0.1 0.1 0.1 light shielding

Fig. 1 Test results under all conditions©International Research Institute for Nuclear Decommissioning

150

Legend(b) Technological development of contamination evaluation for sorting solid waste〇: There is an impact△: Although small there is
an impact―: There is no impact

β ray light emission test: Analysis of right-angled system test results
Table 1 Results of difference processing and remaining effect①
Pre N2②
Post N2③
Reflection④
Scintillation
light⑤
Cherenkov
light⑥
Light emission
inside the camera

〇

―

―

―

ー

ー

20.4

28.7

31.6

―

―

―

〇

ー

ー

11.3

17.3

19.0

―

―

―

〇

ー

ー

6.6

10.5

11.7

―

―

△

〇

ー

ー

4.2

7.7

7.8

―

―

△

〇

ー

ー

1.5

4.1

4.3

〇

〇

ー

―

ー

ー

337.6

648.6

1339.7

―

―

△

〇

ー

ー

167.6

249.0

206.9

Covering sheet

―

―

△

〇

ー

ー

110.7

159.2

111.7

Sr White

Paint

―

―

△

〇

ー

ー

11.1

2.3

-128.3

Sr Gray

Paint

―

―

△

〇

ー

ー

3.1

-0.8

-78.4

SampleSupplementary
information

1

Pm Nitrogen lightemission

No difference
processing

2-1

Pm Pink covering
sheet

3-1

Pm Transparent
covering sheet

4-1

Pm White

Paint

5-1

Pm Gray

Paint

6-11

Sr Nitrogen lightemission

No sample

7-6

Sr Pink

Covering sheet

8-6

Sr Transparent

9-6
10-6

Difference (cps)Differential
no.

20

2×2

4×4

8x8

Covering sheet: Scintillation light

15

Reflection influence ofnitrogen light emission

Amount of light emission
from Sr is absolutely small

10
5
0

2-1

3-1

4-1

5-1

Corresponding differential no.

Fig. 1 Results of difference processing and remaining effect (including statistical error)©International Research Institute for Nuclear Decommissioning151

(b) Technological development of contamination evaluation for sorting solid waste

Purpose of simulation and details of implementation<Purpose of simulation>
To secure α contamination detection performance range (lower α contamination detection limit
depending on the on-site environmental conditions) for β γ surface contamination that gets spatiallydistributed.
1. Simulation of α contamination density and spatial distribution
2. Simulation of β contamination density and spatial distribution
3. Provision as a tool for predicting factors that do not have a measured value or it is difficult tomeasure their value
Example 1
Impact of β rays that are incident on the optical system by chance
Example 2
Light emission of new material<Details of FY2021 implementations>
The following was implemented with a system similar to the β rays light emission test as the basicmodel.
1. Matching the actual nitrogen light emission counting rate in the system in which Am-241 is placedin space.
2. Matching the actual nitrogen light emission counting rate in the system in which Sr-90/Y-90 areplaced in space.

©International Research Institute for Nuclear Decommissioning152

(b) Technological development of contamination evaluation for sorting solid waste

Simulation

Geant4 basic model Straight line system

α camera

Radiation source
Irradiation conditions
Dose: 15.6mSv/h@50cm
Source: Sr-90/Y-90
Source size: φ6cm
Radioactivity (equivalent): 3.32×108Bq

830mm

550mm

Lens
□75mm

Camera obscura

MirrorSampleAnalytical conditions
Number of events: 6×109[beta]
Statistical amount: equivalent to 36.1[sec]Fig. 1 Basic simulation model©International Research Institute for Nuclear Decommissioning153

(b) Technological development of contamination evaluation for sorting solid waste

Simulation: α source system
0.1

0.1

0.3

0.3

0.1

0.1

0.1

0.1

0.2

0.4

0.7

0.9

0.5

0.2

0.1

0.0

0.2

0.9

3.2

4.8

2.5

0.7

0.1

0.1

0.4

1.1

5.3 21.4

3.8

0.8

0.2

0.1

0.2

0.5

3.0

4.5

3.0

0.6

0.2

0.2

0.3

0.3

0.6

1.0

0.6

0.3

0.1

0.1

0.2

0.2

0.2

0.2

0.2

0.2

0.1

0.1

0.1

0.1

0.1

0.1

0.1

0.1

0.0

0.1

Irradiation conditions
Source: Am-241 (5.4Mev, α)Source size: φ3.6cm
Direction of emission: 2π605.2mm

To be measured

Am-241

Simulation:
Actual measured value :Analytical conditions
Number of events: 1×106[Alpha]Statistical amount: equivalent toxxx[sec]

0.3 -0.2

0.2

0.8

1.2

0.0

0.4

-0.3

0.1

0.3

0.7

0.6

0.0

0.1 -0.1

-0.3

0.3

0.7

1.1

1.0

0.3 -0.3

0.3

-0.3

0.1

1.6 16.8

6.6

0.3

0.1

-0.1

0.4

0.8

4.4

2.2

0.0 -0.2 -0.2

0.0 -0.2

0.3

0.8

0.6 -0.1

0.0

0.3

-0.2

0.4

0.0

0.0

0.2 -0.3 -0.2

0.2

0.2

0.4

0.1

0.2

0.4 -0.6 -0.2

0.1

021.4cps
16.8cps

It was verified that simulation is possible with anaccuracy of 0.5≦ (calculated value)/(actual
measured value)≦2 (=Factor2) for radiation source
locations in multiple cases

Fig. 1 α rays nitrogen light emission
simulation results
605.2mm

Measured value of the visual field of thesource (Am-241)

20

40

0.5

0.4

Unit: cps

Fig. 2 α rays nitrogen light emission simulation results (previously reported in FY2020)©International Research Institute for Nuclear Decommissioning154

(b) Technological development of contamination evaluation for sorting solid waste

Simulation: β source straight line system
Results of evaluating the number of detections of β rays
nitrogen light emission by means of simulation

Comparative evaluation of the simulation and actual
measurement

1.9

3.7

4.4

4.3

4.3

4.4

3.5

2.1

1.1

1.2

1.1

1.1

1.1

1.3

1.2

1.2

3.9

5.7

6.6

7.9

8.0

7.4

6.0

3.1

1.5

1.2

1.1

1.2

1.3

1.4

1.5

1.4

4.0

6.0

8.5

9.1

9.1

8.5

7.3

3.3

1.2

0.9

0.9

0.8

0.9

1.1

1.4

1.2

3.6

7.0

9.1 11.1 11.2

8.2

6.9

3.4

Percentage 0.9

0.9

0.8

0.5

0.6

0.9

1.3

1.3

4.5

7.0

9.2 11.7 11.0

9.2

7.1

3.2

1.3

1.0

0.8

0.6

0.6

1.0

1.3

1.1

4.1

6.3

7.3

9.2

8.5

9.0

6.7

3.1

1.3

1.0

0.8

0.9

0.8

1.2

1.4

1.2

3.1

5.7

6.6

7.8

7.8

7.7

6.5

2.6

1.3

1.2

1.2

1.2

1.3

1.5

1.6

1.1

1.7

3.3

4.5

4.7

5.1

4.0

3.6

1.9

0.9

1.1

1.2

1.2

1.4

1.1

1.3

1.3

Results of evaluating the number of detections of β rays
nitrogen light emission by means of a test (Repeated)

0

0.5

1

2

Unit: -

1.7

3.1

3.9

4.1

3.9

3.5

2.9

1.7

0.2

0.6

0.5

0.2

0.5

0.9

0.6

0.4

2.5

4.8

5.9

6.8

6.4

5.4

4.1

2.1

1.4

0.9

0.8

1.1

1.6

2.0

1.9

0.9

3.3

6.4

9.1 11.3 10.5

7.6

5.3

2.7

0.8 -0.5 -0.6 -2.2 -1.4

1.0

2.0

0.6

3.8

7.4 11.1 20.2 19.7

8.8

5.4

2.7

-0.2 -0.4 -2.0 -9.1 -8.5 -0.6

1.4

0.7

3.5

7.4 11.1 20.1 19.6

9.0

5.5

2.9

1.0 -0.4 -1.9 -8.5 -8.6

0.2

1.6

0.3

3.2

6.0

8.6 10.7 10.2

7.3

4.9

2.7

0.9

0.3 -1.3 -1.4 -1.7

1.8

1.8

0.5

2.4

4.6

5.5

6.3

5.9

5.1

4.0

2.2

0.7

1.1

1.1

1.5

1.9

2.6

2.5

0.3

1.9

3.1

3.6

3.8

3.8

3.5

2.8

1.5

-0.2

0.2

0.9

0.9

1.3

0.5

0.8

0.4

40

Unit: cps

20

Differential

-10

0

10

Unit: ⊿ cps

Fig. 1 Results of comparative evaluation of β rays nitrogen light emission (Straight line system)

It was verified that simulation is possible with an accuracy of 0.5≦ (calculated value)/(measured
value)≦2 (=Factor2)©International Research Institute for Nuclear Decommissioning155

(b) Technological development of contamination evaluation for sorting solid waste

Organization of the light emission process: Light emission by nitrogen due to α rays
Event

Evaluation
methodEvaluation items

Evaluated value

α contamination
sourceBibliography

Energy5.486 Mev

Range of α rays

Analysis

In air range

4.1cm

Energy transfer

Analysis

Energy imparted

5.486 MeV/alphaLight emission

Bibliography

Amount of lightemission

Spread of photons

Test/
Geant4

Solid angle

2πIncident to the opticalsystemNumber of photons

2-3 Photon/Alpha

Concentration of light

Reflectivity

80% or more (λ>300 nm)

Incident to thephotoelectric surface

Quantum efficiency

Quantum efficiency (Refer to P.158
Appendix. 1)

Signal processing

Signal processing

Electron multiplication

Signal detection

20.38±0.98 photon/MeV
to 110 Photon/Alpha

Design value
Design value©International Research Institute for Nuclear Decommissioning156

(b) Technological development of contamination evaluation for sorting solid waste

Organization of the light emission process: Light emission by nitrogen due to β rays
Event

Evaluation
methodEvaluation items

Evaluated value

β contamination
sourceBibliography

Energy(Refer to P.143 Fig. 1)

Range of β rays

Bibliography

In air range

Up to 10m

Energy transfer

Analysis

Energy imparted

Refer to P.159 Appendix. 2

Light emission

Bibliography

Amount of lightemission

Spread of photons

Test/
Geant4

Solid angle

4π

Incident to the opticalsystemNumber of photons

3.6x10-8 Photon/Beta

Concentration of light

Reflectivity

80% or more (λ>300 nm)

Incident to thephotoelectric surface

Quantum efficiency

Quantum efficiency (Refer to P.158
Appendix. 1)

Signal processing

Signal processing

Electron multiplication

Signal detection

20.38±0.98 photon/MeV/ 2x10-6 Photon/Beta

Design value
Design value©International Research Institute for Nuclear Decommissioning157

(b) Technological development of contamination evaluation for sorting solid waste

Appendix. 1

Quantum efficiency of PMT
Wavelength range ofnitrogen lightemission due to α rays

©International Research Institute for Nuclear Decommissioning158

(b) Technological development of contamination evaluation for sorting solid waste

Appendix. 2 Stopping Power of β rays

“Stopping-Power & Range Tables for Electrons, Protons, and Helium Ions”
Source: National Institute of Standards and Technology (NIST) ESTAR Database
NIST Standard Reference Database 124
DOI: https://dx.doi.org/10.18434/T4NC7P©International Research Institute for Nuclear Decommissioning159

(b) Technological development of contamination evaluation for sorting solid waste

Quantitative organization of the light emission process<β rays light emission test>
・Light emission test with Sr-90/Y-90, Pm-147 as the radiation source was completed.
・It was verified that paint emits little light.
・It was verified that the covering sheet emits lot of light.
It was verified that scintillation light is generated. Meanwhile, it was found that whether or not Cherenkov light isgenerated needs to be studied in further detail.
・It was verified that there is lot of nitrogen light emission in air.
・It was confirmed that the noise in FY20202 was likely to be mainly due to covering sheet (pink) and nitrogen lightemission.
<Summary of simulation>
・As a result of developing a model simulating the phenomenon of light emission by nitrogen due to α radiation
source and evaluating the amount of light detected by the α camera, it was found that it had a Factor2 accuracy (incomparison with the actual measured value).
・As a result of developing a model simulating the phenomenon of light emission by nitrogen due to β radiation
source and evaluating the amount of light detected by the α camera, it was found that it had a Factor2 accuracy (incomparison with the actual measured value).
<Quantitative organization of the light emission process Important items>
・2 to 3 Photon were detected around the α rays, and 3.6×10-8Photon were detected around the β rays.
・The number of photons detected around α rays as well as β rays largely change depending on the contamination
density and the air dose rate.
・In the light emission test conducted this year, as against a maximum of approx. 17cps at 1ch for Am-241 of 900Bq,
in the case of Sr-90/Y-90 it was a maximum of approx. 20cps for a ch with an air dose rate of 15.6mSv/h.
・The model needs to be extended to include a spatial spread of contamination and change in density in its
simulation.

©International Research Institute for Nuclear Decommissioning160

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Overview of the policy
Table 1 Possible countermeasures for each light emission phenomenon
Light emission
phenomenon

Items that must be considered

Possible countermeasures

Scintillation light<Nitrogen light emission>・Light emission by nitrogen due to α rays as well as β rays has the same wavelength
distribution
・The state of light emission is likely to differ depending on the distribution anddensity of α contamination and β contamination.
・The duration of light emission is longer than that from the sample.
<Light emission other than from nitrogen>・The wavelength range differs depending on the material of the source of lightemission.
・The amount of light emitted differs depending on the material of the source of lightemission.
・Light spreads from where the source of light emission is present.
・The duration of light emission is in the ns order<Nitrogen light emission>
・Distinction using the difference in the state oflight emission
・Indication of the range within which α
contamination can be detected with respect toβ contamination (lower detection limit of α
contamination depending on the on-siteenvironmental conditions).
<Light emission other than from nitrogen>・Filter that uses the difference in wavelengthrange・Distinction using the difference in duration oflight emission
・Prevention by taking measures such as notusing samples that become a source of lightemission, elimination in advance, etc.
・Distinction using the concurrence of lightemission

Cherenkov light・The wavelength distribution is ∝λ-2・Amount of light emission depends on the refractive index n.
・The direction of light emission is conical at an angle of cosθ=c/nv (v: velocity in a
medium) with respect to the incident direction of charged particles.
・The duration of light emission is in the ps order・Filter that uses the difference in wavelength
distribution
・Distinction using the difference in duration oflight emission
・Distinction using the concurrence of lightemission<Noise countermeasures implementation items>
1. Measuring the scintillation light and Cherenkov light by means of the β rays light emission test
2. Checking the current α camera signal processing circuit, and verifying the possibility of temporal discriminationof each type of light emission by means of the above-mentioned test
3. Comprehensively determining the countermeasures that must be taken based on the above©International Research Institute for Nuclear Decommissioning161

(b) Technological development of contamination evaluation for sorting solid waste

Study of noise countermeasures: Current α camera signal processing circuit
The possibility of temporal discrimination based on the light emission phenomenon and the circuit
time constant is studied as a means for noise discrimination.

HV circuit board

Voltage monitor

HV module

FPGA card“Spartan6”

HV set-up
HV buffer

HV module
Upper set-up

Latch

Lower set-up

FPGA card“Cyclone®V”
Or “Cyclone®IV”

X 64 circuit board (Configuration: 16ch/ board x 4 boards, etc.)(In a batch)Preamplifier Waveform shaping unit

Fig. 1Power
supplyunit

Signal detection unit

Current α camera signal processing circuit

Portion of the circuit that affects the time constant (The effect is
explained in the following pages)©International Research Institute for Nuclear Decommissioning162

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Signal processing time constant and countTable 1 Signal processing from photon incidence to signal
Circuit

ImpactCharge sensitive
Gain (x500, 1MHz max)

Signal amplification

Baseline restorer
Pole-zero cancellation

Waveform shaping

V

I

t

Several μstFig. 1 Example of signals before and after signal processing (1 event)
Table 2 Signal counts
Circuit

ImpactWindow discriminatorCut at the upperand lowerV

1 event

Multiple eventsOther

2 Counts

No counts

Schmitt trigger

1 Count
Threshold range that sufficiently
includes light emission due to α, βFig. 2 Example of conversion to the countCurrent α camera circuit has a time constant in the μs order→ Distinction cannot be made based on duration of light emission by the Cherenkov light or the scintillation light©International Research Institute for Nuclear Decommissioning

163

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures : Results of evaluating the time constant of the current circuit(Verification of the actual signal)

Approx. 2μs

Frequency
Trigger typeedge

Fig. 1

Values
SourceStandard
deviation

Average value
Period cannot be foundCoupling

Around
800mV(Gain×500)

Slope

Level

Mode, normal
and hold-off

Signal waveform when α ray source is measured

・When 1 photon is incident, the output is 1.6mV, and by Gain×500 the peak height is around 800mV.
・The width of the output waveform is approx. 2μs between 10% and 90%.
→ The current α camera circuit does not have a temporal resolution of less than μs.
©International Research Institute for Nuclear Decommissioning164

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Study of the likelihood of speeding up
Table 1 Possible countermeasures for each light emission phenomenon
Light emission
phenomenon

Items that must be considered

Possible countermeasures

Scintillation light<Nitrogen light emission>・Light emission by nitrogen due to α rays as well as β rays has the same
wavelength distribution
・The state of light emission is likely to differ depending on the distribution anddensity of α contamination and β contamination.
・The duration of light emission is longer than that from the sample.
<Light emission other than from nitrogen>・The wavelength range differs depending on the material of the source of lightemission.
・The amount of light emitted differs depending on the material of the source oflight emission.
・Light spreads from where the source of light emission is present.
・The duration of light emission is in the ns order<Nitrogen light emission>
・Distinction using the difference in the state of lightemission・Indication of the range within which α
contamination can be detected with respect to β
contamination (lower detection limit of α
contamination depending on the on-siteenvironmental conditions).
<Light emission other than from nitrogen>・Filter that uses the difference in wavelength range
・Distinction using the difference in duration oflight emission
・Prevention by taking measures such as not using
samples that become a source of light emission,
elimination in advance, etc.
・Distinction using the concurrence of light emission

Cherenkov light・The wavelength distribution is ∝λ-2・Amount of light emission depends on the refractive index n.
・The direction of light emission is conical at an angle of cosθ=c/nv (v: velocity ina medium) with respect to the incident direction of charged particles.
・The duration of light emission is in the ps order・Filter that uses the difference in wavelength
distribution
・Distinction using the difference in duration oflight emission
・Distinction using the concurrence of light emission

1) Use of a high-speed amplifier with a band frequency of 100Ghz or more
→ Not realistic considering the cost of improving the circuit system, compatibility with 64ch, etc.
2) Replacement of the I/V amplifier of the ns order with a charge sensitive amplifier→ Since the peak height of the I/V amplifier depends on the number of incident photons, if the number of incident photons changes depending on the
duration of light emission, the peak height changes as well.
(Expected phenomenon)
In the case of scintillation light that is emitted for ns and Cherenkov light that is emitted for ps, if light is detected continuously throughout the durationof light emission, the peak height of the scintillation light rises.

Verification by means of the light emission test in which scintillation light and Cherenkov light are separated©International Research Institute for Nuclear Decommissioning165

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: I/V amplifier verification by means of the β rays light emission test

Layout ②

PMT(H3165-10)

Camera obscura50Ω

Layout ①

Black-out sheet

β raysI/V amplifier unit
C5594-44

Oscilloscope

PMT(H3165-10)

Light emission (Scintillation)

β ray source (Sr-90)

Fig. 1 Light emission measurement system using the I/V amplifier

Table 1 Layout and light emission phenomenon to be measured

PMT
LayoutPerceived light emission phenomenon①

Nitrogen light emission (scintillation light) was measured at the location
where incidence of the beams of electrons was avoided.

②

Nitrogen light emission was shielded and Cherenkov light wasmeasured by allowing incidence of electron beams.

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Noise countermeasures: Results of I/V amplifier verification by means of the β rays light emission test

Approx. -0.5V

Scintillation light from nitrogen
Fig. 1

Waveform in which light emission isdetected in layout ①

・Incidence of nitrogen light emission was 3.6×10-8Photon/Beta(P.157)
・1Photon to several Photon were detected in this system as
well

Approx. -2.5V

Cherenkov light from the window
Fig. 2 Waveform in which light emission is detected inlayout ②
・Number of incident photons was extremely numerous in
Cherenkov light from the window
・There were waveforms in which large quantities of photons
were detected in this system as well・As it is assumed that 1 photon onwards per event is detected at the site, for example in a low air dose rateenvironment it is difficult to identify based on peak height.
・In a high air dose rate environment where photons from multiple events are detected over a certain duration, the lightemission duration and number of incident photons are likely to be correlated, and in such a case it is possible to identify
them.
→ The possibility needs to be ascertained considering multiple factors such as spatial distribution of contamination, its
density, placement of the source of light emission, etc.
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Noise countermeasures: Simultaneous counting method*Principle of the simultaneous counting method and its effect•This is a technique for distinguishing between signal and noise using the concurrence of light emission detection. Hence thesame view is measured in 2 PMTs.
•
If nitrogen light emission is considered as the target example, as against β rays, since the number of incident photons per eventin the case of α rays are larger, the decrease in β rays nitrogen light emission counting rate can be expected to be far greaterthan the decrease in α rays nitrogen light emission counting rate. (S/N improvement can be expected)•
This effect is presumed to be a decreasing effect common to noise generated at random times such as thermal noise, electricalnoise, noise generated by light, noise generated by γ rays, etc. regardless of the light emission principle.
•
However, when the frequency of noise generation increases, noise gets accidentally counted simultaneously leading to a false
detection.
⇒ The sensitivity* of the signal from an ancillary detector that optimizes S/N should be evaluated.
*Generally, if the sensitivity to α rays are enhanced, noise sensitivity such as sensitivity to β rays increases as well.
Light emission of α rays:
・Amount of light emission: approx. 100
photon/α ray
・Duration of light emission: About
several nsec

Signal from α rays Signal that is generated
at random times(Detected)(eliminated)

Noise that is counted
simultaneously by
accident (false detection)

Detector 1(α camera)
Signal from
Detector 1

α rays

Detector 2(Ancillary detector)

Signal from
Detector 2

time
Fig. 1 Principle of the simultaneous counting method©International Research Institute for Nuclear Decommissioning168

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Study of the composition of the ancillary detector
Table 1 List of prospective elements

Elements affecting the detector sensitivity

Detector
8x8 array

Lens

Detector(PMT)

MirrorSingle PMT array /Line (2x2)
Single PMT of
about φ100mm
Single PMT of
about φ10mm

Angle of view

FilterBroad

Narrow

FilterLens

Band-pass filter(UG11, etc.)

φ100mm, short focus
lensFilter with a 300nm to 400nm
nitrogen light emission range
Single nitrogen light emission
wavelength filter such as with alight emission range of 337nm

Long focus lens ofabout φ10mm
None

None

・The combination of detectors that are expected to have the best lower detection
limit in a stand-alone set-up is selected.
・The sensitivity of the option that appears to be theoretically feasible is evaluated.

Shield for γ rays

Transmission ofultraviolet rays
Table 2 Results of evaluating the sensitivity for each composition
Composition

Sensitivity ratio

Remarks

Current system
8×8 array + Short focus lens1

Camera measurementtarget distance 1m

φ10mm Single PMT+ Long focus lens

4.2

φ100mm Single PMT

3.6

α raysLightemissionFig. 1 Principle of the α camera©International Research Institute for Nuclear Decommissioning169

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Preliminary evaluation of the simultaneous counting method
Table 2 Analytical conditions

Table 1 Evaluation system
No.

Classifi
cationCompositionA

Current system
8×8 array + Short focus lens

1

φ10mm Single PMT+ Long focus lensBC

φ100mm Single PMT

Lower detection limit (Bq/cm2)

The combination simulating the simultaneous
counting method
A, A-B, A-C, B-B (Reference)

Filter conditions

Transmission rate (%)

Remarks

Nitrogen lightemission

Cherenkov

275nm to 375nm

100

100

Past

2

315nm to 375nm

100

52.4

Entire nitrogen lightemission

3

337nm to 375

79

31Part of the nitrogenlight emission

4

337nm±10nm

42

18.2

Part of the nitrogenlight emission

12

Combination A
Combination A-B

9

Combination A-C
Combination B-B

6

Lower than the lower detection
limit of4Bq/cm2

3

0
1

2

3

4

→ Evaluation of the simultaneous
counting method with Combination
A-B

Filter condition number
Fig. 1 Results of evaluating the lower detection limit©International Research Institute for Nuclear Decommissioning

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Noise countermeasures: Test manufactured ancillary detector and the test set-up

Step
motor

Tilt

Filter

φ10mmPMT
H3165-10
Fig. 2 Composition of ancillary detector

Band-passfilterFig. 1 External appearance of the ancillary detector

Lens

Ancillary detector
H3165-10
φ10mm-PMT

Example of the simultaneous counting method

Trg-out
APU180164ch-PMT amplifier
Trg

Using only 1 ch

φ130mmLens unit
α camera
45° mirror

Fluorescence produced when
α particles collide against
nitrogen in air
4379 (s-1 in 2π)

φ100mmLens unit
H12428-100
64Ch array PMT

APU180164ch-PMT amplifier

tw

Ch1

→0

Ch2

→1

Ch3
:

→1

Trg-inSimultaneous counting
FPGA boardLAN

tw: 200ns, 500ns, 1μs, 1.5μs, 2μs

Software formeasurement

Windows-PC

Fig. 3 System for evaluating the simultaneous counting method©International Research Institute for Nuclear Decommissioning

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Noise countermeasures
Evaluation of the effectiveness of the simultaneous counting method
Testing conditions
Measurement time: 9720sec
tw (time gate width): 1μs
0Placement of Am-241

15

Unit: cps

8x10-20

Unit: cps

0.78 0.97 0.82 1.45 3.43 2.16 0.98 1.69

1.5E-02

2.9E-03

2.8E-03

2.4E-03

1.6E-03

1.1E-03

1.5E-03

1.5E-020.51 0.80 1.07 1.93 1.89 1.25 0.65 0.72

4.0E-03

1.0E-03

6.2E-04

8.2E-04

8.2E-04

7.2E-04

6.2E-04

4.5E-030.53 1.13 2.67 2.93 1.91 1.28 0.72 0.62

4.3E-03

1.0E-03

1.3E-03

3.9E-03

3.6E-032.0E-03

6.2E-04

3.2E-030.64 1.25 7.86 5.81 5.99 1.36 0.69 0.57

9.5E-03

1.9E-03

2.3E-033.0E-02

4.3E-02

3.4E-03

2.4E-03

7.2E-030.52 1.00 3.98 10.68 15.21 1.59 0.91 1.05

5.9E-03

1.0E-03

3.8E-03

4.5E-02

8.2E-02

5.0E-03

2.3E-03

9.3E-03

0.46 0.72 1.42 3.48 2.90 1.59 0.91 0.57

3.0E-03

5.1E-04

1.9E-03

4.1E-03

5.9E-032.0E-03

1.2E-03

3.2E-030.45 0.83 0.76 0.86 1.31 0.91 0.58 0.41

4.8E-03

1.1E-03

3.1E-04

2.2E-03

2.8E-03

1.3E-03

9.3E-04

2.5E-032.32 0.60 0.84 0.85 0.85 0.85 0.62 0.46

1.6E-02

5.2E-03

2.9E-03

4.6E-03

6.9E-03

4.1E-03

3.9E-03

1.3E-02

Fig. 1 Before adopting the simultaneous counting method

Fig. 2 After adopting the simultaneous counting method

・Noise such as BG, etc. besides the Am-241 view is reduced
・The counting rate is lowered to 0.5%, measurement time is approx. 16 times (Measurement time in this test /normally 10 minutes)→ It is essential to narrow down the process to be applied (Example: high radiation dose region, etc.)©International Research Institute for Nuclear Decommissioning172

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures: Noise reduction by means of the band-pass filter
Image of scintillation light other than from nitrogen

1

Conventional band-pass filter

Strength (relative value)

Nitrogen light emission

0.8
PEN: Polyethylene
naphthalate(*)

0.6
Cherenkov light ∝ λ-2

PET: Plastic bottle(*)

0.4

0.2

0

240 260 280 300 320 340 360 380 400 420 440
Wavelength (nm)Fig. 1 Spectrum of each type of light emission and filter characteristics of the conventional band-pass filter(*) H.Nakamura et.al., “Evidence of deep-blue photon emission at high efficiency by common plastic”, EPL, 95 (2011)©International Research Institute for Nuclear Decommissioning

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Noise countermeasures: Frequency characteristics of the prospective band-pass filter

Transmission rate (relative value)

1.0

0.8
Past
Option 1
Option 2
Option 3
Option 4
Nitrogenlightemission

0.6

0.4

0.2

0.0
240

290

340
Wavelength (nm)

390

440

Fig. 1 Filter characteristics of the prospective band-pass filter

Each band-pass filter is evaluated for sensitivity after it is applied©International Research Institute for Nuclear Decommissioning174

(b) Technological development of contamination evaluation for sorting solid waste

Noise countermeasures
Evaluation of S/N ratio after the application of the prospective band-pass filter
Table 1 Results of evaluating the S/N for each type of prospective band-pass filter
Band-passfilterAmount of nitrogen lightemission (S) (*)

Amount of Cherenkov light (N) (*)

S/NPriority

Past

1.00

1.00

1.00

5

Option 1

1.15

0.75

1.54

4

Option 2

0.96

0.59

1.62

3

Option 3

0.96

0.57

1.68

1

Option 4

1.08

0.66

1.62

2

*Standardized with numerical values of conventional band-pass filters

The following prospects were seen based on the evaluation of prospective band-pass filter 3 that had its
peak at a wavelength longer than 337nm which was the maximum peak of nitrogen light emission.
・The amount of Cherenkov light reduced to approx 1/2 as compared to the conventional band-pass filter
・S/N improved to 1.7 times.

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Noise countermeasures: Conclusion
Proposed countermeasures were selected based on the characteristics of the Cherenkov light andscintillation light and their effectiveness was evaluated.

Feasibility of distinction using the difference in duration of light emission
・It was verified that the current α camera signal processing circuit has a temporal resolution of μs.
・As a result of studying the applicability of the I/V amplifier having a temporal resolution of ns, it was found that the
number of detected photons has a correlation with the peak height, however, it was confirmed that the relation
between the time taken to detect photons with the Cherenkov light and scintillation light depending on the site
environment and the number of photons detected needs to be sorted out.
Feasibility of distinction using the difference in wavelength distribution
・Several band-pass filters were selected in which the transmission characteristics were shifted to the longer sideof the wavelength for Cherenkov light that is distributed on the shorter side of the wavelength as compared tonitrogen light emission.
・It was confirmed that the S/N ratio improved to approx. 1.7 times by using band-pass filters that bring about themost improvement in S/N.
Noise reduction using the simultaneity of light emission detection
・The composition that provides approx. 4 times the sensitivity was selected as the ancillary detector for using thesimultaneous counting method.
・The counting rate and measurement time were evaluated by means of the verification test using the test
manufactured ancillary detector.
・Noise components such as BG, etc. reduced, but it was found that the counting rate decreased to 0.5%.
・It was found that the required measurement time was approx. 16 times.
・It was found that the process to be used for simultaneous counting needs to be narrowed down.
©International Research Institute for Nuclear Decommissioning176

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Consolidation of issues: Issues in distance measurement
Table 1 Issues based on site investigation
Items

Details of the issues

Issues related to distancemeasurement

Since distance is measured based on the time-offlight of infrared rays (859nm), and unless theinfrared rays return their time-of-flight cannot bemeasured, the conditions are such that distancecannot be measured.

Whether or not distance can be measured withrespect to the material found during the site
investigation, using the angle of incidence of theinfrared rays as the variable, will be evaluated.

Since the results are synthesized based on the
shape characteristics, it may not be possible tosynthesize results in the case of wall surfaces, etc.
where there is absolutely no change in shape.

Whether or not it is possible to synthesize results
pertaining to objects that do not change their shape,by using an image of the outer appearance, andcombining the characteristics of the image as wellwhile synthesizing the results, will be clarified.

Issues related tosynthesizing the distancemeasurement results

Study plan

Table 2 Test conditions for verifying the range in which distance measurement is feasible
Assumed target

Material

Shape

Wall surface, floor
surface

Epoxy coating(Gray)

Flat plate

Covering sheet

Polyethylene

Flat plate

CablesVinyl chloride(Black)Cylindrical

Instrument window

Acrylic

Flat plate

Instrument window

Glass

Flat plate

Water accumulation

Water

-

Typicaldimensions

Shooting distance Shooting angle
0°(Front facing) to
90°
0°(Front facing) to90°

Selected from the αcamera angle of view
Uniformly spaced
and the actualfrom 1000 to 3000mm
0°(Front facing) todimensions of the
90°
target object
0°(Front facing) to
90°
0°(Front facing) to90°©International Research Institute for Nuclear Decommissioning

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Evaluation of the range in which distance measurement is feasible: Test set-up (1)Table 1

Measurement target

L: 1000 to 3000

θInfraredrays

Test conditions for verifying the range in which distance measurement is feasible
Typicaldimensions

Assumed target

Material

Shape

Wall surface, floor
surface

Epoxy coating(Gray)

Flat plate

0°(Front facing) to90°

Covering sheet

Polyethylene

Flat plate

0°(Front facing) to90°

CablesVinyl chloride(Black)Cylindrical

Instrument window

Acrylic

Flat plate

Instrument window

Glass

Flat plate

0°(Front facing) to90°

Water accumulation

Water

-

0°(Front facing) to90°

TOF Camera

Selected from the αcamera angle ofview and the actual
dimensions of thetarget object

Shooting distance

Uniformly spacedfrom 1000mm to
3000mmShooting angle

0°(Front facing) to
90°

1000mm to 3000mm
Measurement target(Epoxy resin)

Rotating shaft
TOF
CameraFig. 1 Test set-up

Unit： mm©International Research Institute for Nuclear Decommissioning178

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Evaluation of the range in which distance measurement is feasible: Evaluation results
Table 1 Results of verification with concrete as the target of measurement and L=1000mm
Material

0° (Front facing)

45°

60°

75°

Epoxy resin(100mm square)

Concrete(Solid, 100mm
square)Concrete(Chipped, 100mm
square)Red portion: Distance measurement results not available

Results of verifying the difference in the surface of concrete(1) The middle section of the front facing part could not be measured because of regular reflection.
Can be imputed.
(2) Epoxy resin cannot be measured at 60°, but it becomes impossible to measure concrete(solid) at 75°.
This is presumed to be an impact of the reflectivity or unevenness of the object to be measured.
(3) Concrete (chipped) can be measured up to 75°.
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Evaluation of the range in which distance measurement is feasible: Evaluation results
Table 1 Results of verification with cable as the target of measurement and L=1000mm
Material
Ethylene propylene
rubber(Black, φ20)

Polyvinyl chloride(Black, φ7)0° (Front facing)

15°

30°

Fig. 1 Distance image
Color is the same as the
background (same distance),and hence it has not beenpossible to measure thedistance of the cable

Polyvinyl chloride(Black, φ3)Fig. 2 To be measured

Scope ofmeasurement(Top: Ethylene propylene rubber (Black, φ20)
Center: Polyvinyl chloride (Black, φ7)
Bottom: Polyvinyl chloride (Black, φ3))

Results of verification with cable as the target of measurement(1) A cable of φ20 cannot be measured(i) A cable of φ20 cannot be measured at 15° or less (data missing)(ii) It is not possible to measure the reflection from the cable in the case of a cable of φ20 at 30° or more
and a cable that is φ7 or smaller. Light reflected from the wall surface in the background is measured in that
case.
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Evaluation of the range in which distance measurement is feasible: Evaluation results
Table 1 Results of verification with transparent materials as the target of measurement and L=1000mm
Material

0° (Front facing)

30°

60°

Fig. 1 Distance image

Acrylic(100mm square)

Color is the same as the
background (same distance), andhence it has not been possible tomeasure the distance of thetargetGlass(100mm square)

Red portion: Distance measurement results not available
Fig. 2 To be measured(Left: acrylic, right: glass)

Results of verification with transparent materials as the target of measurement(1) The middle section at 0° could not be measured. This is presumed to be because of thehigh reflection intensity caused by regular reflection.
(2) As the distance to the background is included in the measurement result of the target
portion, it is not possible at any angle. In the case of transparent material, the reflection
intensity of the target is weak, and it is presumed that the reflection of the transmitted
background is measured.
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Evaluation of the range in which distance measurement is feasible: Evaluation results
Table 1 Results of evaluating the impact of the distance when the covering sheet is the target of measurement
Distance

0° (Front facing)

45°

60°

75°

1m

2m

3m

Red portion: Distance measurement results not available

Results based on the difference in distance (covering sheet as the material, with infrared permeability)(1) At a distance of 1m, at 60° or more, distance cannot be measured.
(2) At a distance of 2m or more, distance cannot be measured at any angle. In the case of transparent
material, the intensity of reflected light is weak, and it is presumed that it gets weaker as the distanceincreases.

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Evaluation of the range in which distance measurement is feasible : Test set-up (2)900mm

Tilt angleθInfrared rays

Water
Target of measurement (100mm square x 20mm)

d: Approx. 1mm to 20mm

TOF
CameraFig. 1 Test set-up with water surface as the target of measurement
Verification of the impact of water accumulation found on the floor surface on distance measurement
・The depth of the water accumulation is assumed to be approx. 1mm to 20mm
・Assuming that a TOF camera is installed on a remotely operated cart, it is fixed at a height of approx. 900mm.
・Since the tilt angle range of the remotely operated cart is 0° to 55° (θ = 35° to 90°), θ is considered to be
30° onwards
・Since the distance becomes greater than 3000mm when θ = 75°, measurement is carried out with θ in therange of 30° to 60°.

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Evaluation of the range in which distance measurement is feasible: Evaluation results
Table 1 Evaluation results when d = approx. 1mm
Material

30°

45°

60°

Table 2 Evaluation results when d = 10mm
Material
30°
45°
60°

Epoxy resin(100mmsquare)Epoxy resin(100mmsquare)Concrete(Solid,100mm
square)Concrete(Solid,100mm
square)Only water

Only water

Table 3 Evaluation results when d = 20mm
Material
30°
45°
60°

Red portion: Distance measurement results not available

Epoxy resin(100mmsquare)Concrete(Solid,100mm
square)Only water

Results based on the difference in depth of water and angle
・Measurement was possible regardless of the depth of water and angle in the scope of this evaluation.
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Evaluation of the range in which distance measurement is feasible: Summary of evaluation results
Table 1 Summary of the results of the test for verifying the range in which distance measurement is feasible
Assumed target

Material

Epoxy coating (Gray)

Wall surface, floor surface

Shooting distanceRange in which shooting
is feasible(Angle)1000mm0° (Front facing) to 45°

2000mm0° (Front facing) to 45°

3000mm0° (Front facing) only

1000mm0° (Front facing) to 60°

2000mm0° (Front facing) to 60°

3000mm0° (Front facing) to 45°

1000mm0° (Front facing) to 75°

2000mm0° (Front facing) to 60°

3000mm0° (Front facing) to 45°

1000mm0° (Front facing) to 60°

2000mm0° (Front facing) to 60°

1000mm0° (Front facing) to 60°

2000mm0° (Front facing) to 60°

1000mm0° (Front facing) to 45°

Polyethylene

2000mmMeasurement not possible

3000mmMeasurement not possible

Glass

1000mm to 3000mm

Measurement not possible

Concrete (Solid)

Concrete (Chipped)

Covering sheet + Epoxy coating(Gray)
Floor surface(covered)

Assumed target

Material

Water +
Epoxy coating (Gray)
Water accumulation
Water + Concrete (Solid)

Depth of water

Range in which shooting
is feasible (Angle)

Approx. 1mm

30° to 60°

10mm

30° to 60°

20mm

30° to 60°

Approx. 1mm

30° to 60°

10mm

30° to 60°

20mm

30° to 60°

Covering sheet + Concrete (Solid)

Covering sheet

Instruments

Wires, cables

Acrylic

1000mm to 3000mm

Measurement not possible

Ethylene propylene rubber (Black,
φ20)

1000mm to 3000mm

Measurement not possible

Polyvinyl chloride (Black, φ7)1000mm to 3000mm

Measurement not possible

Polyvinyl chloride (Black, φ3)1000mm to 3000mm

Measurement not possible

The direction of access for each material shall be based on input from the site
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Evaluation of the synthesis of distance measurement results: Configuration study
The offset (error) is minimized by comparing 3 dimensional point clouds using ICP (Iterative Closest
Point) algorithms.
→ This is repeatedly carried out for all measurement data to minimize overall offset.
・An ICP algorithm involves the following:
1. Every point is mapped to its
corresponding point in point clouds foraligning positions

2. The distance between the mapped
points is calculated

3. The position at which the sum total ofthe distances is the minimum is calculated,
and the point cloud is moved

Here the nearest neighbor is mapped

Point cloud A

L3
L1

L2

L4

L6
L5

Point cloud B(Positions are aligned)

Repeated until convergenceFig. 1 Past algorithms for synthesizing distances
When there is no change in the shape of the target object, or when the shape of the target object
is repeated, the position at which the difference in shape (error) is the minimum is uniquely not
asked.
Countermeasures
Using the color information (texture, pattern) of the images that are simultaneously photographed, thedifference in shape and difference in color information is evaluated, and the point cloud is moved tothe position where this difference is the least.
©International Research Institute for Nuclear Decommissioning186

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Evaluation of the synthesis of distance measurement results: Test set-up
Test conditions for evaluating the synthesis of distance measurement results•
The target object does not have any characteristic shape, it is either a plane surface or the shape is repeated.
•
The target object has texture (pattern)
The synthesis of distance measurement results when images are captured under the above-mentioned conditions is verified.
・Assuming that a TOF camera is installed on a remotely operated cart, the camera is installed at a height of approx. 900mm.
・Assuming the measured distance expected at the time of measurement, the camera is installed 1000mm from the front
facing wall surface.
・The tilt angle is in the range of -60° to 60°, and the pan angle is in the rage of 0° to 30°.

TOF
CameraApprox. 900mm

1000mmTilt angle

Wall surface

Pan angle

Infraredrays
Measurement target (floor)
Floor

Measurement target

Measurement target(wall surface)Fig. 1 Distance synthesis evaluation test set-up and external appearance of the object to be evaluated©International Research Institute for Nuclear Decommissioning187

(b) Technological development of contamination evaluation for sorting solid waste

Evaluation of the synthesis of distance measurement results: Evaluation results

Overall result of synthesis

Results of synthesis using
the conventional method

The misalignment has notbeen corrected with shapeinformation alone

Enlarged wall surface

Synthesis using thismethodSynthesis using theconventional method

Fig. 1 Results of distance synthesis of each target object

The effectiveness of the distance synthesis algorithm that uses color information is verified.
©International Research Institute for Nuclear Decommissioning188

(b) Technological development of contamination evaluation for sorting solid waste

Distance measurement system: Summary<Issues concerning the distance measurement system>
Materials found during the site investigation were subjected to a measurement test. It was found thatthose materials can be measured within the following ranges.
Epoxy resin : 0° to 45°
Concrete coat: 0° to 45°
Solid concrete: 0° to 60°
Chipped concrete: 0° to 75°
Φ3-20mm cable: Cannot be measured
Acrylic, glass: Cannot be measured
Covering sheet: 0° to 45°＠1m, Cannot be measured ＠ 2m or more
Concrete coating under water: Entire area

<Issues related to synthesizing the distance measurement results>
The following was implemented for wall surfaces and floor surfaces in which the shape does notchange or in which the same shape is repeated.
・Modified to an algorithm for evaluating the difference in color information (texture, pattern)
・It was verified that distance synthesis is possible even though it could not be performed for the target
object using conventional algorithms.

©International Research Institute for Nuclear Decommissioning189

(b) Technological development of contamination evaluation for sorting solid waste

Reference material: TOF camera and stereo camera
Table 1 TOF camera and stereo camera comparative chart
Items

TOF Camera

Stereo camera

Measurement
principle

Active distance measurement using the TOF ofinfrared rays

Passive distance measurement using trigonometry

Characteristics

・Measurement is possible if the target object
reflects infrared rays
・Ambient light has little impact (Measurement ispossible even in low intensity light)
・Not suitable for measuring long distances as
infrared rays get diffused and their intensity
reduces・The target object needs to have a pattern on thesurface (Not suitable for uniform planes such as
walls, floor, etc.)
・Sufficient ambient light is required
・The range within which measurement is feasible
can be adjusted based on the distance between
the angle of view and the camera

Measurement target
Infrared rays

The distance is calculated based on the timeof-flight of infrared rays
Fig. 1 Schematic diagram of TOF camera

Measurement target
2 cameras

The distance up to the pixel where
it is perceived as the same point iscalculated using trigonometry
Fig. 2 Schematic diagram of stereo camera

A TOF camera is used for this system considering the range of distances measured by α camera, and the likelihoodof not being able to provide sufficient lighting as dismantling is underway,©International Research Institute for Nuclear Decommissioning190

(b) Technological development of contamination evaluation for sorting solid waste

Consolidation of issues: Issues concerning access
Table 1 Issues based on site investigation results
Items

Details of the issues

Study plan

On-site action plan

Issues related to
movement

Cannot pass due to level difference on thefloor surface.

The location of level difference thathinders movement will be clearlyspecified based on the architectural
drawings of the building andphotographs.*

Temporary slope will be
provided where the level
difference hinders movement.

Cannot pass due to narrow passageways, or
the range within which measurement ispossible gets restricted as the direction of the
cart cannot be changed due to narrowpassageways.

The locations where passageways are
narrow will be identified based onarchitectural drawings of the building andphotographs, and the range within whichmeasurement is possible will be clarified
based on the position of the cart while
passing.

―

Due to its combination with the remotely
operated cart, the α camera is installed at ahigher location. Hence the area in the vicinityof the floor surface where measurement isfeasible is likely to get restricted.

The area where measurement is possible
will be clarified by studying thearrangement when the camera is
combined with the remotely controlled
cart.

―

Issues related tomeasurement*To be studied: Unit 3 - 1F, Unit 2 - 5F

α camera
Pan tilt mechanism
Hoisting mechanism
Image when combined with a remotely operated
cart (decontamination system for elevated location)

Crawler©International Research Institute for Nuclear Decommissioning191

(b) Technological development of contamination evaluation for sorting solid waste

Schematic diagram of the α camera measurement system mounted on the cart
Outline of the α camera including the ancillary detector

Drawing of the α camera including the ancillary detector

300

285

237

Ancillary detector
Conventional α camera

434

Unit： mmFig. 1 Outline of the α camera including theancillary detector

Fig. 2 Detailed drawing of the α camera
including the ancillary detector

Proposed structure when the camera is mounted on the existing crawler will be developed©International Research Institute for Nuclear Decommissioning192

(b) Technological development of contamination evaluation for sorting solid waste

Conception picture of the camera mounted on the cart and the results of consolidating
the issues(a) Position when retracted
Hoisting position:
lower limit
Outrigger: Retracted

Power supply box

Cable drum(b) Outrigger is unfolded
Hoisting position:
lower limit
Outrigger: Unfolded(c) Upper hoisting limit
Hoisting position:
upper limit
Outrigger: Unfolded

Ancillary detector
α camera
Pan-tilt

Pan-tilt
control panel

Height fromthe floor surface

Decontamination
Outrigger crawler

Fig. 1 Conception drawing of mounting the camera on the decontamination crawler<Issues concerning the access cart>
・Avoiding the risk of turning over (center of gravity of the crawler)
・The height of the floor surface exceeds the height required for the α camera (Additional installation ofhoisting mechanism, etc.)
・Power shortage for the pan tilt mechanism (due to increase in weight as a result of mounting the ancillarydetector)
・The decontamination crawler can handle a level difference of approx. 40mm. For larger level differencesslopes need to be additionally installed.

©International Research Institute for Nuclear Decommissioning193

(b) Technological development of contamination evaluation for sorting solid waste

Summary

Results so far
✓

✓
✓
✓
✓
✓

The light emission processes from light emission due to α rays and β rays to signal detection were quantitatively
organized, and a simulation model was developed. Further, the accuracy of simulation was evaluated by verifying
the bare minimum simulation model consisting of the radiation source and α camera built based on the lightemission test.
Nitrogen light emission was evaluated using the basic model and it was confirmed that accuracy was 0.5≦
(calculated value)/(measured value)≦2 (=Factor2).
The physical quantity of each process of nitrogen light emission was quantitatively organized based on thesimulation results.
Temporal discrimination using high speed circuit as a noise countermeasure, simultaneous counting method using
the concurrence of light emission, filters that use the difference in wavelength range, were studied, and theirrespective effectiveness was verified.
Based on site investigation results, the range within which measurement is feasible was clearly specified focusing
on materials that could become a hindrance in the distance measurement system. Also, it was confirmed that the
distance measurement results can be synthesized for surfaces that do not undergo any change in shape (wall, etc.).
The conception drawing of the access cart on which the improved α measurement system is mounted was created.
And, the issues anticipated until application to actual equipment were clearly specified.

Issues, etc.
✓✓✓✓

Development of a simulation model simulating the spatial distribution of β contamination
Development of a simulation model simulating light emission sources such as paint, sheets, etc.
Indication of the α contamination detection performance range (lower α contamination detection limit depending on theon-site environmental conditions) when the air dose rate (+ distribution) and material expected on the site are
simulated.
Applicability assessment of α camera based on the α contamination detection performance range and clear
specification of the incidental on-site environmental conditions©International Research Institute for Nuclear Decommissioning194

2. Project details (No. 2)
b. Establishment of concepts of treatment and disposal, and development ofsafety assessment methods(a) Establishment of selecting advance treatment methods[1] Low-temperature treatment technique[2] Study on approach to assessing applicability for treatment technology(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to study the disposal concepts and to establish
safety assessment methods[2] Development of techniques for assessing impact of affecting substances,

etc. on disposal©International Research Institute for Nuclear Decommissioning

195

(a) Establishment of selecting advance treatment methods◼ Achievements up to FY2020
 Techniques (approaches) that enable multifaceted evaluation of technology were presented by identifying the evaluation axes forevaluating both waste characteristics and treatment technology.
 Of the evaluation axes that were considered in connection with low-temperature treatment technologies (cement solidification and
Alkali Activated Materials (AAM) solidification), where data on characteristics of solidified substances was insufficient, it was
acquired through simulated mixed waste system.
◼ Goal To acquire and evaluate data on high temperature treatment technologies and low temperature treatment technologies, required for
identifying stabilization and solidification technologies expected to be applicable to actual treatment, so as to contribute to theestablishment of techniques for selecting advance treatment methods.
◼

Implementation items and overview
Implementation items① Low
temperature
treatment
technologies② Study on
approachtowards
evaluating
applicability oftreatment
technologies

Overview

Indicators for the achievements of goali. Study on techniques for verifying
the feasibility of low-temperature
treatment solidification

・Study of quantification by means of image determination

ii. Investigation of the change inproperties of solidified material due
to heating, etc.

・Acquisition of data on cyclical conditions, etc. simulating the storage
environment

Data concerning the constituent phase ofsolidified material under storage environmental(cyclic) conditions are investigated andinvestigation results are presented.

iii. Evaluation of the relationshipbetween the inventory of Cs, etc.
and temperature of solidified
material

・Analysis related to the impact during accumulated storage of squareshaped containers

The temperature attained when square shaped
containers are accumulated and stored isevaluated and the method of calculating it ispresented.

iv. Investigation and evaluation offactors affecting long-term
degradation

・Investigation of the method of accelerated testing of degradation
・Data acquisition through accelerated testing and study of applicability

The method of accelerated testing ofdegradation is investigated and the investigation
results are presented.

i. Organization of acquired dataand evaluation axes used forcomparing technologies

Summarization of the data acquired through tests along with existing data,and organization of evaluation axes

Techniques (approach methods) for evaluating
solidification technologies that are likely to be
applicable to various solid waste are presented.

ii. Investigation related to volume
of Cs volatilization during hightemperature treatment and its
control・Acquisition of data related to Cs volatilization control based on tests

Results of investigation of the volatilization
control effect of various events is presented.

Screening technique incorporating the image
determination method is presented.

・Study of applicability while taking the extensive waste generated at the
Fukushima Daiichi Nuclear Power Station into consideration, and in addition,acquisition of required data

©International Research Institute for Nuclear Decommissioning196

(a) Establishment of selecting advance treatment methods- Relationship between the developed techniques and the technology selection process -Prospective applicabletechnologies

Items analyzed(Evaluation axis)

廃棄物Aの特性値

溶融技術Aの技術情報
評価軸Ｂ

溶融

AAM固化

Scope ofapplication

セメント固化

評価軸Ａ

評価軸Ａ

Technical
informationＡ

Selection
process

Analysis of propertiesof wasteＢ

Ｃ Ｄ・・・・・・・・
溶融技術

Identification of prospective applicabletechnologies[Treatment technology selection process]Economic
・・・・・
経済性
efficiency

Versatility
汎用性

Safety安全性・・・・・・
Correlation
政策との
with
policies
関係性
メンテナ
Maintainability
Easeof ンス性運転容Social
社会的acceptability
受容性
operation
・・・・・ Compatibility
易性
互換性
・・・・・

Selection of treatment technology

Techniques to be developed
during this study

Analysis
methods
Inspection
methodSelection
process

Analysis of properties ofwaste

Verification of solidification/composition study

Setting actual treatment conditions[Treatment parameter selection process]©International Research Institute for Nuclear Decommissioning

197

(a) Establishment of selecting advance treatment methods- Flow of the study for establishing the approach (techniques) towards evaluating applicability of treatment technologies ✓✓⚫
⚫

Performance
性能

During low temperature solidification treatment, the relationshipbetween change in properties of solidification material due to thephysical and chemical properties of waste, and the performancerequired of the solidified material needs to be considered.
The required time period differs depending on the performance(Fig. 1)

Ascertaining performance during and after treatment
Confirming changes in the constituent phases focusing on longterm stability (heating, drying (radiation), duration)

Treatment
処理

Storage
保管

Disposal
処分①
Fluidity①流動性②
Hardening rate
②硬化速度
③ Compressive③圧縮強度
strength④ Hydrogen④水素発生速度generation rate
⑤ Nuclide⑤核種溶出率leaching rate

Fig. 1 The time period during which performance factors and properties of major
solidified substances determining the limit values (waste filling rate) determinedby waste properties during the solidification treatment are required
Concluded in FY2020
Investigation of property
information on eachtreatment technology

・Hydrogen generation rate
・Nuclide leaching rate
・Solidification inhibitors

Ascertaining performance
during and after treatment

Concluded in FY2020
Evaluation of the scopeforhigh-temperaturesolidification treatment

Verification methodof solidification

Changes inproperties due toheating and drying

Temperature
conditions leading tochange in properties

Limit values of the waste fillingrate based on basic
characteristics (① to ③)
Limit valuesof waste filling rate based onstability

Limit valuesof waste filling rate

Limit valuesof waste

Performance ofsolidified
substances

Actual facility
configuration• Organization of technical comparison
items (evaluation axis)• Consolidation of information on
restrictions• Consolidation of technical information

Comparison
Evaluation of highest
temperature attained

Changes inproperties withpassage of time

Temperature attained
depending on filling rate

Effect of Cs
volatilization
controlLong-term stability

Effect ofvolatilization control
by operation

Preparation of figures
and tables
Evaluation of applicability oftreatment technologies
Establishment of approaches(techniques)

Waste package
specifications

Information on process,etc.

©International Research Institute for Nuclear Decommissioning198

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the possibility of low-temperature solidification treatment- Details of implementation ◼

Achievements up to FY2020• Techniques for determining the feasibility of low-temperature solidification treatment (cement, AAM) were studied based on properties of various
types of solid waste, and a technique consisting of primary screening (evaluation of possibility of solidification) and secondary screening(composition study and verification of performance of the solidified substances) was developed. Also, quantification was carried out on a trial
basis using image determination during the primary screening, to verify applicability.
• The following issues were identified with respect to on-site application and further refinement of data.
✓ Creation of quantitative criteria for determination during primary screening.
✓ Extensive verification using chemical reagents, etc. assuming various effects on the solidification material, in order to generalize theexamination method.
◼ Objective
To study quantification by means of image determination for the low-temperature solidification treatment technologies of cement solidification and
AAM solidification, and make proposals. Also, to study applicability of the proposed technique taking the extensive waste generated at Fukushima
Daiichi Nuclear Power Station into consideration, and in addition, to acquire and compile the required data.

Details of implementation
Verification of the effect of fluidity improvement due to chemical admixtures• Chemical admixtures (water reducing agents) that are expected to have a fluidity improvement effect with respect tocement and AAM will be investigated, and selected.
• As secondary screening (composition study and verification of performance of solidified substances), the effect ofadding the selected chemical admixtures expected when actual solidification conditions are set up will be studied.
Study of quantification by means of image determination• Addition of inspection items and the method of acquiring data on solidification properties will be studied.
• Quantification by means of image determination equipment will be studied and will be reflected in the screening
technique.

Indicators for the achievements of goal
Screening technique incorporating the image determination method is presented.

©International Research Institute for Nuclear Decommissioning199

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Selection of chemical admixtures •••This study focused on 4 types of typical chemical components from amongst water-reducing agents and high-performance water-reducing
agents (hereinafter described as chemical admixtures) used for improving the fluidity of cement-based material.
From the investigation results, prospective chemical admixtures were selected which had varied chemical compositions that played a rolein water reduction (Table 1).
The name was selected considering experience or track record of past use or information obtained from the manufacturer.

Table 1 Selected chemical admixtures
Type

Molecular structure(Schematic diagram)

Effect in the presence of cement

Effect in the presence of AAM

Name

Lignin sulfonicacid based

Adheres to cement particles, because of which theparticles get negatively charged and an electrostatic
repulsion effect is achieved.
Advantage: Cheap
Disadvantage: Cement hydration is delayed. Water
reduction effect is low.

There are only a few examples of research using AAM.
Some of the research reports indicate water reduction
effect but the views are not consistent.
Tolerant to highly alkaline substances (expected).

Pearllex DP
Pearllex NP

Naphthalene
sulfonic acidbased

Adheres to cement particles, because of which theparticles get negatively charged and an electrostatic
repulsion effect is achieved.
Advantage: It is difficult to delay hydration reaction
Disadvantage: Fluidity reduces easily with to passageof time.

There are comparatively many examples of researchusing AAM.
Some of the research reports indicate water reduction
effect but the views are not consistent.
Tolerant to highly alkaline substances (expected).

Mighty 100

Melamine
sulfonic acidbased

Adheres to cement particles, because of which theparticles get negatively charged and an electrostatic
repulsion effect is achieved.
Advantage: It is difficult to delay hydration reaction
Disadvantage: Fluidity reduces easily with to passageof time.

There are very few examples of research using AAM.
Water reduction effect is unknown.
Tolerant to highly alkaline substances (expected).

SikamentFF86

Polycarboxylic
acid based

Adhered to cement particles. Charging effect is weak,
but steric repulsion effect is achieved for particles
belonging to the same side chain.
Advantage: Fluidity is easily maintained even withpassage of time. Water reduction effect is high.
Disadvantage: Expensive. Cement hydration isdelayed.
*The largest number of use examples are found in thecivil engineering and architectural domain.

There are comparatively many examples of researchusing AAM.
Some of the research reports indicate water reduction
effect but the views are not consistent.
There are concerns about tolerance to highly alkaline
substances.

Sikament
1200N©International Research Institute for Nuclear Decommissioning

200

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Verification of the effect of fluidity improvement due to chemical admixtures ➢ Verification of the effect of fluidity improvement due to chemical admixtures•

Tests were conducted using the selected chemical admixtures (Table 1) to find out the kind of impact that the chemical admixture has on physicalproperties of solidification material (cement, AAM).

•

As there are differences in the usage results of and research on chemical admixtures with respect to cement and AAM, implementation details were
established for each solidification material.
 Cement: As there are ample findings indicating fluidity improvement effect in general, the tests will be conducted focusing on studies related tosimulated mixed waste systems (carbonate slurry: CS, iron co-precipitation slurry: IS).
• Verification of the effect of the selected chemical admixture (base material, simulated mixed waste system)• Acquisition of data pertaining to properties of solidified substances in the simulated mixed waste system when the chemical admixture isadded (Adjustment of the chemical admixture addition rate and the waste filling rate (Fig. 1))
 AAM: As there are only a few research examples concerning the addition of chemical admixtures, whether or not the chemical admixture has
an effect will be investigated.
• Verification of the effect of the selected chemical admixture (base material, simulated mixed waste system)• Investigation of the composition of AAM that has a fluidity improvement effect

Table 1 Selected chemical admixtures
Besides acquisition of data on each solidification material, thechanges in properties of chemical admixtures due to radiation were
investigated as well.
 Investigation of the impact of radiation on chemical admixtures• Investigation of the change in properties of chemical
admixtures due to irradiation• Investigation of the irradiation characteristics of solidified
cement containing chemical admixtures*Definition of waste filling rate

•

Class

Type

Product name
Pearllex NP (OPC)

Lignin sulfonic acid based
Pearllex DP (AAM)
Chemical
admixture

Manufacturer
Nippon Paper
Industries Co., Ltd.
Nippon Paper
Industries Co., Ltd.

Naphthalene sulfonic acidbased

Mighty 100

Kao Chemicals

Melamine sulfonic acid based

SikamentFF86

Sika Japan Ltd.

Polycarboxylic acid based

Sikament 1200N

Sika Japan Ltd.

Waste
Waste filling rate (mass%)
Pulvulerant base material + liquid base material + waste
Waste filling rate 30mass%
Ratio of water and solids 45mass% (ratio of waterand cement 80mass%)

Waste filling rate 50mass%
Ratio of water and solids 45mass% (ratio of water andcement 164mass%)

Water
Water

CementDried and milled simulated waste

Dried and milled simulated waste

CementWhen the ratio of water and solids is kept constant, as the waste filling rate increases, the proportion of cement or AAM powder decreases.

Fig. 1 Definition of waste filling rate
©International Research Institute for Nuclear Decommissioning201

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Verification of the effect of fluidity improvement due to chemical admixtures (Cement) ➢ Impact of chemical admixtures on cement• A paste of base material and simulated mixed waste was made and fluidity data was acquired through a small flow test
(Fig. 1) using a small ring (φ25×50 mm) to understand the fluidity improvement effect of the selected chemical
admixtures.
• There were 5 types of simulated waste, namely, light calcium carbonate, silica powder, heavy calcium carbonate,
simulated carbonate slurry, and simulated iron co-precipitation slurry.
• The obtained small flow values were compiled in terms of the rate of increase in fluidity as against systems that do notcontain chemical admixtures (Table 1).

Fig. 1 Small flow test
Table 1 Results of verifying the fluidity of the selected chemical admixtures
Evaluation based on the rate of increase in small flow values* 1

Type
Base material

Simulated mixed waste system

CementLight calcium
carbonate

Silica powder

Heavy calcium
carbonate

Simulatedcarbonate slurry

Simulated ironco-precipitation
slurryLignin sulfonic acid based

Pearllex NP

◎

〇

〇

〇

〇

〇

Naphthalene sulfonic acidbased

Mighty 100

◎

〇

〇

〇

－

〇

Melamine sulfonic acid based

Sikament FF86

◎

〇

〇

〇

－

〇

Polycarboxylic acid based

Sikament 1200N

◎

◎

◎

◎

－

◎

*1 Explanation of the evaluation symbols ◎: Large improvement (＋30% or more), 〇: Improvement (0 to +30%), －: No improvement✓ All chemical admixtures showed fluidity improvement effect except in simulated carbonate slurry.
✓ Only lignin sulfonic acid based admixture showed an improvement effect in simulated carbonate slurry.
✓ All chemical admixtures showed fluidity improvement effect in simulated iron co-precipitation slurry, but the effect was particularly high withpolycarboxylic acid based admixture.
✓ In the case of cement it was found that there is a chemical admixture having a fluidity improvement effect with respect to all simulated waste.

©International Research Institute for Nuclear Decommissioning202

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Acquisition of data on properties of solidified substances when chemical admixtures are added (Cement) ◼ Data on properties of solidified substances containing chemical admixtures was acquired for simulated carbonate slurry and simulated iron coprecipitation slurry which are simulated secondary wastes generated from contaminated water treatment.

➢ Test method• The chemical admixture that showed the maximum fluidity improvement effect on each simulated slurry was used based on the results of verifying thefluidity improvement effect of chemical admixtures.
 Simulated carbonate slurry: Lignin sulfonic acid based admixture (Pearllex NP)
 Simulated iron co-precipitation slurry: Polycarboxylic acid based admixture (Sikament 1200N)••

Data was acquired by changing the waste filling rate (Table 2) based on the evaluation criteria (Table 1) and the method for testing properties of thesolidified substances.
The amount of chemical admixture to be added was indicated as weight % of the effective amount of admixture for the weight of the pulverulentmaterial (cement + simulated waste) (P).
Table 1 Method of testing the properties of solidified substances and the evaluation criteria
Fluidity

Condensation

Compressive strength

Material segregation (Cement)

TestmethodSmall flow test
Small ring of Φ25 x 50 [mm]Condensation test using thevicat needle

Measurement ofcompressive strength
Small solidified body of 2 x 2x 3 [cm]

Verification of the number of days
until bleeding disappears

Evaluation
criteria

Small flow test
= 66 to 81 [mm] (Cement)= Unit 110 to 130[mm](AAM)

No flash setting, and solidifies
without bleeding within 24
hours

1.47 [N/mm2] or more
(Material age 28 days)

In principle disappears within 24hours
Not observed after 3 days in thecase of waste for which it is difficult
to attain the criteria.

Table 2 Composition of the mixture
Ratio of water and powder (W/P)[mass%]

Waste filling rate [mass %]Simulated carbonate slurry
mixture system

53.8

30 to 40

Simulated iron co-precipitation
slurry mixture system

43.0

20 to 50

©International Research Institute for Nuclear Decommissioning203(a) Establishment of selecting advance treatment [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Acquisition of data on properties of solidified substances when chemical admixtures are added: Simulated carbonate slurry (Cement) ➢ Properties of the simulated mixed waste system containing chemical admixtures• Data was acquired on properties of solidified substances when chemical admixture (lignin sulfonic acid based) that shows
fluidity improvement effect in simulated carbonate slurry (CS) mixture system is added (Fig. 1, Fig. 2 and Fig. 3).

Small ring flow (mm)OPC+CS
120
100

y = 22.53x + 43.17
R2 = 0.992

8060
y = 26.81x + 44.54
R2 = 0.998

Loading rate 30
Loading rate 35
Loading rate 40

20
0
0.00

1.00

2.00

3.00

4.00

5.00

Effective addition rate (P×%)
Fig. 1 Relation between the rate of addition ofchemical admixtures and fluidity✓ Even if the filling rate is increased, by adjusting
the amount of chemical admixture added,
fluidity can be secured.

10

8

OPC+CS7d
28d

AD:Px3.16%8
6

OPC+CS

7

Final setting time (d)

140

40

Condensation

Strength
Compressive strength (N/mm2)

FluidityAD:
Px5.56%

4
2

Loading rate 30
Loading rate 35
Loading rate 40

6
5
4
3
2
1

0

0
30

35

40

Loading rate (mass%)

1

2

3

4

5

Effective addition rate (P×%)

Fig. 2 Relation between filling rate and strength✓ As filling rate increases the strength
decreases✓ Waste can be mixed up to a 40mass%
filling rate.

0

Fig. 3 Relation between the rate of addition of chemical
admixtures and number of days required for hardening✓ The hardening is completed within 24
hours at all levels with the considered
composition of up to 40mass% filling rate
and about Px4% addition rate of chemical
admixtures.

Ease of material segregation✓ Bleeding water disappears within 24 hours at all levels with the considered composition of up to 40mass% filling rate andabout Px4% addition rate of chemical admixtures.
✓ The filling rate (30mass%) of simulated carbonate slurry into solidified cement in the typical composition determined last year can be
improved to up to 40mass% in the scope of investigation this time, by addition of chemical admixtures.

©International Research Institute for Nuclear Decommissioning204

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Acquisition of data on properties of solidified substances when chemical admixtures are added: Simulated iron coprecipitation slurry (Cement) Properties of the simulated mixed waste system containing chemical admixtures• Data was acquired on properties of solidified substances when chemical admixture (polycarboxylic acid based: Sikament 1200N) thatshows fluidity improvement effect in simulated iron co-precipitation slurry (IS) mixture system is added.

y = 290.8x + 37.33
R2 = 0.906

OPC+IS
120
100
80
60
40
20
0
0.0

y = 202.2x + 33.33
R2 = 0.979
Loading rate 20
Loading rate 35
Loading rate 40
Loading rate 45
Loading rate 50
0.2

0.4

y = 187.4x + 20.25
R2 = 0.996
y = 110.9x + 38.25
R2 = 0.822
0.6

0.8

1.0

35

Condensation
9

OPC+IS30

AD:
P×0.17% AD:
P×0.30% AD:
P×0.37%

25
20

7d
28d

8

Final setting time (d)140Compressive strength (N/mm2)

Strength

Fluidity

Small ring flow (mm)➢AD:
P×0.51%15
10
50

OPC+IS

7
6
5
4

Loading rate 20
Loading rate 35
Loading rate 40
Loading rate 45
Loading rate 50

3
2
1

20

35

40

45

0
0.0

50

Loading rate (mass%)

0.2

0.4

0.6

0.8

1.0

Effective addition rate (P×%)

Effective addition rate (P×%)
Fig. 1 Relation between the rate of addition ofchemical admixtures and fluidity

Fig. 2 Relation between filling rate and strength

Fig. 3 Relation between the rate of addition of chemical
admixtures and number of days required for hardening✓ Even if the filling rate is increased, by adjusting
the amount of chemical admixture added,
fluidity can be secured.

✓ As filling rate increases, strength
decreases.
✓ The evaluation criteria of 1.47N/mm 2 ismet even with a filling rate of 50mass%.

✓ With the increase in filling rate the numberof days required for hardening increases as
well.
✓ There is a possibility of being able to reduce
the number of days required for hardening
by using hardening accelerators.

Ease of material segregation✓ Bleeding disappeared within 24 hours at all levels even with a filling rate of 50mass%.
✓ In the case of simulated iron co-precipitation slurry mixture system, even though fluidity can be secured by adding
chemical admixtures, as condensation (number of days required for hardening) gets delayed, the waste filling rate is
not expected to improve.
©International Research Institute for Nuclear Decommissioning205

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Verification of the effect of fluidity improvement due to chemical admixtures (AAM) ➢ Impact of chemical admixtures on AAM• A paste of base material (M: only Metakaolin, MB40: M + slag 40mass%) and a system containing simulated waste was made and fluidity data was
acquired through a small flow test using a small ring (φ25x50mm) to understand the fluidity improvement effect of the selected chemical admixtures.
• There were 2 types of simulated wastes, namely, simulated carbonate slurry, and simulated iron co-precipitation slurry.
• The obtained small flow values are shown in Fig. 1, and their compilation in terms of the rate of increase in fluidity as against systems that do notinclude chemical admixtures is shown in Table 1.
Polycarboxylic acidbased admixture:
Decrease in fluidity

110

M
Small ring flow (mm)100

Table 1 Results of verifying the fluidity of the selected chemical admixtures
Evaluation based on the rate of increase in flow values with φ25 [mm] ring *1

Type
90

AAM (M)80

Base metal

Simulatedcarbonate
Slurry70
60
50
40 No chemical Lignin
Naphthalene Melamine Polycarboxylic
減水剤なし
LS
NS
MS
PC
admixtures sulfonic acid sulfonic acid sulfonic acid
based

110

MB40
Small ring flow (mm)

100

based

acid based

based

Polycarboxylic acidbased admixture:
Decrease in fluidity90

AAM (MB40)
Simulated
Iron coprecipitation
SlurryBase metal

Simulatedcarbonate
SlurrySimulated
Iron coprecipitation
SlurryLignin sulfonic acid based

Pearllex DP

－

〇

－

－

〇

－

Naphthalene sulfonic acid based

Mighty 100

－

◎

－

－

◎

－

Melamine sulfonic acid based

Sikament FF86

－

◎

－

－

◎

－

#1 Explanation of evaluation symbols ◎: Large improvement (+30% or more), ○: Improvement (0to+30%), －:improvement,
Reduced fluidity
Sikament×:
1200N
x
x
x
x
x
Polycarboxylic
acidNo
based

x

80

*1 Explanation of evaluation symbols ◎: Large improvement (+30% or more), 〇: Improvement (0 to +30%), －: No improvement, ×: Reduced fluidity70(M: Only Metakaolin, MB40: Metakaolin + slag 40mass%)

60
50
40

LigninNaphthalene
Polycarboxylic
NochemicalMelamine
減水剤なし
LS
NS
MS
PC
admixtures sulfonic acid sulfonic acid sulfonic acid acid based
based
based
based

Fig. 1 Small flow values of basematerials in which chemical
admixtures were added(Top: AAM (M), Bottom: AAM(MB40)✓ Fluidity reduced under all conditions in the case of polycarboxylic acid based admixture.
✓ Improvement in fluidity due to chemical admixtures was not seen in the case of base material system.
✓ Fluidity improvement was seen in simulated carbonate slurry mixture system in the case of all
admixtures except polycarboxylic acid based admixture.
✓ None of the chemical admixtures showed a fluidity improvement effect in simulated iron coprecipitation slurry mixture system.
✓ None of the chemical admixtures showed an effect throughout all cases under AAM, but it was foundthat the effect is seen depending on the waste mixed.

©International Research Institute for Nuclear Decommissioning206

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Verification of the effect of fluidity improvement due to chemical admixtures (AAM) ➢ Background of verification• In the case of AAM, the chemical admixture that didn’t have an effect on the base material, showed different tendencies such as some fluidity
improvement effect in the simulated mixed waste system.
• It is believed that this difference in fluidity improvement effect depends on the composition of the solution, and in the case of AAM, the compositionof the alkaline solution is expected to have an impact on the effect of the chemical admixture.
• Hence, the effect of chemical admixture when the concentration of Na and Si in the alkaline mixing water is changed in a mixed waste system wasinvestigated.
➢ Testing conditions• AAM (M＋CS (filling rate 30%)), AAM (M＋IS (filling rate 20%)), AAM (MB40＋CS (filling rate 30%)), AAM (MB40＋IS (filling rate 20%))• Composition of the alkaline mixing water: Concentration of Na and Si in the solution is changed• The changes in small flow values when the composition of the alkaline solution is changed are shown in Fig. 1.

Lignin sulfonic acid based

Naphthalene sulfonic acid based
14

10

8 Percent change insmall flow values
6420Area in which it is not feasible due to
水ガラスの組成上、実現不能領域the compositionof water glass

0

1

2

3

Si concentration (mol/L)

Composition of the solution in the
typical composition of AAM (M+CS)
identified during studies until FY2020(Si: 2mol/L, Na: 5mol/L)

4

10

8 Percent change insmall flow values
6
4
2
0

5

～-20%
-20%～-10%-10%～10%
10%～20%20%～

12Area in which it is not feasible due to
水ガラスの組成上、実現不能領域the compositionof water glass

0

1

2

3

Si concentration (mol/L)

4

Polycarboxylic acid based

Melamine sulfonic acid based
14

Na concentration (mol/L)～-20%-20%～-10%-10%～10%
10%～20%20%～

12

Na concentration (mol/L)

Na concentration (mol/L)

Small
flow
percentchange 14

5Cannotbe mixed
混練不可～-20%-20%～-10%-10%～10%
10%～20%
20%～

12
10
8

Flash
setting
急結

There was no
fluidity
improvement
effect in any of thecompositions

Percent change in

6 small flow values
420Area in which it is not feasible due to
水ガラスの組成上、実現不能領域the compositionof water glass

0

1

2

3

4

5

Si concentration (mol/L)

Fig. 1 Difference in the water reduction effect depending on the composition of mixing
water and the type of chemical admixture(Legend: Percent change in small flow values, AAM (M+CS))

With a decrease in Na
concentration, the effect of thechemical admixture increases.

➢ Test results✓ Water reduction effect largely changed depending on the composition of mixing water in the case of all chemical admixtures.
✓ Polycarboxylic acid based admixtures did not have a water reduction effect in the scope of any of the combinations.
✓ It was found that with a decrease in the concentration of Na in the alkaline mixing water, water reduction effect increases. In particular, it was
found that naphthalene sulfonic acid based admixtures and melamine sulfonic acid based admixtures tend to have a larger water reduction
effect.
✓ When chemical admixtures are added to AAM, it is necessary to adjust the composition once again for each chemical admixture used.
©International Research Institute for Nuclear Decommissioning207

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study on the scope of application of solidification treatment when chemical admixtures are added (Cement) ◼ Data acquired on the properties of solidified cement mixed in simulated waste in which chemical admixtures are added, wasreflected in the triangular diagram presented in FY2020. (Portion enclosed by the ○ was added).
➢ Solidification characteristics of solidified cement in the simulated carbonate slurry + lignin sulfonic acid based admixture(Pearllex NP)

as
s%
)

(m

ial

as
s%
)

(m

ial

ate
r

Ba
se
m

as
s%
)

ate
ria
l (m

ate
r

Ba
se
m

)
s%
as

ate
r

Ba
se
m

as
s%
)

as
s%
)

(m

ial

ate
r

Ba
se
m

(m

Comprehensive
evaluation

20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

Ease of material
segregation

Waste (mass%)

Strength

)
s%
as

ter
Wa

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

不可possible
Not
Partially
一部不可possible
Possible可Possible
可 (+AD)(+AD)

28d compressive strength (N/mm²)～ 1.47
1.47 ～ 5
5～
5 ～ (+AD)

(m

0

Ba
se
m

Bleeding disappearance time (hr.) Base material:Cement
72 ～Waste:CS
024 ～ 72
100
～ 24
10
90
～ 24 (+AD)

Waste (mass%)

Table 1 Scope of feasibility of solidification treatment whenchemical admixtures are added (Cement + CS system) -Scope of application of solidified substancesthat do not contain chemical admixtures

Base material:Cement
Waste:CS

Waste (mass%)

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

ter
Wa

Waste (mass%)

0

)
s%
as

)
s%
as

)
s%
as

Waste (mass%)

Hardeningtime

Base material:Cement
Waste:CS

(m

(m

(m

20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

Final setting time (hr.)
72 ～24 ～ 48～ 24～ 24 (+AD)

ter
Wa

ter
Wa

Fluidity

ter
Wa

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

φ25 small ring flow (mm) Base material:Cement～ 65
Waste:CS
65 ～ 80
0
100
80 ～
10
90
80 ～ (+AD)

(m

0

ial

Base material:Cement
Waste:CS

possible
Partially possible
Not possible

Comprehensive
evaluation

Withoutchemical
admixturesWithchemical
admixtures

Simulatedcarbonateslurry15 to 40
[mass%]

15 to 45
[mass%]

Cement

30 to 55
[mass%]

20 to 55
[mass%]

Water

30 to 45
[mass%]

30 to 45
[mass%]

Fig. 1 Scope of feasibility of solidification treatment of solidified cement in the simulated carbonate slurry when chemical admixture(Pearllex NP) is added✓ The scope of feasibility of solidification of the carbonate slurry was increased by adding the chemical admixture.
©International Research Institute for Nuclear Decommissioning208

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study on the scope of application of solidification treatment when chemical admixtures are added (Cement) -➢ Solidification characteristics of solidified cement in simulated iron co-precipitation slurry + polycarboxylic acid based admixture(Sikament 1200N)

(m
as
s%
)

28d compressive strength (N/mm²)～ 1.47～ 1.47 (+AD)
1.47 ～ 5
1.47 ～ 5 (+AD)
5～
5 ～ (+AD)ater

Ease of material
segregation

ial

(m

as
s%
)

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

Ba
se
m

as
s%
)

(m

ial

ial

ate
r

総合評価
Comprehensive
evaluation
不可possible
Not
Not
possible
不可 (+AD) (+AD)
一部不可possible
Partially
一部不可possible(+AD) (+AD)
Partially
可Possible
Possible
可 (+AD) (+AD)

(m

)
s%
as

Ba
se
ma
ter

ate
r

as
s%
)

(m

ter
Wa

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 1501 160 70 80 90 100

ial

Ba
se
m

as
s%
)

(m

ial

ate
r

Ba
se
m

0

0

Strength

Waste (mass%)

Waste (mass%)

Scope of application of solidified substancesthat do not contain chemical admixtures

Base material:Cement
Waste:IS

Base material:Cement
Waste:IS

)
s%
as
(m

Waste (mass%)

Waste (mass%)

Hardening
time

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

Bleeding disappearance time (hr.)
96 ～96 ～ (+AD)
72 ～ 96
72 ～ 96 (+AD)～ 72～ 72 (+AD)

)
s%
as
(m

)
s%
as
(m

)
s%
as
(m

40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 50 60 70 80 90 100

0

ter
Wa

ter
Wa

Fluidity

Base material:Cement
Waste:IS

ter
Wa

100
10
90
20
80
30
70
40
60
50
50
60
40
70
30
80
20
90
10
100
0
0 10 20 30 40 76
50 60 70 80 90 100

Final setting time (hr.)
72 ～72～ (+AD)
24 ～ 72
24 ～ 72 (+AD)～ 24～ 24 (+AD)

Ba
se
m

0

ter
Wa

φ25 small ring flow (mm) Base material:Cement
Waste:IS～ 65
0～ 65 (+AD)
100
10
65 ～ 8090
65 ～ 80 (+AD)
20
80
80 ～
30
70
80 ～ (+AD)

Base material:Cement
Waste:IS

Possible
Partially possible
Not possible

Table 1 Scope of feasibility of solidification treatment
when chemical admixtures are added (Cement + IS
system) Withoutchemical
admixturesWithchemical
admixtures

Simulatediron coprecipitation
slurry15 to 25
[mass%]

Same as on
the left

Cement45 to 50
[mass%]

Same as on
the left

Water

30 to 35
[mass%]

Same as on
the left

Comprehensive
evaluation

Comprehensive
evaluation

Waste (mass%)

Fig. 1 Scope of feasibility of solidification treatment of solidified cement in the simulated iron coprecipitation slurry when chemical admixture (Sikament 1200N) is added✓ When fluidity improves, condensation gets delayed. Hence ultimately the scope of application did not increase.
©International Research Institute for Nuclear Decommissioning209

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Irradiation test: Irradiation characteristics of the chemical admixtures ➢ Investigation of the change in properties of chemical admixtures due to irradiation
•
•
•
•
•

Changes in the properties of chemical components that show water reduction effect of the chemical admixture, due to radiation, were confirmed
through γ irradiation test.
Samples of 5 types of chemical admixtures indicated in Table 1 were irradiated, they were diluted so that the effective amount of admixture was
20mass% and the moisture content was about 80mass%, and placed in a glass vial.
γ irradiation was carried out under the conditions mentioned in Table 2.
The samples before and after irradiation were compared by means of FT-IR analysis (Attenuated Total Reflection method) (Fig. 1).
After irradiation, the hydrogen gas inside the container was analyzed, and the G value was calculated (Fig. 2).

Table 1 Material to be irradiatedChemical admixture

Table 2 Irradiation conditions
Form of sample
Powder (aqueous
solution)
Powder (aqueous
solution)
Powder (aqueous
solution)

1

Lignin sulfonic acid based (Pearllex NP)

2

Lignin sulfonic acid based (Pearllex DP)

3Naphthalene sulfonic acid based (Mighty 100)

4

Melamine sulfonic acid based (Sikament FF86)

Liquid5

Polycarboxylic acid based (Sikament 1200N)

LiquidTest details

γ irradiation of chemical admixtures

Testing site

Takasaki Advanced Radiation ResearchInstitute

Radiation source
Dose rate
Integral dose
(Planned)

Co-60
Approx. 3 [kGy/h]
3 [kGy]Sikament
シーカメン1200N
ト1200NUnirradiated
非照射Sikament
3kGy
シーカメン122N
ト1200N
3ｋＧｙ

0.6✓ During FT-IR, no differences were seen in the shape of peaks before
and after irradiation, in the case of all chemical admixtures.
✓ Except Sikament 1200N which is a polycarboxylic acid based
0.3
admixture, the G value of all chemical admixtures was lower than the0.2
G value of water (0.45).
✓ The G value of Sikament 1200N was high, but it is presumed based
0.1
on the results of FT-IR that this was not due to breakdown of
0.0
パール Pearllex
パール Sikament
シーカ Sikament
シーカ
マイティ
Mighty Pearllex
admixtures.
1200N
レックス
レックス
メント
メント
FF86
100
DP
NP
100
DP
NP
FF86
1200N
✓ Changes in properties of chemical admixtures due to radiation were
Fig. 2 G value of chemical admixtures
not found.

4000 3600 3200 2800 2400 2000 1600 1200 800 400

Wavenumber (cm-1)
パールレッ
クスNP
非照射
PearllexNP
Unirradiated
パールレッ
クスNP
3ｋGy
PearllexNP
3kGy

G value
G値 (H2/100eV)

0.5

0.4

4000 3600 3200 2800 2400 2000 1600 1200 800 400

Wavenumber (cm-1)

Fig. 1 Results of FT- IR analysis of chemical admixtures©International Research Institute for Nuclear Decommissioning

210

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Irradiation test: Irradiation characteristics of the solidified cement containing chemical admixtures ➢ Irradiation characteristics of solidified cement containing chemical admixtures
•

•
•
•

Investigation was carried out to check whether the properties of the chemical admixtures in solidified substances change due to the impact ofradiation leading to leaching of organic components.
Solidified cement indicated in Table 1 was used as the sample for irradiation.
γ irradiation was carried out under the conditions mentioned in Table 2.
Leaching test was conducted for the samples before and after irradiation. Further, the changes in organic components in the solid phase after
the leaching test were verified by CHN analysis.

Table 2 Irradiation conditions

Table 1 Material to be irradiatedSolidific
ationmaterial

W/C[mass%]
45

Cement25

Small flow
values[mm]

66-81

Chemical admixture
Sikament 1200N

CuringperiodSamplesize

7 daysΦ10×20[mm]

Test details

γ irradiation of chemical
admixtures

Testing site

Takasaki Advanced Radiation
Research Institute

Radiation source

Dose rate• Results of the CHN analysis of the solid phase are indicated in Fig. 1.

Containment rate (%)

3.0

OPC

N
C
H

2.5
2.0
1.5
1.0
0.5
0.0
W/C=0.45
Unirradiated
非照射

W/C=0.45
3kGy

W/C=0.25
Polycarboxylic
ポリカルボンacid酸系
based

Integral dose
(Planned)

Co-60

Approx. 3 [kGy/h]
3 [kGy]✓ During the leaching test, there was no leaching of organic matter before andafter irradiation.
✓ Data showing an increase in the proportion of carbon in all samples after
irradiation was obtained, but the leaching characteristics were not affected.
✓ The possibility of uptake of carbon from outside (for example the gaseouscomponent in the container, etc.) can be considered.
✓ Leaching of the chemical admixture component due to radiation could not be
verified, and hence it is believed that it had only a small effect on leaching
characteristics.

W/C=0.25
Polycarboxylic
ポリカルボンacid based
酸系
3kGy
3kGy

Fig. 1 Results of the CHN analysis of the solid phase after the leaching test
©International Research Institute for Nuclear Decommissioning211

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Determination of fluidity improvement effect based on primary screening ➢ Determination of fluidity improvement effect based on primary screening• If the effect of chemical admixtures can be determined during primary screening (determined using a combination of waste andinspection liquid), the effect can be determined more easily than performing secondary screening (determined using a paste
made by mixing waste and solidification material).
• It was verified that the effect of the chemical admixture can be determined during primary screening, by measuring thedispersibility of waste using inspection liquid (cement equilibrium water, 40g) in which chemical admixture (0.1g) is added.
• The results of primary screening (verification of dispersibility of waste and inspection liquid with a centrifuge tube) (Table 1))
and the results of secondary screening (fluidity verification test using the paste) (Table 2) were compared.
 Cement
Table 1 Results of dispersibility verification by means of primary screening method
Results of primary screening (solidification feasibility evaluation)(Evaluation based on the state of sedimentation of particles*1)

Type of simulated
waste

Table 2 Results of fluidity verification by means of secondary screening method
Evaluation based on the rate of increase in small flow values *2

Type

SilicapowderHeavy
calciumcarbonate

SimulatedcarbonateslurrySimulated
Iron
coprecipitationslurryInspection liquid(Basis of comparison:
Cement equilibrium water)

Cementfor researchLightcalciumcarbonate

Cement equilibrium water＋ Pearllex NP

〇

〇

〇

〇

－

〇

Cement equilibrium water＋Sikament FF86

〇

◎

◎

◎

－

〇

Cement equilibrium water＋Mighty 100

〇

◎

◎

◎

－

〇

Cement equilibrium water＋Sikament 1200N

〇

◎

◎

◎

－

〇

Type of simulated
waste
CementLight
calciumcarbonate

SilicapowderHeavy
calciumcarbonate

SimulatedcarbonateslurrySimulated
Iron
coprecipitationslurryPearllex NP

◎

〇

〇

〇

〇

〇

Sikament FF86

◎

〇

〇

〇

－

〇

Mighty 100

◎

〇

〇

〇

－

〇

Sikament 1200N

◎

◎

◎

◎

－

◎

Type of chemical
admixture*2 Explanation of the evaluation symbols ◎: Large improvement (＋30% or more), 〇:Improvement (0 to +30%), －: No improvement*1 Explanation of the evaluation symbols ◎: Dispersibility improvement (All particles), 〇:Dispersibility improvement (Only fine particles), －: No change✓ The dispersibility of particles improved due to addition of chemical admixture with cement alone and simulated waste (total 5 types) except simulated carbonate
slurry (Table 1).
✓ The fluidity improvement effect was consistent for all simulated waste except simulated carbonate slurry (Table 2).

 AAM✓ The trend seen in the primary screening (change in dispersibility) and the trend seen in secondary screening (change in fluidity) were not consistent.
✓ In the case of cement, the results are consistent for all simulated waste other than simulated carbonate slurry, and hence the effect can be
determined by means of the screening method even for systems containing chemical admixtures.
©International Research Institute for Nuclear Decommissioning212

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study on addition of inspection items and the method of acquiring data on solidification characteristics: Condensation verification test
➢ Study of the method of verifying condensation of solidified substances by means of the untrasonic method

Table 1 Compounding conditions•

Samplesolidified
substance

Fillingfactor[mass%]

W/P[mass%]

Cement-30

Cement + CS

30

53.8

Cement + IS

20

48

•

20

Solidified
substance

Hardening

Paste

18
16

Vicat
needle12
OPC
OPC+CS
OPC+IS

10
8

0

2

4

6

8 10 12 14 16 18 20 22 24

Time (hr.)

Hardening

14

End

Start

Ultrasonic propagation time (µs)•

Condensation, which is an evaluation item of solidification characteristics, is supposed to be tested andevaluated using the vicat needle in secondary screening, but from the perspective of reducing exposure,etc. it is desirable to complete the operation in a sealed system such as the centrifuge tube.
The method of obtaining condensation from the outer surface of the centrifuge tube by means of ultrasonicwaves was studied.
Cement paste (Table 1) was mixed, and the Ultrasonic Propagation Time (UPT) from right after it was filled
in a centrifuge tube up to 24 hours was obtained.

UltrasonicmethodFig. 2 Image illustrating the relation between the change incharacteristics at the time of cement hardening and the acquired data

Fig. 1 Relation between elapsed time and the ultrasonicpropagation time

✓ All samples indicated that there was an inflection point at which the ultrasonic propagation time largely decreases (Fig. 1).
✓ As the propagation velocity of ultrasonic waves changes in the paste and in solidified substances, it is believed that it was possible to ascertain
how the paste changes to solidified substance with passage of time.
✓ It is believed that the start and end time of condensation obtained using the vicat needle, and the condition of the solidification material thatindicates the propagation time obtained by means of the ultrasonic method is as shown in Figure 2.
✓ It is thus implied that the time required for the paste to solidify can be obtained without making contact with the sample by inserting the vicat
needle.
©International Research Institute for Nuclear Decommissioning213

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study on addition of inspection items and the method of acquiring data on solidification characteristics: American Nuclear Society leaching test -➢ Study of the method of evaluating leaching of solidified substances• While evaluating solidification characteristics since understanding leaching is important as well besides the current evaluation criteria, the method
for easily evaluating leaching was studied.
• From the perspective of reducing exposure, etc. it is desirable to complete the operation in a sealed system such as the centrifuge tube.
Table 1 Comparison of the testing conditions of the ANS
leaching test and the simple test

• The applicability of the centrifuge tube method was studied by conducting a leaching
test using a 50ml centrifuge tube (hereinafter, centrifuge tube method), and theresults were compared with the results of the ANS leaching test (hereinafter, ANS
method) (Table 1).
• About 15 cc sample was filled in a centrifuge tube and hardened, and it was sealed
and cured up to a material age of 28 days. Thereafter, 30cc ultrapure water waspoured on the sample in the centrifuge tube, it was allowed to stand, and thensolution was collected and refilled. The element concentration of the collected
solution was measured.
102
ANS centrifuge
遠沈管法 tube method
OPC+CS
Sr
Sr
101
Sn
Sn
100
Cs
Cs
Ce
Ce
10-1

60
40
20
0

100

10-2

ANS centrifuge
遠沈管法 tube method
Sr
Sr
Sn
Sn
Cs
Cs
Ce
Ce

10-3
10-4

0

24

48

72

Immersion period (hr.)

96

120

10-5

0

24

48

72

96

Immersion period (hr.)

Fig. 1 Comparison of the leaching rate obtained by means of the
ANS method and the centrifuge tube method Cement + CS
(Left: Real axis, right: Logarithmic axis)

120

Quantity of solution(cm3)/ Sample contact
surface ratio

Leaching period[h]Quantity ofsolution collected

ANS leaching test
(Implemented in
FY2020)10

2, 7, 24, 48,
72, 96, 120

Around 1.2 [L] /round

Centrifuge tube
method(Implemented this
year)

2.62

2, 7, 24, 48,
72, 96, 120

30 [cc] / round

102
M+CS
ANS centrifuge
遠沈管法 tube method
101
Sr
Sr
Sn
Sn
100
Cs
Cs
Ce
Ce
10-1

M+CS

80

Leaching rate (%)

80

Leaching rate (%)

Leaching rate (%)

OPC+CS

Leaching rate (%)

100

Testing conditions

60
40
20
0

10-2
ANS centrifuge
遠沈管法 tube method
Sr
Sr
Sn
Sn
Cs
Cs
Ce
Ce

10-3
10-4

0

24

48

72

Immersion period (hr.)

96

120

10-5

0

24

48

72

96

120

Immersion period (hr.)

Fig. 2 Comparison of the leaching rate obtained by means of the
ANS method and the centrifuge tube method AAM + CS)(Left: Real axis, right: Logarithmic axis)✓ The easily leachable and not so easily leachable elements obtained as a result of the centrifuge tube method for both the cement system as well as the AAM systemwere consistent with those obtained by means of the ANS method (Fig. 1, Fig. 2).
✓ The relationship of magnitude of the leaching rate for each element was consistent in the cement system. In the AAM system, the relationship of magnitude of theleaching rate for the not so easily leachable elements was not consistent.
✓ The centrifuge tube method is likely to be applicable for determining the ease of leaching of the elements.
✓ If the relationship of the ANS method and the wet surface area can be controlled, there is the possibility of being able to evaluate leaching (obtain the L value) using
the centrifuge tube.
✓ It thus implied that there is the possibility of being able to ascertain leaching behavior of elements by means of a small scale leaching test (centrifuge tube method).
©International Research Institute for Nuclear Decommissioning214

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study of quantification using the image determination test equipment ➢ Study of quantification by means of the image determination test equipment
•
•

•
•

In order to establish the determination criteria for the verification technique, each determination criteria is quantified using the image determinationequipment (Fig. 1).
For establishing the determination criteria, data needs to be accumulated on waste indicating both “appropriateness” and “inappropriateness” with respect
to the determination criteria.
So far, it has been found that solidifiable simulated waste can be determined by means of the image determination equipment similar to visual inspection.
In this study, data was obtained by means of the image determination equipment on the 3 types of simulated waste (Table 1) that were determined as
inappropriate as a result of visual determination, from among 12 types of simulated waste studied during the FY2020 investigation.

Shutter box for shielding light

Table 1 3 types of simulated waste determined as inappropriate as a result of visual determination

Analysis ofmeasured values

Thermalimaging

VisibleCentrifuge
imaging
tube
Light source
Light source

Simulated waste

Specifications of the image
determination test equipment(1) Visible imaging camera(Light source at 1 locations, thefront and the back)(2) Thermal imaging camera(3) Load cell

Properties determined as “inappropriate”

Inspection liquid

Aluminum powder

Foaming after dripping of inspection liquid

Simulated cement
kneading waterMontmorillonite

Swelling after dripping of inspection liquid

Pure water

Magnesium sulfate

Sudden heating after dripping of inspection liquid

Pure water

Fig. 1 Configuration, specifications and external appearance of the image determination test equipment➢ Acquisition of images indicating inappropriate properties✓ 3 types of simulated wastes were loaded on to the equipment. The obtained images are shown Table 2.

Table 2 Images of each simulated waste
Aluminum powder +Simulated cement
kneading waterMontmorillonite +Pure water

Magnesium sulfate +Pure water
Surrounding
temperature
measurement
range
Sampletemperature
measurement
range

・Confirmation of foaming：〇

・Swelling: 〇・Heating： 〇・Acquisition oftemperature：〇 Aluminum powder✓ Foaming was confirmed from visible imaging
 Montmorillonite✓ The state of slight swelling was confirmed from visible
imaging Magnesium sulfate✓ The state of heating was confirmed by means of thermal
imaging✓ Sudden heating after dripping of inspection liquid was
obtained as numerical values along with passage of time.

✓ Inappropriate properties can be determined similar to visualinspection using the image determination equipment.
©International Research Institute for Nuclear Decommissioning215

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment- Study of quantification using the image determination test equipment ➢ Study on quantification of inappropriate properties••

It was verified whether or not inappropriate properties can be quantified with respect to foaming and swelling.
It was verified whether the images where foaming can be seen can be adjusted using an analytical software so that selecting the portion with thefoaming and obtaining the area of that section can be automated (Fig. 1).

Portionwith actual
foaming

Portion in which
there is no foamingis included

Perceived as smaller than
the actual area withfoaming

Fig. 1 Perception of foaming by image adjustment (area of the red portion to be calculated)✓ The portion with foaming was perceived as smaller than the size of the portion with actual foaming. Hence the area of only the portion withfoaming could not be calculated.
✓ The portion with the shadow of the centrifuge tube, etc. was perceived as an area with the same contrast and thus sections besides the portion
with foaming got included as well.
✓ In order to automate the process of obtaining the amount of foaming from the image, the study needs to include conditions during shooting as well.
•

The determination criteria of heating was established on a trial basis focusing on the values of rise in temperature (Δ Temperature =[Temperature of sample] ー [Surrounding temperature])
Table 1 Maximum Δ temperature of simulated waste✓ Besides simulated waste, the Δ temperature of the kneaded cement paste wasobtained by means of the image determination equipment, and it was compared with
the Δ temperature of the simulated waste that was obtained in a similar manner bymeans of the image determination equipment (Table 1).
✓ As long as the range of Δ temperature of simulated waste does not exceed the Δtemperature (1.2℃) of the cement base material, it can be determined as solidifiable.
✓ The Δ temperature of simulated iron co-precipitation slurry is 3.4℃ which is higher
than the cement paste, but it has been confirmed to be solidifiable in the studies
conducted up to last year.
✓ The determination criteria of heating is estimated to be between 3.4 and 24.1℃.
✓ In order to derive a clear determination criteria, data needs to be accumulated bymeans of tests using samples that offer a variety of heat generation.

Maximum Δ
temperature [℃]

Determination ofsolidification

Cement1.2-

Magnesium sulfate

24.1

Inappropriate

Aluminum powder

1.2

Appropriate

Montmorillonite

1.3

Appropriate

Heavy calcium carbonate

1.7

Appropriate

Simulated carbonate slurry

1.3

Appropriate

Simulated iron coprecipitation slurry

3.4

Appropriate©International Research Institute for Nuclear Decommissioning216

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniquei. Study on techniques for verifying the feasibility of low-temperature solidification treatment

Results so far✓ Chemical admixtures were investigated and 5 types of chemical admixtures that are likely to have aneffect on cement and AAM were identified.
✓ The effects that the identified chemical admixtures have on change in characteristics of solidified
substances such as fluidity, condensation, strength and ease of material segregation were verified anddata was obtained.
✓ Cement solidification: Lignin sulfonic acid for simulated carbonate slurry and polycarboxylic acid for
simulated iron co-precipitation slurry were selected as the chemical admixtures with a powerfulimprovement effect. It became possible to improve the maximum filling rate of simulated carbonate
slurry from 40 mass% to 45 mass% by adding the chemical admixture.
✓ AAM solidification: Chemical admixtures did not have an effect on AAM In order to use chemical
admixtures for AAM, the composition needs to be adjusted once again.
✓ As an improvement in the screening technique, the evaluation technique that enables verification ofcondensation without making contact with the sample by using a centrifuge tube is expected to havepotential. It was found that there is the possibility of being able to ascertain leaching behavior ofelements by means of a leaching test using a centrifuge tube.

Issues, etc.
✓ The simple measurement method that can be used as a verification technique needs to studied
further.
✓ In order to generalize the verification techniques, verifications assuming various solid wastes
need to be carried out.
✓ In order to develop a technique for quantification of foaming and swelling based on images andto derive a criteria for determining feasibility of solidification, data needs to be accumulated bymeans of tests using samples that offer a variety of heat generation.
©International Research Institute for Nuclear Decommissioning217

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Details of implementation ◼ Achievements up to FY2020• Investigations and tests related to the change in properties of solidified substances due to heating and drying wereconducted, that made it clear that heating and drying contributed to the decrease in the strength of solidified substances.
• The following issue was identified with respect to on-site application and further refinement of data.
✓ Acquisition of data on change in properties of solidified substances in actual storage environment, and evaluation of
the attained temperature.
◼ GoalTo conduct tests simulating heating, drying, etc. assuming the temperature and humidity in actual environment forinvestigating the impact that storage environment has on the change in properties of solidified substances, in order toinvestigate the impact on solidified substances during the period from treatment to storage.

Details of implementation

Acquisition of data on cyclic conditions, etc. simulating the storage environment• Setting of conditions to which the solidified substances are exposed assuming the temperature and humidity in thestorage environment• Evaluation of change in properties of solidified substances after introducing cyclic conditions

Indicators for the achievements of goal

Data concerning the constituent phase of solidified material under storage environmental (cyclic)
conditions are studied and investigation results are presented.

©International Research Institute for Nuclear Decommissioning218

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test conditions 1. Item to be tested and method of exposure ➢ Test conditions• Item to be tested: Prepared solidified substances (Table 1)• Method of preparing the solidified substances: A 1 cm space was created at the top of a cylindrical mold (φ5 x 10cm) (Fig. 1)• Exposure testing method: Testing material of material age 7 days (φ5 x 9 cm) was exposed inside temperature and humidity control
equipment that was set to various exposure conditions (Fig. 2).
Table 1

Composition of solidified substances
Waste filling rate

Water / Solidratio

Si / Al ratio

Na / Si
ratio[mass%][mass%][mol/mol][mol/mol]Base material

-

45

-

-

Cement + CS

30

54

-

-

Cement + IS

20

48

-

-

Base material

-

150

1.8

1.2

M+CS

30

92

1.8

1.2

M+IS

20

100

1.9

1.0

Base material

-

100

1.8

0.8

MB40+CS

30

80

1.8

0.8

MB40+IS

20

87

1.9

0.95

Sample name Label*

Cement

AAM
(M)

AAM
(MB40)*Explanation of abbreviations
CS: Simulated carbonate slurry
IS: Simulated iron co-precipitation slurry
Cement: Ordinary Portland Cement (OPC)
AAM(M): Metakaolin 100mass%
AAM(MB40): Metakaolin 60mass% + blast furnace slag 40mass%

Fig. 1 Solidified substance placed in the mold
*By using a translucent mold, the status inside themold (amount of bleeding, etc.) can be roughly
observed without removing the contents of the mold.

Fig. 2 The inside of the temperature and humidity
control equipment©International Research Institute for Nuclear Decommissioning

219

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test conditions 2. Setting the cyclic conditions simulating the storage environment ➢ Setting of cyclic conditions simulating the storage environment[Environmental load conditions to which solidified substances are expected to get exposed during storage]• Stored in sealed solidification containers (No contact with outside air → conditions not conducive todrying)• Cyclic changes in temperature during the day and at night and changes in temperature due to theseasons occur in the storage environment (indoors)
Fig. 1 Set temperature conditions to which
solidified substances get exposed• Range of change in temperature: about -5 to 45℃
Set up by adding the increase in temperature in a system wherein multiple drums are arranged
Table 1 Temperature conditions set for 1 cycle (6h)(Cs-137, 1E+9 Bq/cm3, natural convection current 20℃→27℃) to the Namie town AMeDAS
Temperature set in
Olympic data (Summer: 37.3℃ in 2018, Winter: -12.4℃ in 2014).
Time [min]the constantCyclic conditions: In order to accelerate the change in solidified substances, a cycle
temperature reservoirof maximum temperature 45℃ → Minimum temperature -5℃ → Maximum
10
20℃→45℃
temperature 45℃ was set as 1 cycle of 6 hours. (4 cycles in 24 hours, 20 minutes
each of rise and fall in temperature, maximum and minimum temperature160
45°C
maintained for 160 minutes) (Fig. 1, Table 1)
20
45℃→-5℃➢ Method of evaluating the change in properties of solidified substances• The phenomena that are likely to occur including results of studies until FY2020 were
investigated, and property data (Table 2) and sites to be analyzed (Fig. 2) were
determined.

3cm

Table 2

Property data to be evaluatedAnalysis methods

9cm

No.

Property

Analysis sites①

Compressive strength

Ultrasonic waves (Non-destructive)-②

DensityNon-destructive (Calculated based on
dimensions and weight)-③

Amount of moisture(Free water)

Coarse crushing ⇒ Method of drying
at 105℃

Top and bottom④

Crystalline structure(XRD)

Pulverizing ⇒ XRD

Top and bottom⑤

Pore size distribution

Coarse crushing ⇒ Pretreatment ⇒MIP

Top and bottom

6cm

160

-5°C

10

-5℃→20℃

Top

Diametricaldirection

Bottom(Vertical center)Fig. 2 Image of the sites to beanalyzed on the sample•

In order to verify the internal temperature of thesolidified substances during exposure, athermocouple was embedded into 1 solidified
substance sample of each composition.

©International Research Institute for Nuclear Decommissioning220

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Change in external appearance 1. Photographs -➢ Change in external appearance after exposure (Table 1)
Table 1
CementCement + CS

Photograph of the solidified substances after exposure to cyclic conditions
Cement
+IS

AAM
(M)

AAM
(M+CS)AAM(M+IS)

AAM(MB40)

AAM(MB40+CS)

AAM(MB40+IS)

100th
cycle

200th
cycle

280th
cycle

✓
✓
✓
✓

100th cycle: Cracks developed at the top and bottom in all test specimens of only AAM (M + CS). Other solidified substances had no abnormalities.
200th cycle: Cracks developed at the top and bottom in all test specimens of only AAM (M + CS). Other solidified substances had no abnormalities.
280th cycle: Multiple cracks developed in all test specimens of only AAM (M + CS). Other solidified substances had no abnormalities.
(Summary) Only the AAM (M + CS) solidified substance developed cracks due to cyclic temperature load.

©International Research Institute for Nuclear Decommissioning221

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- External appearance 2. Investigation of the solidified substances that were damaged➢ Investigation related to solidified substances in which damage was seen•

Data from the thermocouple embedded into 1 solidified substance sample of each composition was confirmed for verifying internal
temperature of the solidified substances during exposure (Fig. 1, Fig. 2).
60

60

OPC
OPC+CS
OPC+ISAAM(MB40)
AAM(MB40+CS)
AAM(MB40+IS)

40
30
20
10
0
-10

40
30

✓ It was found that the internal
temperature of the solidified substances
was different only in the case of theAAM (M + CS) solidified substancesthat developed cracks.

20
10
0

1225

1230

1235

1240

-10

1245

Time (hr.)
Fig 1. Changes in the internal
temperature of solidified
substances

1225

1230

1235

1240

1245

Time (hr.)
Fig 2. Change in the internal temperatureof solidified substances (Showing only
Cement + CS, AAM (M + CS))

AAM (M + CS) solidified substance of the same composition was additionally prepared for cause investigation, and temperature load
test was conducted under cyclic conditions (Fig. 3).
60✓ The internal temperature of the additionally prepared AAM (M + CS) solidified substanceAAM(M+CS)
AAM(M+CS)-New
showed a similar tendency. Thus it was verified that this tendency is peculiar to AAM (M +50
CS) solidified substance.
Temperature (°C)•

OPC+CSAAM(M+CS)

50

Temperature (°C)

Temperature (°C)

50

AAM(M)AAM(M+CS)
AAM(M+IS)

40✓

This implies that the occurrence of the phenomenon of intermittent change in the internal
temperature during rise in temperature and during fall in temperature is associated withchanges in thermal conductivity and thermal capacity of the solidified substance causedby temperature, and absorption and generation of heat.

✓

Since a tendency of internal temperature that is different from that of other solidified
substances was verified in the first cycle of the cyclic conditions, it is presumed that thecracks developed in the first cycle.

✓

The cause for the correlation between this peculiar tendency of internal temperature andthe cracks needs to be studied in detail in the future.

30
20
10
0
-10
960

965

970

975

Time (hr.)

980

985

990

Fig. 3 Changes in the internal temperature of AAM (M +CS) solidified substances(The portion that was initially prepared and the portion thatwas prepared additionally)©International Research Institute for Nuclear Decommissioning222

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test results 1. ① Compressive strength, ② Density, ③ Amount of free water -

80

32
days

57
days

82
days

OPC
OPC+CS
OPC+IS

60

40

20

0

0

100

200

Estimated compressive strength(N/mm2)① Estimated compressive strength

Estimated compressive strength(N/mm2)

Number of days
since placement

AAM (M) based
Number of days
since placement

300

80

82
days

57
days

AAM(M)AAM(M+CS)
AAM(M+IS)

60

M+CS
280サイクル目
推定不可

40

20

0

0

100

200

300

Cycle (Count)

Cycle (Count)

✓

32
days

AAM (MB40) based
Number of days
since placement
Estimated comressive strength(N/mm2)

Cement based*The measurement method for AAM (M+ CS) in which cracks developed waschanged. (Ultrasonic waves wereapplied from the sides of the solidified
substance)

80

32
days

57
days

82
days

100

200

300

AAM(MB40)
AAM(MB40+CS)
AAM(MB40+IS)

60

40

20

0

0

Cycle (Count)

Along with the increase in number of cycles, the strength of cement appeared to increase. However, this strength development is believed to be caused by the progress of the hydration reaction andhence it is inferred that it is not an impact of the cyclic conditions.
The strength of AAM (M), AAM (MB40) solidified substances decreased from 0th → 100th cycle.
2.0

2.0

Upper part

Upper part
Density (g/cm3)

Density (g/cm3)

1.7
OPC
OPC+CS
OPC+IS

1.6
1.5

0

100

200

M+CS
280サイクル目
測定不可

1.8
1.7
1.6
1.5

300

1.7

0

100

200

1.5

300

0

100

200

300

Cycle (count)40

Upper part

OPC
OPC+CS
OPC+IS

35
30
25
20
15

0

100

200

300

Amount of free water (mass%)

There was no major change in any of the solidified substances. (There was no difference in the top and bottom parts of the solidified substance.)
Amount of free water (mass%)③ Amount of free water

1.8

Cycle (count)

Cycle (count)✓

AAM(MB40)
AAM(MB40+CS)
AAM(MB40+IS)

1.6

Cycle (count)✓

Upper part
1.9

40
35
30
25

Upper part
AAM(M)AAM(M+CS)
AAM(M+IS)

20
15

0

100

200

Cycle (count)300

Amount of free water (mass%)② Density

1.8

2.0

AAM(M)AAM(M+CS)
AAM(M+IS)

1.9

1.9

Density (g/cm3)✓40

Upper part

35
30
25
AAM(MB40)
AAM(MB40+CS)
AAM(MB40+IS)

20
15

0

100

200

300

Cycle (count)

None of the solidified substances showed any significant change.

©International Research Institute for Nuclear Decommissioning223

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test results 2. ④ Crystalline structure (XRD)④ Change in the crystalline structure of solidified substances due to cyclic temperature load
• As there was no difference in the top and bottom portion of the solidified substance, the results ofTable 1 Crystalline structure analysis results
the bottom portion were used for comparison (Fig. 1, Table 1).
Cycle number

AAM (M)
AAM(M)-0Cycle
AAM(M)-100Cycle
AAM(M)-200Cycle
AAM(M)-280Cycle

AAM(M+CS)-0Cycle
AAM(M+CS)-100Cycle
AAM(M+CS)-200Cycle
AAM(M+CS)-280Cycle

AAM(M+IS)-0Cycle
AAM(M+IS)-100Cycle
AAM(M+IS)-200Cycle
AAM(M+IS)-280Cycle

Cement

AAM
(M)

10

20

30

40

50

AAM(MB40)-0Cycle
AAM(MB40)-100Cycle
AAM(MB40)-200Cycle
AAM(MB40)-280Cycle

AAM(MB40+CS)-0Cycle
AAM(MB40+CS)-100Cycle
AAM(MB40+CS)-200Cycle
AAM(MB40+CS)-280Cycle

100

200

280

Base material

―

―

―

―

CS

―

―

―

―

IS

―

―

―

―

Base material――✓ Formed
Faujasite✓ Increase inamount formed
Faujasite

CS

―

―

✓ Formed
Faujasite✓ Increase inamount formed
Faujasite

IS

―

―

✓ Formed
Faujasite✓ Increase inamount formed
Faujasite

Base material

―

―

―

―

CS

―

―

✓ Formed
Faujasite✓ Formed
Faujasite

IS

―

―

✓ Formed
Faujasite✓ Formed
Faujasite60

2θ/θ (°)

AAM(MB40)

0

AAM(MB40+IS)-0Cycle
AAM(MB40+IS)-100Cycle
AAM(MB40+IS)-200Cycle
AAM(MB40+IS)-280Cycle

AAM(MB40)

Legend

10

20

30

40

50

60

2θ/θ (°)

Change in constituent phase ―: Absent, ✓: Present(The name of the mineral listed represents the mineral facies that is formed)✓ Cement solidified substance: There was no change in mineral facies.
✓ AAM (M) solidified substance: With the increase in the number of cycles, there
was an increase in the quantity of Faujasite formed in the base material andCS, IS mixed solidified substances. (Faujasite is a type of zeolite.)✓ AAM (MB40) solidified substance: CS and IS mixed solidified substances
formed Faujisite from 200th cycle onwards.

Fig. 1 Crystalline structure analysis results (AAM(M), AAM(MB40))©International Research Institute for Nuclear Decommissioning224

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test results_3. ⑤ Pore size distribution (MIP) - *As the sample taken after subjecting to cyclic load

⑤ Pore size distribution due to cyclic temperature load
• As there was no difference in the top and bottom part of the solidified substances, the results of the top
part were used for comparison (Fig. 1).
60

40
30
20
10
0
0.001

0.01

0.1

1

10

100

AAM(M)-0 Cycle
AAM(M)-100 Cycle
AAM(M)-200 Cycle
AAM(M)-280 Cycle

50
40
30
20
10
0
0.001

Cumulative porosity (Vol%)

60

AAM(MB40)

Cumulative porosity (Vol%)

Upper part

0.01

0.1

1

10

100

1000

Average pore diameter (µm)
Upper part
AAM(MB40)-0 Cycle

40

100 Cycle · 200 Cycle: ? ? ? ?

30
20
10
0
0.001

0.01

0.1

1

10

100

1000

Average pore diameter (µm)

✓ Cement

30
20
10

0.01

0.1

1

10

100

40

20
10

0.01

0.1

1

10

100

30
20
10

Upper part

30
20
10
0
0.001

0.01

0.1

1

10

100

1

1000

Upper part

10

100

1000

AAM(M+IS)-0 Cycle
AAM(M+IS)-100 Cycle
AAM(M+IS)-200 Cycle
AAM(M+IS)-280 Cycle

30
20
10
0
0.001

1000

AAM(MB40+CS)-0 Cycle
AAM(MB40+CS)-100 Cycle
AAM(MB40+CS)-200 Cycle
AAM(MB40+CS)-280 Cycle

40

0.1

40

60

50

0.01

50

Average pore diameter (µm)

Average pore diameter (µm)

Fig. 1

40

60

30

0
0.001

OPC+IS-0 Cycle
OPC+IS-100 Cycle
OPC+IS-200 Cycle
OPC+IS-280 Cycle

50

Average pore diameter (µm)

AAM(M+CS)-100 Cycle
AAM(M+CS)-200 Cycle
AAM(M+CS)-280 Cycle

50

IS
Upper part

0
0.001

1000

Average pore diameter (µm)
60Upper part
AAM(M+CS)-0 Cycle

60

AAM(MB40)-100 Cycle
AAM(MB40)-200 Cycle
AAM(MB40)-280 Cycle

50

40

0
0.001

Cumulative porosity (Vol%)

AAM
(M)

Cumulative porosity (Vol%)

60

50

1000

Average pore diameter (µm)

60
OPC+CS-0 Cycle
OPC+CS-100 Cycle
OPC+CS-200 Cycle
OPC+CS-280 Cycle

Cumulative porosity (Vol%)

50

CS
Upper part

Cumulative porosity (Vol%)

OPC-0 Cycle
OPC 100-Cycle
OPC 200-Cycle
OPC-280 Cycle

Cumulative porosity (Vol%)

Base material
Upper part

Cumulative porosity (Vol%)

CementCumulative porosity (Vol%)60

broke down from the time it was removed from the mold,it was not able to be measured.

0.01

0.1

1

10

100

1000

Average pore diameter (µm)
Upper part
AAM(MB40+IS)-0 Cycle
AAM(MB40+IS)-100 Cycle
AAM(MB40+IS)-200 Cycle
AAM(MB40+IS)-280 Cycle

50
40
30
20
10
0
0.001

0.01

0.1

1

10

100

1000

Average pore diameter (µm)

Results of measuring pore size distribution: Along with the increase in number of cycles, the pore size became compact, but it is believed that this was because of crystal growth due to the progress in hydration
reaction. Hence it is presumed that this was not an impact of the cyclic conditions.
: No change in base material. The CS mixed sample was unable to be measured as it had broken down. The pore size of the IS mixed sample increased from 0th → 100th✓ AAM (M)cycle
✓ AAM(MB40) : No change in base material. The porosity of the CS mixed sample somewhat increased and the pore size became slightly larger as well. The pore size of the IS mixed
sample increased from 0th → 100th cycle.
©International Research Institute for Nuclear Decommissioning225

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Test results_4. At a glance (Changes observed due to the load of cyclic conditions) • Each analysis was conducted 4 times at 0th, 100th, 200th, and 280th cycle. The changes seen between 2 milestone cycles are listed in the table below.
Changes between 0th → 100th cycle

Cycle number

Cementbased

AAM(M) based

AAM(MB40)
based

Changes between 100th → 200th cycle

Changes between 200th → 280th cycle

CracksXRD

MIP

CracksXRD

MIP

Cracks

XRD

MIP

Base
material――✓
Compactness――✓
Compactness

―

―

―

CS

―

―

✓
Compactness――✓
Compactness

―

―

―

IS

―

―

✓
Compactness――✓
Compactness

―

―

―

Base
material

―

―

―

―

✓
Formed――✓Increase inamount formed✓
Increase in size――✓
Formed―✓Increase inamount formed

―

CS

IS

―

―

✓
Increase in size―✓
Formed✓Compactness―✓
Increase inamount formed✓
CompactnessBasematerial

―

―

―

―

―

―

―

―

―

CS

―

―

✓
Increase in size―✓
Formed✓Increase in size―✓
Formed✓Compactness

IS

―

―

✓
Increase in size―✓
Formed✓Compactness―✓
Formed―LegendCracks ―: Absent, Photos: present

Change in constituent phase (XRD) ―: Absent, ✓: Present
Change in porosity (MIP) ―: Absent, ✓: Present✓ There was no impact of cyclic conditions on cement solidified substances.
✓ Cracks developed or there were changes in the constituent phase of AAM (M) solidified substances due to load of cyclic conditions.
✓ There were changes in the constituent phase of AAM (MB40) solidified substances due to load of cyclic conditions.
©International Research Institute for Nuclear Decommissioning226

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueii. Investigation of the change in properties of solidified substances due to heating, etc.
- Summary -Results so far✓ Tests were conducted to investigate the changes in properties of solidified substances due tochange in temperature within the range of -5℃ to 45℃ that is determined by considering theamount of heat generated by solidified substances in addition to the maximum and minimum
temperatures in Namie town.
✓ Results of observing the changes in properties of solidified substances exposed to cyclic
conditions simulating storage environment are given below.

✓ There were no cracks or changes in mineral facies in the cement solidification substances.
Hence it was found that the properties do not change even in storage environment.
✓ It was found that cracks are likely to develop in AAM (M) solidified substances depending on
the type of waste mixed.
✓ Although cracks were not seen in the AAM (MB40) solidified substances, as change in theconstituent phase (change from amorphous to mineral facies) was observed, it was foundthat the solidification performance such as leaching characteristics of waste, etc. is likely tochange (deteriorate).
Issues, etc.
✓ As the solidification performance of AAM (M) and AAM (MB40) solidified substances is likely
to deteriorate due to the impact of storage environment depending on their combination with
waste, methods for checking the combination with waste in advance are necessary.

©International Research Institute for Nuclear Decommissioning227

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances•

As the performance of solidified substances is affected by temperature in the case of low-temperature treatment technology (cement, AAM), the upper
temperature limit depends on the material.
As heat transfer depends on the system, the limit value of waste filling rate is expected to change depending on the shape and size of the container.

•

By analyzing the maximum temperature and clarifying the changes due to shape and dimensions, the solidifiable inventory limit value
studied based on decay heat is acquired as data.
◼

Achievements up to FY2020••◼

Heat generation by the inventory contained in the system in which stand-alone
solidified substances and cylindrical solidified substances are arranged in an
accumulated state was analyzed with the shape of the container, solidificationmaterial and radioactive nuclides as parameters, and a relational expression foreasily estimating the temperature of the solidified substance was derived.

Analyticalvalue

Conversion coefficient for
converting nuclide inventories into
calorific values (drum can)The following issue was identified with respect to on-site application and further
refinement of data.
✓
Increasing knowledge pertaining to accumulation type storage form (squarecontainers) and deriving a formula for evaluating the highest temperature
attained taking that knowledge into consideration.
Goal

Estimation of the thermal
environment when filled withsolidified substances and setting
the limits

Coefficient with regard to the profile difference (filling amount) (drum can, square container)

Coefficient with regard to the material (thermal conductivity) (drum can)Coefficient with regard to the layout (drum can, square containers)•

The temperature of the solidified substance when radioactive waste is solidified will be analyzed, and the relationship between nuclides such as Cs, etc. containedin the solidified substance and the temperature of the solidified substance will be evaluated.

•

The inventory that can be contained while maintaining integrity of the solidified substance will be studied based on the evaluation results, thereby contributing to
the estimation of heat impact.

Details of implementation

Investigation of thermal effect when square containers are stored in an accumulated state
・Creating the system model ・Setting the analysis conditions ・Analyzing thermal effect of Cs and other nuclides
Derivation of the formula for estimating thermal effect・Derivation of formula for obtaining the temperature and amount of heat generated by containers that are arranged in astand-alone state ・Derivation of formula for obtaining the temperature and amount of heat generated by containers that arearranged in an accumulated state

Indicators for the achievements of goal
The temperature attained when square shaped containers are accumulated and stored is evaluated andthe method of calculating it is presented.
©International Research Institute for Nuclear Decommissioning228

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Analysis of the effect of arranging square containers in an accumulated state on heat generation 1. Model and analysis conditions ➢Creation of accumulation type arrangement model for square containers, and setting of analysis conditions• Solidified substance containers were stored in an accumulated state by arranging multiple containers horizontally and vertically, and the increase intemperature due to the thermal effect amongst containers was estimated.
• A model was created for analyzing the amount of heat generated and the temperature with respect to square containers that have excellent container
storage efficiency as compared to drum cans, and which have undergone analysis when arranged in an accumulated state, for on-site application of
the analytical data.
• Analysis conditions were set up assuming secondary waste generated from contaminated water treatment as waste that has undergone cement
solidification.
X1
X2
X3

The focus was on thetemperature of the blue
container (the container at
the center of 27 accumulated
containers)

Y
X➢

Analysis conditions
Analysis conditions

Details

Container

27 square 1m3 containers (Thickness 0.2 [cm])

Solidification material

CementSolidified substancedensity[g/cm3]

2.69

Activity concentration [Bq/cm3]
Software used✓

X

Fig. 2 Image illustrating the division of thesolidified substances(Stand-alone containers and bird's-eye view)Fig. 1 Image of accumulation type arrangement

1.0E+05, 06, 07, 08, 09
PHITS (Analysis of amount of heat generated)
COMSOL (Temperature analysis)➢

Rough divisionof column X3

YFig. 3 Image illustrating the division of thesolidified substances(Containers in an accumulated state andbird’s-eye view)

Information on the accumulation type arrangement model for square containers• Cement with uniform radiation is filled in 1m3 square containers up to 80 % of
their height and 27 such containers were accumulated.
• The containers were stacked so that the length, breadth and height was 3 x 3x 3 containers such that it was possible to observe the thermal impact of thesurrounding on the container at the center. (Fig. 1)• The portion with solidified substance in a single container was divided into 108
cells of length breadth and height of 6 x 6 x 3 (Fig. 2) (For the sake of
convenience while setting up the software used for analyzing the amount ofheat generated, the division of the containers in the X3 column was roughly
set up. (Fig. 3))

An accumulated type arrangement model with 1m 3 square containers which have a good storage efficiency was created for on-site applicability of the analytical data.

©International Research Institute for Nuclear Decommissioning229

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Analysis of the effect of arranging square containers in an accumulated state on heat generation 2. Analysis of amount of heatgenerated ➢Analysis of amount of heat generated when square containers are arranged
in an accumulated state• The amount of heat generated in the case of accumulation typearrangement in which heat generation is expected to increase due toradioactive interaction between containers, was analyzed in order to obtainthe values used for temperature analysis.
➢
Analysis conditions• The change in the amount of heat generated by solidified substances due
to accumulation is anticipated to be different for different nuclides
depending on the type of radiation.
• Sr-90, Cs-137, Co-60 which emit radiation of different intensities wereanalyzed.
• In order to evaluate the impact of accumulation, the amount of heatgenerated by each stacked container was arranged against the amount ofheat generated by a stand-alone container assuming that heat generated
by the stand-alone container is 1.00.
➢

Results of analyzing the amount of heatgenerated✓ In the case of Co-60 and Cs-137 the amountof heat generated increased as compared tothe stand-alone type arrangement (Fig. 1, Fig.
2), and the magnitude relationship of the rateof increase was consistent with magnituderelationship of the energy from gamma rays.
(Table 1)✓ The impact of surrounding containers depends
on the energy from the emitted gamma rays.
✓ In the case of Sr-90, the amount of heatgenerated from the stand-alone typearrangement did not increase. (Fig. 3)✓ In the case of nuclides that do not produce
gamma rays, the impact of accumulation was
only due to heat transfer.
✓

1.05

1.07

1.05

1.07

1.09

1.07

1.05

1.06

1.05

1.06

1.08

1.07

1.09

1.11

1.09

1.06

1.08

1.06

1.05

1.07

1.05

1.07

1.09

1.07

1.05

1.07

1.05

Lower level

Upper level

Mid level

Fig. 1 Rate of increase in amount of heat generated considering that theamount of heat generated by a 1m 3 container is 1.00 (Co-60)

1.02

1.03

1.02

1.03

1.04

1.03

1.02

1.03

1.02

Table 1 Decay energyof nuclides

1.03

1.04

1.03

1.04

1.05

1.04

1.03

1.04

1.03

Nuclides

Gammarays [MeV]

1.02

1.03

1.02

1.03

1.04

1.03

1.02

1.03

1.02

Co-60

1.173
1.332

Cs137

0.662
0.03210.0365

Sr-90

None

It was verified that the increase in amount of heat generated due toaccumulation depends on the type of radiation released by the nuclide
and the energy of the gamma rays.

Lower level

Upper level

Mid level

Fig. 2 Rate of increase in amount of heat generated considering that theamount of heat generated by a 1m 3 container is 1.00 (Cs-137)

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

Lower level

Mid level

Upper level

Fig. 3 Rate of increase in amount of heat generated considering that theamount of heat generated by a 1m 3 container is 1.00 (Sr-90)©International Research Institute for Nuclear Decommissioning230

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Analysis of the effect of arranging square containers in an accumulated state on heat generation 3. Temperature analysis ➢Temperature analysis when square containers are arranged in an accumulated state
The temperature in the case of accumulation type arrangement in which heat generation is expected to increase due to thermal interaction between containers, wasanalyzed under 2 conditions namely, when there is no convection current and when natural convection is taken into consideration.
• The initial temperature of the ambient air and the solidified substance was set at 20℃, and the temperature was analyzed using the results of the analysis of amountof heat generated (Cs-137).
• During actual storage, the heat removal (cooling effect) due to the natural convection current resulting from the difference in the temperature of the solidified
substance and the ambient air needs to be considered.
• First, on the safer side, the condition wherein convection current because of which the temperature result is expected to be on the higher side, is not considered(gravity was not set), was analyzed.
• Next, the gravity was added to the same model, so that the condition wherein natural convection current because of which a more realistic result is expected, isconsidered, was analyzed.
•✓ The maximum temperature of solidified
substances was 38.70℃ under natural convection
current conditions.

Fig. 1 Results of temperature analysis
under conditions without convection current➢•

Fig. 2 Results of temperature analysis under naturalconvection current conditions

Estimation of the upper limit value of inventory using the analysis results
In order to evaluate the inventory that can be contained while maintaining integrity of the solidified substance, the results of analyzing conditions considering
natural convection current were plotted on a graph with the activity concentration of the solidified substance on the x axis and the maximum temperature of thesolidified substance on the y axis (Fig. 1), and the inventory limit value was obtained using the approximation formula.

✓ The relational expression between activity concentration and maximum temperature of the solidified
substance was derived from the plotted analysis results.
𝑦 = −6.35 × 10−19 𝑥 2 + 1.94 × 10−8 𝑥 + 20 (Formula 1)•
𝑥: Activity concentration [Bq/ cm3]•
𝑦: Maximum temperature of solidified substance [℃]✓ Since 𝑦 = 60, the inventory limit value of Cs-137 is estimated to as 2.22 × 109 Bq/cm3✓ The inventory limit value of Cs-137 obtained using the relational expression between activity
concentration and maximum temperature of solidified substance derived using the temperature analysis
results considering natural convection current was 2.22 × 109 Bq/cm3.
Fig. 1 Relation between activity concentration and maximum temperature of the solidified substance©International Research Institute for Nuclear Decommissioning231

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 1. Stand-alone type arrangement 1/2 -

•

•

•
•

➢

Estimation of thermal impact in a stand-alone type arrangement
The formula for obtaining temperature under arbitrary conditions was derived from theanalysis results to facilitate estimation of the temperature of solidified substances
containing radioactive waste.
The relational expression between the amount of heat generated and the temperature wasderived with the shape of the containers arranged in a stand-alone state as the parameter.
(e)
The relational expression between activity concentration and amount of heat generated
was derived with the nuclide, shape of container and solidification material as theparameters. (a, b. c)
The 2 formulas were used in combination to obtain the temperature of solidified
substances with arbitrary parameters.

Derivation of the relational expression between the amount of heatgenerated and the temperature (e)•The formula (Formula 2) for obtaining the maximum temperaturefrom the amount of heat generated was derived based on the
results of analysis using each system model. (Table 1)
T = 𝑜W 2 + 𝑝W + 𝑞 (Formula 2)••

𝑇 : Maximum temperature [℃]
𝑊 : Amount of heat generated [W/m3](Stand-alone type arrangement Cell with maximum temperature)aAnalytical
value

e

Conversion coefficient for
converting nuclide inventories into
calorific values (drum can)Estimation of the thermal
environment when filled withsolidified substances and setting
the limits

Coefficient with regard to the profile difference (filling amount) (drum can, square container)

Coefficient with regard to the material (thermal conductivity) (drum can)Coefficient with regard to the layout (drum can, square containers)bcd

Table 1 List of relational expressions pertaining to the
maximum temperature in each system

System modelStand-alone type arrangement➢

Relational expression between the amount ofheat generated and the maximum temperature

1m3 square

T = −2.16 × 10−4 W 2 + 0.70W + 20

6m3 square

T = −3.03 × 10−3 W 2 + 1.70W + 20

1m3 squarewith shield

T = −1.33 × 10−4 W 2 + 0.43W + 20

Pails

T = −6.40 × 10−5 W 2 + 0.10W + 20

Drum cans

T = 3.06 × 10−5 W 2 + 0.35W + 20

HIC

T = −2.76 × 10−3 W 2 + 1.28W + 20

©International Research Institute for Nuclear Decommissioning232

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 2. Stand-alone type arrangement 2/2 ➢Analyticalvalue

Derivation of relational expression between activity concentration and amount of heat generated (a, b.
c)•
The formula for obtaining the amount of heat generated from the activity concentration was derived
using the results of analyzing the nuclides, shape of container and solidification material.

ae

Conversion coefficient for converting
nuclide inventories into calorificvalues (drum can)Estimation of the thermal
environment when filled withsolidified substances and setting
the limitsbcdCoefficient with regard to the profile difference (filling amount) (drum can, square container)

Coefficient with regard to the material (thermal conductivity) (drum can)Coefficient with regard to the layout (drum can, square containers)•

First, basic conditions were established, and the relational expression between activity
concentration and amount of heat generated under those basic conditions was derived. (Formula 3)𝑊=•

1.21 × 10−13

× 𝑥 (Formula 3)••

Table 1 Basic conditions

𝑊: Amount of heat generated [W/m3]
𝑥: Activity concentration [Bq/ cm3]

Next, the results under basic conditions and other conditions were multiplied with the formuladerived from the basic conditions as coefficients of each parameter and the formula (Formula 4) for
obtaining the amount of heat generated under arbitrary conditions was derived.
1. Analysis conditions (selected)
𝑊 = 1.21 × 10−13 × 𝑎𝑁𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137 × 𝑎𝐶𝑜𝑛𝑡𝑎𝑖𝑛𝑒𝑟 Τ𝑎𝑑𝑟𝑢𝑚 ×
𝑎𝑆𝑜𝑙𝑖𝑑𝑖𝑓𝑖𝑐𝑎𝑡𝑖𝑜𝑛 𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 Τ𝑎𝐶𝑒𝑚𝑒𝑛𝑡 (𝐿𝑖𝑡𝑒𝑟𝑎𝑡𝑢𝑟𝑒 𝑑𝑎𝑡𝑎) × 𝑥 (Formula 4)

Container

SquareNuclidesCs-137

Shape ofcontainer

Drum cans

Solidificationmaterial

Cement3. Estimated amount of heat generated(Cell with maximum temperature in the standalone type arrangement)

Solidificationmaterial

Nuclides•••➢Coefficient of nuclides: 𝑎𝑁𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137
Coefficient of the shape of container: 𝑎𝐶𝑜𝑛𝑡𝑎𝑖𝑛𝑒𝑟 Τ𝑎𝑑𝑟𝑢𝑚
Coefficient of solidification material: 𝑎𝑆𝑜𝑙𝑖𝑑𝑖𝑓𝑖𝑐𝑎𝑡𝑖𝑜𝑛 𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 Τ𝑎𝐶𝑒𝑚𝑒𝑛𝑡 (𝐿𝑖𝑡𝑒𝑟𝑎𝑡𝑢𝑟𝑒 𝑑𝑎𝑡𝑎)

2. Activity concentration (entered)

5. Estimated maximum temperature(stand-alone type arrangement)

Activity concentration

Fig. 1 Example of a calculation sheet for
Combined use of the formulas for estimating the temperature and amount of heat generated
estimating maximum temperature
Finally, by using Formula 4 and Formula 2 from the previous page in combination, it became possible
in this section to estimate the maximum temperature of solidified substances under arbitrary
conditions and activity concentration in the range of analysis results. (Fig. 1)
Table 2 Values used as coefficients3•
For example, if the nuclide is configured as Co-60, shape as a 1m square, solidification
Coefficient
Conditions
material as AAM(M) and activity concentration as 1.0E+08 [Bq/cm3] (Table 2),
based on𝑎𝑁𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137
Co-60
𝑊 = 1.21 × 10−13 × 2.96 × 1.01 × 0.92 × 1.00 × 108
𝑎𝐶𝑜𝑛𝑡𝑎𝑖𝑛𝑒𝑟 Τ𝑎𝑑𝑟𝑢𝑚
1m3 square
the amount of heat generated by the solidified substance was estimated to be 3.31 × 101
container
3[W/m ], and based on−42−1
𝑎𝑆𝑜𝑙𝑖𝑑𝑖𝑓𝑖𝑐𝑎𝑡𝑖𝑜𝑛 𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 Τ𝑎𝐶𝑒𝑚𝑒𝑛𝑡 (𝐿𝑖𝑡𝑒𝑟𝑎𝑡𝑢𝑟𝑒 𝑑𝑎𝑡𝑎
T = −2.16 × 10 W + 7.04 × 10 W + 20
AAM (M)the temperature of the solidified substance was estimated to be 43.05℃.

•

Values
2.96
1.01
0.92

✓ The formula for estimating the amount of heat generated was derived using the formula for estimating the maximum temperature of the solidified substance having an arbitrary shapeand the coefficients of nuclides, shape of container, and solidification material respectively.
✓ By using the 2 formulas in combination, it became possible to estimate the maximum temperature of solidified substances under arbitrary conditions and activity concentration.
©International Research Institute for Nuclear Decommissioning233

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 3. Accumulation type arrangement 1/2 ➢Estimation of thermal impact in an accumulation type arrangement
The results of the stand-alone type arrangement that has sufficient analysis results(a, b, c) with nuclides and solidification material as parameters, were extended tothe accumulation type arrangement to facilitate estimation of the temperature ofsolidified substances in an accumulation type arrangement. (d)•The relational expression between the amount of heat generated and thetemperature was derived. (e)•The relational expression between the amount of heat generated in a stand-alone
type arrangement and in an accumulation type arrangement was derived. (d)•The 2 formulas were used in combination to obtain the temperature of solidified
substances with arbitrary parameters.
•
The scope of reflection of the formulas derived for the accumulation typearrangement is as follows.
•
Nuclide: Cs-137•
Solidification material: Cement•
Accumulated containers: Drum can, 1m 3 square container
•

•

a

e

Conversion coefficient for converting
nuclide inventories into calorificvalues (drum can)Estimation of the thermal
environment when filled withsolidified substances and setting
the limits

Coefficient with regard to the profile difference (filling amount) (drum can, square container)

Coefficient with regard to the material (thermal conductivity) (drum can)Coefficient with regard to the layout (drum can, square containers)bcd

Derivation of the relational expression between the amount of heat generated and the temperature (e)The relational expression (Formula 5) for obtaining the maximum temperature from the amount of heat generated in an accumulation typearrangement when there was no convection current was derived based on the results of analysis using each system model. (Table 1)

T = 𝑜𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛 2 + 𝑝𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛 + 𝑞••

Table 1 List of relational expressions pertaining to the
maximum temperature in each system(Formula 5)

𝑇 : Maximum temperature [℃]𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛 : Amount of heat generated [W/m3] (Container withmaximum temperature in accumulation type arrangement)

System model

Accumulation typearrangement➢

Analytical
value

Relational expression between the amount of heatgenerated and the maximum temperature

108 Drum cans
Accumulation typearrangementT2= −1.70 × 10−2 W𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛+ 6.25W𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛 + 20

27 square 1m3
containers
Accumulation typearrangementT2= −9.6 × 10−3 W𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛+ 5.07W𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑖𝑜𝑛 + 20

©International Research Institute for Nuclear Decommissioning

234

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 4. Accumulation type arrangement 2/2 ➢aDerivation of the relational expression between the amount of heat generated in a stand-alone typearrangement and in an accumulation type arrangement (d)•The relational expression between the amount of heat generated in a stand-alone type arrangement and inan accumulation type arrangement was derived based on the analysis results (Formula 6)

Analytical
value

•
•

Estimation of the thermal
environment when filled withsolidified substances and setting thelimitsCoefficient with regard to the profile difference (filling amount) (drum can, square container)

𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑒𝑑 = 𝑏 × 𝑊𝑆𝑢𝑚 (Equation 6)•eConversion coefficient for converting
nuclide inventories into calorificvalues (drum can)Coefficient with regard to the material (thermal conductivity) (drum can)

𝑊𝑆𝑢𝑚 : Amount of heat generated [W/m3] (Sum of single units arrangement)𝑏 : Coefficient (Depends on system model, nuclide)
𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑒𝑑 : Amount of heat generated [W/m3] (Sum of the containers with maximum
temperature in accumulation type arrangement)

Coefficient with regard to the layout (drum can, square containers)bcd

1. Analysis conditions (selected)

Container

SquareSolidification material•

The coefficient b of each system can be expressed by the following equations (Equation 7, Equation 8) based on
the decay energy of the nuclides to be analyzed.
••

Drum cans: 𝑏 = 0.18 × 𝑙 + 1 (Equation 7)
1m3 square container:𝑏 = 0.06 × 𝑙 + 1 (Equation 8)•

Nuclides2. Activity concentration (entered)
Activity
concentration𝑙: Largest proportion of photon
energy emitted during decay ofeach nuclide

3. Estimated amount of heat generated(Cell with maximum temperature in the stand-alone type arrangement)•

Wsum was obtained with the following relational expression (Equation 9) based on the total amount ofheat generated in the stand-alone type arrangement.
•
𝑊: Amount of heat generated [W/m3] (Cell with maximum temperature in
𝑊𝑆𝑢𝑚 = 𝑘 × 𝑊 (Equation 9)
stand-alone type arrangement)•𝑘 : Coefficient (Depends on nuclide)•
𝑊𝑆𝑢𝑚 : Amount of heat generated [W/m3] (Sum of the single units
arrangement)➢
Combined use of the formulas for estimating the temperature and amount of heat generated•
By using Formula 6 and Formula 5 from the previous page in combination, it became possible in this section to estimate themaximum temperature of solidified substances under arbitrary conditions and activity concentration extended from theresults of the stand-alone type arrangement, in the range of analysis results. (Fig. 1)

4. Estimated amount of heat generated (Containers with maximum
temperature in the accumulation type arrangement)(27 containers_X2Y2Z3 container)(27 containers_X2Y2Z3 container)
5. Estimated maximum temperature (stand-alone type arrangement)

6. Estimated maximum temperature (accumulation type arrangement)(27 containers_X2Y2Z3 container)•

For example, if the nuclide is configured as Co-60, shape as a 1m3 square, solidification material as AAM(M) and activity
concentration as 1.0E+08 [Bq/cm3] (Table 1),
based on𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑒𝑑 = 0.06 × 𝑙 + 1 × 𝑘 × 𝑊
the amount of heat generated by the solidified substance is estimated to be 4.00 × 101 [W/m3], 𝑎𝑛𝑑 𝑏𝑎𝑠𝑒𝑑 𝑜𝑛
2
T = −9.6 × 10−3 𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑒𝑑+ 5.07𝑊𝑎𝑐𝑐𝑢𝑚𝑢𝑙𝑎𝑡𝑒𝑑 + 20
the temperature of the solidified substance is estimated to be 207.31℃.

✓ In this section, the equation for estimating the maximum temperature of solidified substances in an arbitrary accumulation typearrangement and the equation for estimating the amount of heat generated in an accumulation type arrangement from theamount of heat generated in a stand-alone type arrangement, in the range of the analysis results, were derived.
✓ By using the relational expressions in combination, it became possible to estimate the temperature of solidified substances in
the range of accumulation extended from the results of the stand-alone type arrangement.

Fig. 1 Example of a calculation sheet for estimating
maximum temperature

Table 1 Values used as coefficients
Coefficient

Conditions

ValuesW

Co-60
1m3 square container
AAM (M)

3.31 × 10−5

𝑙

Co-60

1.3320

𝑘

Co-60

0.97

©International Research Institute for Nuclear Decommissioning235

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Summary -Results so far
✓

✓
✓

✓

✓

An accumulation type arrangement model for 1m3 square containers which have a good storage efficiency
was created for on-site application of the analytical data.
It was found that the increase in amount of heat generated due to accumulation of containers depends onthe energy of gamma rays released by the nuclide.
The inventory limit value of Cs-137 obtained using the relational expression between activity
concentration and maximum temperature of solidified substance derived using the temperature analysis
results was 2.22 × 109 Bq/cm3.
The equation for estimating the maximum temperature of the solidified substances having an arbitrary
shape and the equation for estimating the amount of heat generated using coefficients of nuclides, shapeof container, and solidification material respectively, in the range of analysis results, were derived . By
using equations in combination, it became possible to estimate the maximum temperature of solidified
substances under arbitrary conditions and activity concentration.
The equation for estimating the maximum temperature of solidified substances in an arbitrary
accumulation type arrangement format and the equation for estimating the amount of heat generated inan accumulation type arrangement from the amount of heat generated in a stand-alone type arrangement,in the range of the analysis results, were derived. By using the equations in combination, it became
possible to estimate the temperature of solidified substances in the range of accumulation by extending
the results of the stand-alone type arrangement.

Issues, etc.
✓

For more detailed study such as for the investigation of heat removal effect due to air convection, etc., it isnecessary to conduct analysis in which conditions that take into account actual storage environment are
set up.

©International Research Institute for Nuclear Decommissioning236

Supplementary materials(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 2. Stand-alone type arrangement 2/2 Supplementary materials ➢The following coefficients are listed as supplementary information for - Estimation of the thermal impact based on
analysis results 2. Stand-alone type arrangement 2/2 Table 1 List of coefficients (𝑎𝑛𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137 )of nuclides in Equation 4
Nuclides
Sr-90
Y-90
Tc-99
Ru-106
Rh-106
Ag-110m
Cd-113m
Sn-119m
Sn-123
Sn-126
Sb-125
Te-123m
Te-125m
Te-127
Te-127m
I-129
Cs-134
Cs-137Ce-144Pr-144Pr-144m

𝑎𝑛𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137
1.49
1.23
0.14
1.87
2.12
3.31
0.25
0.02
0.71
3.41
0.71
0.19
0.05
0.31
0.42
0.12
2.04
1.00
1.76
1.64
1.31

NuclidesPm-146
Pm-147
Sm-151
Eu-152
Eu-154Eu-155
Pu-238
Pu-239Pu-240Pu-241
Am-241
Am-242m
Am-243
Cm-243
Cm-244
Mn-54
Co-60
Ni-63
Zn-65
H-3

𝑎𝑛𝑢𝑐𝑙𝑖𝑑𝑒 Τ𝑎𝐶𝑠−137
1.01
0.08
0.03
1.53
1.80
1.63
9.56
8.69
8.49
1.91
8.61
25.87
374.13
16.14
26.53
0.98
2.96
0.02
0.66
0.01

Table 2 List of coefficients (𝑎𝑐𝑜𝑛𝑡𝑎𝑖𝑛𝑒𝑟 Τ𝑎𝑑𝑟𝑢𝑚 ) of
the shape of containers in Equation 4
Shape of container
Pails
Drum cylinders
HIC
1m3 square container
6m3 square container
3
1m square container with shield𝑎co𝑛𝑡𝑎𝑖𝑛𝑒𝑟 Τ𝑎drum 𝑐𝑎𝑛
0.871
1.000
1.045
1.010
0.979
0.973

Table 3 List of coefficients(𝑎𝑆𝑜𝑙𝑖𝑑𝑖𝑓𝑖𝑐𝑎𝑡𝑖𝑜𝑛 𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 Τ𝑎𝐶𝑒𝑚𝑒𝑛𝑡 (𝐿𝑖𝑡𝑒𝑟𝑎𝑡𝑢𝑟𝑒 𝑑𝑎𝑡𝑎)of solidification material in Equation 4)

Solidification material
Cement (Literature
data)
Cement(Measurements)
AAM (M)

𝑎S𝑜𝑙𝑖𝑑𝑖𝑓𝑖𝑐𝑎𝑡𝑖𝑜𝑛 𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 Τ𝑎𝑂𝑃𝐶 (Literature data)

AAM(MB20)

0.92

AAM(MB40)

0.93

1.00

0.95
0.92

©International Research Institute for Nuclear Decommissioning

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Supplementary materials(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiii. Evaluation of the relationship between the inventory of Cs, etc. and the temperature of solidified substances- Estimation of the thermal impact based on analysis results 4. Accumulation type arrangement 2/2 Supplementary materials ➢The following coefficients are listed as supplementary information for - Estimation of the thermal impact based on
analysis results 2. Accumulation type arrangement 2/2 Table 2 Coefficients for each nuclide in the Equations 6 to 9
(Accumulation type arrangement of 1m 3 square containers)

Table 1 Coefficients for each nuclide in Equations 6 to 9
(Accumulation type arrangement of drum cans)

Nuclides
Sr-90
Y-90
Tc-99
Ru-106
Rh-106
Ag-110m
Cd-113m
Sn-119m
Sn-123
Sn-126
Sb-125
Te-123m
Te-125m
Te-127
Te-127m
I-129
Cs-134
Cs-137Ce-144Pr-144Pr-144m𝑙
0.00000.00000.00000.00000.51200.65800.00000.0252
1.08900.08760.02740.15900.02740.41800.02740.02970.60500.66200.13400.69700.0359b
1.00001.00001.00001.00001.09391.12071.00001.0046
1.19971.01611.00501.0292
1.00501.07671.00501.00541.11101.12141.0246
1.12781.0066

k
1.01
1.01
1.01
1.00
0.98
0.89
1.00
1.00
0.99
0.94
0.94
0.93
1.00
0.99
1.02
1.01
0.90
0.93
1.02
0.99
0.88

NuclidesPm-146
Pm-147
Sm-151
Eu-152
Eu-154Eu-155
Pu-238
Pu-239Pu-240Pu-241
Am-241
Am-242m
Am-243
Cm-243
Cm-244
Mn-54
Co-60
Ni-63
Zn-65
H-3

𝑙
0.45300.12100.02150.04000.12300.08650.01590.01590.01590.01560.01610.01790.07470.27760.01690.83501.33200.00001.11600.0000b
1.0831
1.0222
1.00391.0073
1.02261.01591.00291.00291.00291.00291.00301.00331.01371.05091.0031
1.15311.24431.00001.20461.0000

k
0.91
1.01
1.01
0.91
0.91
1.00
1.01
1.01
1.01
1.00
1.02
1.01
1.00
1.01
1.01
0.88
0.89
1.01
0.88
1.01

NuclidesSr-90
Cs-137
Co-60
Pu-238Am-241H-3𝑙0.00000.66201.33200.01590.01610.0000b1.00001.12141.24431.00291.00301.0000

k
1.00
0.98
0.97
1.01
1.02
1.00

©International Research Institute for Nuclear Decommissioning238

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Details of implementation ◼ Achievements up to FY2020• While a certain applicability has been confirmed through verification of the feasibility of estimating long-term changes inproperties of solidified substances, using thermodynamic equilibrium calculations, it was found that knowledge pertaining
to long term mineral facies changes needs to be acquired.
• The following issue was identified with respect to on-site application and further refinement of data.
✓ Acquisition of tentative knowledge for making it possible to estimate long-term changes in properties of solidified
substances◼ GoalIn order to investigate long term property change behavior of solidified substances, the method of accelerated testing of
long-term property change in solidified substances will be investigated, and the applicability of the investigated method ofaccelerated testing to evaluating longer term property change behavior of solidified substances will be studied.
Details of implementation
Investigation of the method of accelerated testing related to change in properties
Investigation and organization of the factors affecting change in properties of cement, concrete, materials of the engineered barriers forgeological disposal, glass, etc. which have been used in the past, the change in their properties and methods for evaluating the change
in properties (including accelerated testing), etc.
Data acquisition through accelerated testing and study of applicability
Implementation of actual accelerated testing based on the methods of accelerated testing that were investigated, and acquisition ofdata.

Indicators for the achievements of goal
The method of accelerated testing of degradation is studied and the investigation results are presented.

©International Research Institute for Nuclear Decommissioning

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(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment technique(b) iv. Investigation and evaluation of factors affecting long-term change in properties- Investigation of the methods of accelerated testing ➢ Investigation of the methods of accelerated testing related to change in properties•

The factors affecting change in properties of cement, concrete, materials of the engineered barriers for geological disposal, glass, etc. whichhave been used in the past, the change in their properties and methods for evaluating the change in properties (including accelerated testing),
etc. were investigated and organized (Table 1).
Table 1 Results of investigation and organization related to accelerated testing

TargetFactors leading to change
in properties

Temperature

Effect of the factors leading to change inproperties

Phenomenon that could
occur

Method of evaluating the change inproperties (an example)

Accelerated testing (anexample)Test of shrinkage due todryingJIS A 1151(2002)

Shrinking associated with drying

CracksStrength reduction

Visual confirmation, measurement ofchange in length, strength measurement,
etc.

Change in capacity due to change intemperature

CracksStrength reduction

Visual confirmation, strength
measurement, etc.

Heating-cooling cyclic test

JIS A 6909(2014)

Change in capacity due to freezing andthawing

CracksPeeling
Popping out

Strength reduction

Visual confirmation, elastodynamic
coefficient measurement, etc.

Freezing-thawing test

JIS A 1148 (2001)

Change in volume due to heat

CracksStrength reduction

Visual confirmation, measurement ofchange in length, strength measurement,
etc.

Heat curing test

Change in generation phase

CracksStrength reduction

Visual confirmation, measurement ofchange in length, strength measurement,
XRD, etc.

Heat curing test

Dry-wet cyclic test

Humidity

Change in capacity due to change inhumidity

CracksStrength reduction

Visual confirmation, measurement ofchange in length, strength measurement,
etc.

Salt content

Expansion due to steel corrosion

CracksStrength reduction

Visual confirmation, spontaneous
potential measurement,
strength measurement, etc.

Salt water immersion test
Salt spray test

JCI(2014)

Change in capacity due to carbonation

CracksStrength reduction

Steel corrosion due to decrease in pH of
concrete

Strength reduction

Visual confirmation, carbonation depth
measurement,
strength measurement, etc.

Accelerated carbonation
test

JIS A 1153(2003)

CracksImmersion test

Concrete

Carbon dioxide gas

Decomposition of cement hydrates

Strength reduction

Visual confirmation, determination of thequantity of hydrates,
strength measurement, etc.

Decomposition and leaching of cement
hydrates

CracksPeeling

Strength reduction

Visual confirmation,
determination of the quantity of hydrates(TG/DTA), strength measurement, etc.

Immersion test
Water flow test
Electrical enhancement
test
Chemical enhancement
test

Movement of water soluble components

Efflorescence

Deterioration of view
Strength reduction

Visual confirmation,color tone measurement (color difference
meter), etc.

Efflorescence test

Load

Deformation of components due to tensilestressCracksStrength reduction

Measurement of change in dimensions,strength measurement, etc.

Point load test

Temperature

Change in capacity due to change intemperature

CracksPeeling

Reduction in adhesive
ability

Heating-cooling cycle, etc.

Heating-cooling cyclic test

JIS A 6909(2014)

Visual confirmation,color tone measurement (color difference
meter), etc.

Accelerated weathering
test

JIS A 6909(2014)

Visual confirmation, strength
measurement, etc.

Heating-cooling cyclic test

JIS R 3223(2017)

Liquid phase composition analysis, etc.

Immersion test
Flowing water solubility
test

JIS R 3503(2007)

Liquid phase composition analysis, etc.

Immersion test
Flowing water solubility
test

JIS R 3503(2007)

Water (Rainfall)

Polymer
cementmortarUltraviolet rays

Polymer degradation

Chalking

Heat

Change in capacity due to change intemperature

CracksWater

Acid, alkali, etc.

Leaching of components

ErosionWeight reduction

Weight reduction

Strength reduction

Strength reduction

Strength reduction✓ The accelerated tests specified in JIS were
only for evaluating performance related tostrength reduction.
✓ However, evaluating only strength reduction
is not sufficient for evaluating theperformance of waste bodies (deterioration).

✓ It is believed that an evaluation method isrequired for verifying the impact of change inproperties that affects changes in the mineral
facies that contributes to the confinement
function of solidified substances.
•

CracksPeelingAcid, alkali, etc.

Glass

Testing standards,
etc. for reference

Performancedegradation

A testing method believed to be likely toaffect changes in mineral facies was
identified by focusing on the effects offactors leading to change in properties.
✓ Heat curing test
✓ Efflorescence test•

Further, the following testing methods thatenable quantification of acceleration basedon JIS standards were identified forevaluating the impact on strength
reduction.
✓ Freezing-thawing test
✓ Accelerated
carbonation test

©International Research Institute for Nuclear Decommissioning240

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Selection of the methods of accelerated testing ➢ The following 4 tests were identified and implemented as tests for verifying the long-term property change behavior of solidified substances (Table 1)• In addition, the anticipated factors leading to change in properties were reviewed considering the properties of 1F waste, the treatment methods,
etc.

Table 1 Selected testing method
Type of test

Details of anticipated change in properties

Anticipatedfactors
leading tochange inproperties

Freezing-thawing test

Mixing using comparatively greater amount of water is required for the lowtemperature solidification treatment of waste resulting from water treatment (mainlyslurry). The integrity (strength, etc.) of the solidified substances is likely to largely
reduce in response to the effect of freezing and thawing when the storage
environment of the solidified substances is inland, in a cold region, in indoor areas
where air-conditioning doesn’t work, in outdoor areas.

Temperature
and waterStrength reduction

Available(JIS A 1148)

Accelerated carbonation
test

If the integrity of the container in which solidified substances are stored is lost, theproperties of solidified substances are likely to change in response to the impact ofcarbon dioxide gas in the atmosphere.

Carbon dioxidegas

Strength reduction

Available(JIS A 1153)

Heat curing test

The amorphous and crystalline phases that are formed as a result of the impact ofwater supplied from outside the solidified substance or the change in temperatureduring long-term storage, undergo change in phase, due to which the properties ofthe solidified substances are likely to differ largely.

Temperature
and water

Change in mineral
facies phase

Not available*

Efflorescence test

If the containers in which solidified substances are stored lose their integrity, themoisture content in the solidified substances gradually evaporates due to which theelements in the solidified substance are likely to move to the surface of thesolidified substance and in addition a change in phase is likely to occur.

Water

Mass transfer
Change in mineral
facies phase

Not available*

References
Equilibrium solubility
test(Implementation in
FY2020)

It is believed that leaching from the solidified substance depends on the amount ofmoisture content that comes in contact per unit of time. Leaching of water solublecomponents is expected to be accelerated when the amount of moisture content
that comes in contact with respect to weight of solidified substance (solid to liquidratio) is changed.

Water

Leaching of watersoluble components

Not available(There are someideas)

Evaluation
indicator

Method ofquantifying
acceleration*Although it has not been established as a method of accelerated testing of degradation (at this point intime there is no equation for evaluating and calculating the acceleration), it is presumed to be similar tothe property change mechanism when cement solidified substances are stored, and is believed to benecessary for evaluating the performance of solidified substances (change in mineral facies).
©International Research Institute for Nuclear Decommissioning241

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Freezing-thawing test 1. Overview and test conditions ➢ Freezing-thawing test: Test for evaluating the internal damage of concrete caused due to frost damage. (JIS A 1148)
Table 1 Overview of the freezing-thawing test
Factors
leading tochange inproperties

Evaluation items

Temperature

Frost damage
resistance(Frostdeterioration)

Criteria

Dimensions ofsolidified substance[mm]

Material
age at the
beginning

Measurement items during
each cycle

End period

Mutual elastodynamic coefficient
60% or more
(Refer to Table 2)

40×40×160

28 daysFirst resonance frequency
Ultrasonic propagation time(Every 30 cycles*)

300 cycles

Sealed curing 28 days
10°C

Table 2-20°CFirst resonance frequency andultrasonic propagation time ismeasured every 30 cycles

40×40×160

Ultrasonic propagation time
Measurement ofcompressive strength
MIP Measurement*Measured at the 5th cycle and 15th cycle only immediately after the test is started,and thereafter measured every 30 cycles.

➢ Freezing-thawing test conditions

300
CyclesMeasurement items
pertaining to samples
after the tests areconcluded

Test conditions
・Freezing in air and thawing in air
・Freezing process: 2 hours(10℃ → -20℃ in 45 minutes, -20℃ retained for
75 minutes)
・Thawing process: 2 hours(-20℃ → 10℃ in 45 minutes, -20℃ retained for
75 minutes)

Ultrasonic propagation
speed
Compressive strength
Ultrasonic propagation
speed
MIP Measurement
Compressive strength: Number of tests 2(variation control)
MIP Measurement: Number of tests 1

Placement of mutual elastodynamic coefficient in academic conferences

Name of literature

Architectural Institute of Japan
Recommendations for Durability
Design and Construction Practiceof Reinforced Concrete Buildings
Japan Society of Civil Engineers
Standard Specifications forConcrete Structures

Mutualelastodynamic
coefficient
60% when cracks areat a critical stage

60% at the minimum
limit value

Additional Notes

The concrete deterioration indicator when cracks
are formed that give rise to safety problemsa s s o c i a t e d wi t h u s e , i s called a s m u t u a lelastodynamic coefficient.
Cross-sectional thickness about 20[cm] or more.
(When freezing and thawing is frequently repeated,
and the structures are exposed as usual)Fig. 1 Freezing-thawing test
equipment and test status©International Research Institute for Nuclear Decommissioning

242

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Freezing-thawing test 2. Calculation of acceleration ➢ Investigation of acceleration in the freezing-thawing test

The extent to which test conditions specified in JIS A 1148 “Method of test for resistance of concrete to freezing and thawing” are accelerated ascompared to the conditions in actual environment was calculated based on the “Recommendations for Durability Design and Construction Practiceof Reinforced Concrete Buildings” of the Architectural Institute of Japan.
➢ Results of investigating acceleration
In the “Recommendations for Durability Design and Construction Practice of Reinforced Concrete Buildings, 2016 edition” the following method has been described forverifying performance against frost damage.
• The limit state of the cracks due to frost damage is set in terms of the mutual elastodynamic coefficient.
• When the mutual elastodynamic coefficient reaches 60% that state is set as the design limit state. (Design limit: The state when there are safety problems
associated with use)
• The number of years in which cracks reach the limit is obtained using the equation below.

𝑆𝐿𝑑 = 𝐶𝑦𝑛 Τ𝐶𝑦𝑝

SLd: The number of years in which cracks reach the limit [years]
Cyn: Number of test cycles when the mutual elastodynamic coefficient of the limit state is reached according to the ASTM C 666 (JIS A 1148)
A standard [cycles]*1： ASTM：American Society for Testing and Materials
Cyp: Number of annual ASTM*1 equivalent cycles *2 [cycles/year]*2： Number of cycles expressing the effect of freezing and thawing on concrete over a
period of 1 year under the regional weather conditions.
About Cyp
About Cyn
It is assumed that information is obtained using reliable material from meteorological data of theUnder normal conditions, test results are used in principle, but in this study C yn isconstruction site.
⇒ Cyp from the literature of Hasegawa et al, from where the durability design and construction
assumed for calculating acceleration.
practices are cited, is used.
⇒ Assumed to be 300 cycles* (= Cycle when test is concluded according to the JIS A
1148 standards)
C : 6.3 in Fukushima (The region is set to Fukushima)yp

𝑆𝐿𝑑 = 300Τ6.3 = 47.6
≒ 48 years

Number of years in which cracks
reach the limit in Fukushima
48 years

Takuya Hasegawa and others; Method of estimating degradation of concrete due to frost damage andcomparison of service life based on the difference in climatic conditions (Proceedings of the Japan
Concrete Institute, Vol. 37, No. 1, 2015, pp859-864)

Time required for 300 cycles: 1200 hours= 50 days≒ 2 months

Calculated considering
1 cycle of 4 hours

Time required for 300 cycles of thefreezing-thawing test
Approx. 288X speed
2 months✓ When the in-water freezing-thawing test method is adopted and a solidified substance that reaches 60%
mutual elastodynamic coefficient in 300 cycles is used, the 300 cycle freezing-thawing test is equivalent
to the 48 year (288 times) period in Fukushima.

Temperature at the central part ofthe specimens
供試体中心部の温度*Since according to the guidelines of the Architectural Institute of Japan, Japan Society of
Civil Engineers, etc. “Concrete for which the mutual elastodynamic coefficient is 60% orlower until 300 cycles are completed must not be used in structures”, Cyn was assumed
to be 300 cycles.

Thawing
融解行程Freezing
凍結行程

process

process

1 cycle: More than 3 hours less
1サイクル：3時間以上4時間以内
than 4 hours (According to the JIS(JIS
A 1148による規定)
A 1148standards)

時間 cycle and JIS A
Fig. 1 Freezing-thawing
1148 standards©International Research Institute for Nuclear Decommissioning243

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Freezing-thawing test 3. Test results ➢ Results of the freezing-thawing test

Table 1 Results of the freezing-thawing test
Evaluation
items

Measurement items
Sample name

External observation

First resonancefrequency

Ultrasonic propagation time
Ultrasonic propagation rate

Strength

MIP

DamagedMutual elastodynamic
coefficient

Changes and property changes from the initialsampleFrost damage
resistance

Number ofyears in whichcracks reachthe limit

Precipitate

Cracks

Scaling*

Cement base material

－

－

－

－

✓

✓ (100%)No change (107%)

No major change＋＋＋

48 years or
more

Cement + CS

＋

＋

＋

＋

✓

✓ (86%)

No change (86%)

No major change＋＋＋

48 years or
more

Cement + IS

－

＋

－

－

✓

✓ (94%)

No change (102%)

No major change＋＋＋

48 years or
more

AAM(M) base material

＋

＋

－

＋

x

× (59%)- (14%)

Increase in number of poresof size 10-2 to 100＋

3 years

AAM (M) + CS

＋

－

＋

－

✓

✓ (91%)＋ (245%)Decrease in number ofpores of size 10-2 to 100＋＋

48 years or
more

AAM (M) + IS

＋

－

－

－

✓

✓ (77%)- (33%)

Increase in number of poresof size 10-1to 100＋＋

48 years or
more

AAM(MB40) base material

＋

－

－

－

x

✓ (97%)- (34%)

Increase in number of poresof size 10-2to 100＋

43 years

AAM(MB40) + CS

＋

－

－

－

✓

✓ (71%)△ (74%)

Increase in number of poresof size 10-1to 100＋＋

48 years or
more

AAM(MB40) + IS

＋

－

－

－

✓

✓ (71%)- (25%)

Increase in number of poresof size 10-2to 100＋＋

48 years or
more✓: 60% or more
×: 60% or less✓: 60% or more
×: 60% or lessValue in ( ) indicates lowest
value
of
the
ultrasonicpropagation speed rate during
the period＋: Increase△: Reduction ofless than 50%－: Reduction of
50% or more－

Number of ＋:
Extent of frost
damageresistance－Description of the signs andsymbols＋: Formed－: Not formed*Scaling ・・・ A frost damage phenomenon because of which concrete deteriorates, and in which there is peeling on the surface due to the effect of freezing and thawing.
✓

Cement based solidified substance : There was little change in properties (strength reduction and change in porosity) due to freezing and thawing in all cases✓

AAM(M) based solidified substance: Strength of the base material reduced. However, it is presumed that AAM(M+CS) and AAM(M+IS) are unlikely to be affected as compared to the base material.

✓

AAM(MB40) based solidified
substance✓

Freezing-thawing test

: Strength of the base material reduced. However, it is presumed that AAM(MB40+CS) and AAM(MB40+IS) are unlikely to be affected as compared to the basematerial.
: Change in properties of solidified substances due to freezing and thawing can be accelerated quantitatively (equivalent to 48 years (288 times) in the Fukushimaregion, by means of the 300 cycle test where the mutual elastodynamic coefficient is 60%)) and the cracks in the solidified substance and strength reduction canbe evaluated.
©International Research Institute for Nuclear Decommissioning244

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Accelerated carbonation test 1. Overview and test conditions ➢ Accelerated carbonation test: Test for evaluating degradation of material by accelerating carbonation of concrete due to carbon dioxide
gas in atmosphere (JIS A 1153)
Table 1 Overview of the accelerated carbonation test
Factors leading
to change inproperties

Carbon dioxide
gas

Evaluation items

Dimensions ofsolidified substance[mm]

Evaluation of the impact
of carbon dioxide gas inatmosphere

Material age at the
beginning

Acceleration
period[weeks]Measurement items

20×20×50

Carbonation depth
28 days old(Including 20℃, 60%RH
drying)(Carbonation, drying)

0*1, 1, 4, 8

➢ Accelerated carbonation test conditions－*2 Measured in addition to the JIS standards

Measurement of carbonation depth

Accelerated carbonation test conditions
Start of acceleration(After 28 days of sealed curing)・Temperature: 20°C
・Relative humidity: 60%
・Carbon dioxide
concentration: 5%

20×20×30

AAM(MB40) basematerial
with cracks all over

MIP measurement (Measured at 0, 1, 4, 8
weeks)XRD measurement (Measured at 0, 1, 4, 8
weeks)EPMA measurement (Measured at 8 weeks)

Ultrasonic propagationtime*2Compressive strength*2

20×20×30*1 Initial sample before accelerated carbonation

20×20×50

Measurement items pertaining to samples
after the test is completed *2← Cracks
Cement basematerial

Fig. 1 Status of formation of cracks in the carbonated cross section

Phenolphthalein
spray

Cracking at thecenter••Carbonation depth•MIP
XRD (Solidified substance,
precipitate)
EPMA (8 weeks)• Ultrasonic propagation speed• Compressive strength test
• Carbonation depth

Before spraying
phenolphthalein solution

After spraying
phenolphthalein solution

Fig. 2 Status of cracked surface before and after spraying phenolphthalein solution(AAM(MB40) base material
Acceleration period 4 weeks)©International Research Institute for Nuclear Decommissioning

245

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Accelerated carbonation test 2. Calculation of acceleration -➢ Investigation of the extent of acceleration under the accelerated carbonation test conditions➢

The extent to which test conditions under the accelerated environment specified in JIS A 1153 “Method of accelerated carbonation test for concrete” are accelerated as compared to theatmosphere was calculated based on the “Recommendations for Durability Design and Construction Practice of High Durability Reinforced Concrete Structures” of the Architectural Instituteof Japan and past research (Abe and others, 1990) that formed the basis of the test method.
Michihiko Abe and others; Research related to evaluation of theaccelerated carbonation test method for concrete,
Results of calculating acceleration(The Architectural Institute of Japan's Journal of Structural and Construction
Engineering , Vol. 409, March 1990, pp 1-10)

JIS Conditions
Dimensions of specimen: 10 x 10 x 40 [cm]Curing method: 20℃ underwater
Curing period: 1 day in the mold + 27 days
underwater
Drying period: 28 days (20℃, 60%)

If the carbonation depth under these conditions is expressed

Accelerated test temperature: in terms of period ・・・
20℃
Accelerated test humidity: 60%
CO2 gas concentration: 5%

𝑪=𝑨 𝒕

C: Carbonation depth
t: time period
A: Carbonation acceleration
coefficient

Carbonation
depth

As a result of conducting the test by changing the CO2 concentration,
Abe et al proposed a formula for calculating the carbonation depth under environmental conditions
other than 5% CO2 concentration.
𝑪 = 𝑨 𝑪𝑶𝟐 Τ𝟓 𝒕
When CO2 concentration is 5% (= standard condition), 𝑪𝑶𝟐 Τ𝟓 = 1⇒ 𝑪 = 𝑨 𝒕 is true

PeriodFig. 1 Relation between period
and carbonation depth

Conversion of the accelerated carbonation test conditions
Using the above method of calculation as reference, the period required for the carbonation depth in atmosphere to change to accelerated carbonation
depth at 8 weeks of acceleration under the JIS conditions was calculated. The carbon dioxide gas concentration in atmosphere was considered to be
0.03% indoors and about 0.1% outdoors (assuming that all other conditions remain constant)
Carbonation depth under JIS

𝑪=𝑨 𝒕

conditions:
Carbonation depth in atmosphere: 𝑪 = 𝑨 𝑪𝑶𝟐 Τ𝟓 𝒕′

𝑨 𝒕 = 𝑨 𝑪𝑶𝟐 Τ𝟓 𝒕′
0.03% outdoors
0.1% indoors

Outdoors: 𝑡′ = 167𝑡
(Period was 167 times of that
under JIS conditions)
Indoors: 𝑡′ = 50𝑡(Period was 50 times of that
under JIS conditions)✓ If the carbonation depth at 8 weeks of acceleration period is converted to carbonation depth in atmosphere, it is equivalent to about 26 years (outdoors) and about 8 years (indoors).

167 times

50 times©International Research Institute for Nuclear Decommissioning246

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Accelerated carbonation test 3. Test results (Cement based solidified substances) ➢ Results of the accelerated carbonation test
Table 1 Results of accelerated carbonation test of cement based solidified substances (up to 8 weeks of acceleration period)
Measurement items
External observation
Sample nameEvaluation items

Changes and change in properties from the initial sample (Accelerated carbonation period: 0 weeks)

Precipitate

CracksCarbonation
Depth (8 weeks)Strength

XRD

MIP

EPMA

Cement base material

－

＋

4.9mm

＋ (172%)

Calcite formation

No major change

Ca, SI, Al: Decrease in concentration
up to 5 mm from the surface＋＋

Cement + CS

－

－

1.3mm

＋ (137%)

Belite peak reduction

No major change

Na: Increase in surface concentration
Cl: Decrease in surface concentration,
increase in concentration from 5mm
and deeper

＋＋＋

＋＋

Number of ＋:
Extent ofcarbonation
resistance－

Cement + IS

Description of thesigns and symbols－＋: Formed－: Not formed＋: Formed－: Not formed

6.8mm＋ (139%)Belite peak reduction
Disappearance ofHaliteNo major change

Na: Increase in concentration up to 3
mm from the surface
Cl: Decrease in concentration up to 3
mm from the surface, increase inconcentration from 3mm and deeper－＋: Increase△: Decrease of less than50%
ー: Decrease of 50% or more
Rate of change from theinitial sample is indicated by
the numerical value in ( )

－

－

－

Increase in Na concentration up
to about 3 mm from the surface

Initial[Na]

Outdoors: 26 years(Accelerated 167
times)
Indoors: 8 years(Accelerated 50 times)－

Increase in Cl concentration
at 3 mm or deeper
Decrease in Cl concentration up to
about 3 mm from the surface

Initial

8w

Carbonation
resistance

Period anticipatedunder actual
environment
equivalent to 8
weeks ofacceleration period

Carbonation depth 6.8 mm

Carbonation depth 6.8 mm

[Cl]

8w

Fig. 1 Some of the EPMA mapping images and concentration distribution (Cement + IS: Samples with the deepest carbonation and element mapping images withsignificant fluctuations are selected)✓ Cement based solidified substance: Strength increased and changes in the mineral facies phase were observed in thecase of base material, CS mixture as well as IS mixture.
©International Research Institute for Nuclear Decommissioning

247

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Accelerated carbonation test 4. Test results (AAM (M) based solidified substances) -➢ Results of the accelerated carbonation test

Table 1 Results of accelerated carbonation test of AAM(M) based solidified substances (up to 8 weeks of acceleration period)
Measurement items
External observation

Sample name

Precipitate

AAM(M) base material

Cracks－

Evaluation items

Changes and change in properties from the initial sample (Accelerated carbonation period: 0 weeks)－

Carbonation
Depth (8 weeks)Strength

No change
(88%)

4.1mm

Carbonation
resistance

XRD

MIP

EPMA

No major change

Increase in numberof pores of size 10-2
to 10-0Si:
Increaseinconcentration up to 5mmfrom the surface
Na:
Increaseinconcentration up to 5mmfrom the surface

＋＋＋

＋＋

Outdoors: 25 years(Accelerated 167
times)AAM(M+CS)＋(Nahcolite)－

0.92mm＋ (201%)

Northupite formed

Decrease in porosity
Decrease in numberof pores of size 10-2
to 100Ca, SI, Al, Mg: Decreasein concentration up to2mm from the surface
Na, Cl: Increase inconcentration up to 2mmfrom the surface(Formation of precipitate)

AAM(M+IS)＋(Halite)－

1.2mm

No change (105%)

No major change

No major change

Na, Cl: Increase inconcentration up to 1mm
from the surface(Formation of precipitate)＋＋＋－＋: Increase△: Decrease of less than50%
ー: Decrease of 50% or
moreRate of change from theinitial sample is indicatedby the numerical value in ( )

－

－

－

Number of ＋:
Extent ofcarbonation
resistance＋: Formed－: Not formed

Description of the signs
and symbols＋: Formed－: Not formed

IncreaseinSi
concentration up to
about 5mm from thesurfaceIndoors: 8 years(Accelerated 50 times)－

IncreaseinNa
concentration up to
about 5mm from thesurface

Initial

InitialCarbonation depth 4.1mm[Si]

Period anticipatedunder actual
environment
equivalent to 8
weeks ofacceleration period

8w

[Na]

8w

Carbonation depth 4.1mm

Fig. 1 Some of the EPMA mapping images and concentration distribution (M base material: Samples with the deepest carbonation and element mapping images withsignificant fluctuations are selected)✓ AAM(M) based solidified substances: There was no change in strength and mineral facies phase in the case of base material and IS mixture. Strength increased andchanges in the mineral facies phase were observed only in the case of CS mixture.
©International Research Institute for Nuclear Decommissioning

248

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Accelerated carbonation test 5. Test results (AAM (MB40) based solidified substances) ➢ Results of the accelerated carbonation test
Table 1 Results of accelerated carbonation test of AAM(MB40) based solidified substances (up to 8 weeks of the acceleration period)
Measurement items
External observation
Sample name
Precipitate

CracksCarbonation
Depth (8 weeks)Strength

XRD

MIP

EPMA

―

＋

35.6mm- (41%)

Disappearance of C-SH
Formation of
Aragonaite

No major change

Ca: Decrease in concentration in the
carbonated part
Na: Increase in surface concentration＋

Formation of Faujasite

Increase in numberof pores of size 10-2
to 10-1

Ca, SI, Al, Mg: Decrease inconcentration up to 7mm from thesurfaceNa,
Cl:
Increaseinsurfaceconcentration,
decreaseinconcentration up to 1 to 5mm from thesurface＋＋＋＋

Number of ＋:
Extent ofcarbonation
resistance

AAM(MB40+CS)―

AAM(MB40+IS)＋＋: Formed－: Notformed

―

6.9mm

―

Ca (mass%)

Initial

5.3mm＋ (123%)

Formation of Faujasite
Disappearance ofHaliteIncrease in numberof pores of size 10-2
to 10-1－＋: Increase△: Decrease of less than50%
ー: Decrease of 50% or
moreRate of change from theinitial sample is indicatedby the numerical value in
()

－

－

－

MB40

15
InitialCarbonation 8w

Carbonation depth
35.6mm10

Initial

5

✓
✓

8w

0

0

5

10 15 20 25 30 35 40
Distance from exposed face (mm)

45

50

8w

－

InitialCarbonation 8w

Increase in Na
surface concentration

10

5

0

[Na]

Outdoors: 26 years(Accelerated 167
times)
Indoors: 8 years(Accelerated 50 times)

MB40

Decrease in Ca concentration(Carbonated part)[Ca]

Period anticipatedunder actual
environment
equivalent to 8
weeks of theacceleration period

Carbonation
resistance

Ca: Increase in concentration up to 3 to
5 mm
Si, Al: Increase in concentration up to7mm from the surface
Na,
Cl:
Increaseinsurfaceconcentration,
decreaseinconcentration up to 1 to 5mm from thesurface＋: Formed－: Not formed

15

No change(107%)Na (mass%)

AAM(MB40) base material

Description of thesigns and symbols

Evaluation items

Changes and change in properties from the initial sample (Accelerated carbonation period: 0 weeks)

Carbonation depth
35.6mm0

5

10 15 20 25 30 35 40
Distance from exposed face (mm)4550Fig. 1 Some of the EPMA mapping images and concentration distribution (MB40 base material: Samples with the deepest carbonation and element mapping images with significant
fluctuations are selected)
AAM(MB40) solidified substance: Cracks were formed on the base material. Zeolite mineral (Faujasite) was formed in the CS mixture and IS mixture.
Accelerated carbonation test
: This implies the possibility of being able to evaluate the strength of solidified substances and changes in the mineral facies by quantitatively
accelerating change in properties (carbonation due to carbon dioxide gas). (It is necessary to check whether it is the same as the status of change
in properties due to carbonation in actual environment)©International Research Institute for Nuclear Decommissioning

249

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Heat curing test 1. Overview and test conditions ➢ Heat curing test

Table 1 Overview of the heat curing test

Factors
leading tochange inproperties

Evaluation items

Dimensions of solidified
substance [mm]Material age at the
beginning

Temperature [℃]CuringMeasurement items

Test period [weeks]

Temperature
and water

Mass transfer, recombination

50×50×20

28 days20
50
80

SealedHumid

MIP, XRD
SEM/EDS4, 8

Sealing bag

Wrap

Sealed curing

Humid curing• XRD Measurement

Sealed curing• MIP Measurement• SEM/EDS

20°C
Status of heat curing test

H

Br

Fr C

H

P
Fr

Br

Fr C

H

P
Fr

Br

Fr

C

H

P
Br

5

10

15

C

20

C

P

H

P

Fr Be

C

H P

P

Fr Be

C

H P

C

C

30

C

Br
C

4

C

Be

C

C

C

C

C

H P
CC

H PC
C

35

40

Br
P
Be

H

P
CC

H

C

Br

H C BrP

H C Br

Fa

Br

Fa

Br
P

CC

P

50

2

C : Calcite [CaCO3], CSH: C-S-H
Fr : Friedel's salt [3CaO・
Al2O3・
CaCl2・
10H2O], H: Halite [NaCl]
K : Katoite [3CaO・
Al2O3・
SiO2・
4H2O], P: Portlandite [Ca(OH)2]

K HC

55

60

5

Pi Br
T

H

Pi
C
PiPiPi
Pi T
C

Br Pi
Br
H
PiC PiPi An C
PiPi Pi Br
HPi
Pi
C Pi Br
C
Pi
PiPi
Pi Br
H Br
T
Pi
Pi PiBrPi
Pi C
Pi
Pi An CTPi C An H
Br
Pi H Pi
Fa
H
Pi Pi
Fa
Pi
C
Br Pi
Pi Pi
Br
C
Br
Pi
Fa PiPi BrFa
PiPi Fa T
Pi
HC
Pi Fa
Pi FaPiC
Fa Fa
An H
An
Pi
Br T
Pi HPi Pi Pi
Pi H
Fa Sd Fa
Pi
Pi Pi
Br
C PiPiSd
Br
PiC
Pi Br
Pi Br Pi
Pi Fa T
Pi Fa Pi
Fa
Sd Fa
Fa C
An C
T
An H
Pi
Fa Fa
Pi

P

Br
P

K

45

H

P

Br
P
Be

P
C

Formation of2θ (deg.Cu Kα)
Katoite Be: Belite [Ca SiO ], Br: Brucite [Mg(OH) ]
2

Be

Br
C

Br
K

H

25

Br

Fr Be

P
K

Zeolite formation
due to crystallization of
HBr
C
N-A-S-H
Pi Pi Pi60

M+CS_Initial
M+CS_20°CSealed_4W
M+CS_50°CSealed_4W
M+CS_80°CSealed_4W

10

15

Pi Br

20

Pi Pi
Pi C An H

25

H

30

35 Fa 404550

2θ (deg.Cu Kα)55

60

Cumulative Porosity (%)

Intensity (a.u.)

P
Fr

P

80°C

*SEM/EDS is measured by
selecting samples based on
the XRD results

MIP
M+CS

C

P

50°C

XRD

OPC+CS_Initial
OPC+CS_20°CSealed_4W
OPC+CS_50°CSealed_4W
OPC+CS_80°CSealed_4W

Disappearance of
AFm phase

20°C

60
OPC+CS_Initial
OPC+CS_20°CSealed_4W
OPC+CS_50°CSealed_4W
OPC+CS_80°CSealed_4W

50
40
30

No majorchange

20
10
0
10-3

10-2

10-1
100
101
102
Average Pore Diameter (μm)

An: Anatase [TiO2], Br: Brucite [Mg(OH)2], C : Calcite [CaCO3]
Fa : Faujasite [Na2CaAl4Si10O28·20H2O]
H : Halite [NaCl], Pi : Pirssonite [CaNa2(CO3)2(H2O)2]
Sd : Sodalite [Na4Si3Al3O12Cl], T: Thermonatrite [Na2CO3・
H2O]

Cumulative Porosity (%)➢ Results of the heat curing test
OPC+CS

80°C

Humid curing

Intensity (a.u.)

Fig. 1

50°C

103

SEM

M+CS_Initial
M+CS_20°CSealed_4W
M+CS_50°CSealed_4W
M+CS_80°CSealed_4W

50
40
30

Changes in the

20 pore structure due

to crystallization
10 of N-A-S-H
0
10-3

10-2

Decrease in pores
formed due to mixingof carbonate slurry

10-1
100
101
102
Average Pore Diameter (μm)

103

Fig. 2 Measurement results of the heat curing test (Cement + CS, AAM(M+CS);
Test period 4 weeks)•

The heat curing test gave rise to changes in the mineral facies in thesolidified substances and in the pore structure.
⇒ The results of the heat curing test were consolidated by solidified
substance.

AAM(M+CS)_20℃
sealing_4weeks

Traces ofdisappearance ofneedle crystals
will be added inaccordance withthe products.
AAM(M+CS)_80℃ sealing_4weeks©International Research Institute for Nuclear Decommissioning

250

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Heat curing test 2. Test results (Cement based solidified substances) -➢ Results of the heat curing test (Cement based solidified substances)
Table 1 Results of heat curing test of cement based solidified substances (XRD, MIP Curing period of 8 weeks)
Samplename

Sealed20 [℃]

Cementbasematerial

Cement +CS

Cement +
IS

ー

ー

ー

Humid

50 [℃]

80 [℃]

20 [℃]

50 [℃]

80 [℃]

ー・Monosulfate that is the AFm phase
disappeared・Thermodyanmically stable Katoite wasformed.
・Size of the pores increased due to drying,and number of pores that were of the size 100[µm] or less increased

・Number of poresof the size 10-1 to
100 [µm]
decreased due tohydration

・Number of pores of thesize 10-1 to 100 [µm]decreased due tohydration

・Monosulfate that is the AFm phase
disappeared・Thermodyanmically stable Katoite wasformed.
・Belite of clinker mineral disappeared.
・Number of pores of the size 10-1 to 100 [µm]decreased due to hydration

・ Belite of clinker
mineral disappeared・ Belite of clinker mineral disappeared.
・Friedel’s salt that is the AFm phase
disappeared.
・Thermodyanmically stable Katoite wasformed.

・Number of poresof the size 10-1 to
100 [µm]
decreased due tohydration
・Number of poresof the size 10-2[µm] or lower
increased due tostructural change
in C-S-H.

・ Belite of clinker
mineral disappeared.
・Number of pores of thesize 10-1 to 100 [µm]decreased due tohydration
・Number of pores of thesize 10-2 [µm] or lower
increased due tostructural change in CS-H.

・ Belite of clinker mineral disappeared.
・Friedel’s salt that is the AFm phase
disappeared.
・Thermodyanmically stable Katoite wasformed.
・Number of pores of the size 10-1 to 100 [µm]decreased due to hydration
・Number of pores of the size 10-2 [µm] orlower increased due to structural change in CS-H.

・ Belite of clinker
mineral disappeared.
・Thermodyanmicallystable Katoite wasformed.

・ Belite of clinker mineral disappeared.
・Friedel’s salt that is the AFm phasedecreased.
・Thermodyanmically stable Katoite wasformed.
・Size of the pores increased due to drying,and number of pores that were of the size 100[µm] or less increased・ Belite of clinker
mineral disappeared.
・Thermodyanmicallystable Katoite wasformed.

・ Belite of clinker mineral disappeared.
・Friedel’s salt that is the AFm phase
disappeared.
・Thermodyanmically stable Katoite wasformed.

ー✓ The AFm phase (Monosulfate, Friedel’s salt) disappeared at 80℃ in all samples, and instead the thermodyanmically stable Katoite was formed.

✓ Katoite which was not formed in cement base material and cement + CS solidified substances even at 50℃ was formed in the case of cement + IS solidified substances.
⇒ Due to the increase in Fe content associated with mixing of waste, Fe solid solution Katoite that can form even at room temperature is likely to be formed.
✓ In the case of humid curing samples, since hydration of the samples progressed further, it was confirmed that the peak of Belite disappeared and the pore structurechanged.
(In the case of sealed curing samples as well, change in the peak of Belite and change in the pore structure were verified.)©International Research Institute for Nuclear Decommissioning251

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Heat curing test 3. Test results (AAM (M) based solidified substances) ➢ Results of the heat curing test (AAM(M) based solidified substances)
Table 1 Results of heat curing test of AAM(M) based solidified substances (XRD, MIP Curing period of 8 weeks)
Samplename

AAM(M)
basematerial

AAM(M+CS)AAM(M+IS)

SealedHumid

20 [℃]

50 [℃]

80 [℃]

ー・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100µm
increased.

・Faujasite wasformed・Pores of size
about 100 [µm]formed due to CS
mixing decreased(cause unknown)ー

20 [℃]

50 [℃]

80 [℃]

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.

ー・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 [µm]increased.
・Pores of size about 100[µm] formed due to CS
mixing decreased (cause
unknown)・Faujasite was formed
・Sodalite was formed
・Due to formation of
Sodalite (crystallization),
pores of size 10-1 [µm] orless decreased.
・Pores of size about 100[µm] formed due to CS
mixing decreased (cause
unknown)・Faujasite wasformed・Pores of size
about 100 [µm]formed due to CS
mixing decreased(cause unknown)

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 [µm]increased.
・Pores of size about 100[µm] formed due to CS
mixing decreased (cause
unknown)・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 [µm]increased.
・Pores of size about 100[µm] formed due to CS
mixing decreased (cause
unknown)・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.
・Pores of size about 10-1[µm] decreased (cause
unknown)・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.

・Faujasite was formed
・Due to formation of
Faujasite (crystallization),
pores of size 10-1 to 100[µm] increased.
・Pores of size about 10-1[µm] decreased (cause
unknown)ー✓ Due to heat curing, N-A-S-H in the matrix portion of the solidified substance crystallized, and zeolite minerals (Faujsite, Sodalite) were formed.
⇒ In some of the samples, the same zeolite minerals were found even at 20℃, due to which it is believed that the reaction got accelerated due toheat curing.
✓ The formed zeolite mineral differed depending on the compounding or curing conditions, and the change in pore structure also differed depending on
the formed zeolite mineral.
⇒ It implies that the formed zeolite mineral changes depending on the composition of the solidified substance, and the difference in its formation
process leads to change in pore structure.
©International Research Institute for Nuclear Decommissioning252

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Heat curing test 4. Test results (AAM (MB40) based solidified substances) ➢ Results of the heat curing test (AAM(MB40) based solidified substances)
Table 1 Results of heat curing test of AAM(MB40) based solidified substances (XRD, MIP Curing period of 8 weeks)
Samplename

Sealed20 [℃]

50 [℃]

AAM(MB40)
Basematerial

・C-S-H was formed

・C-S-H was formed

MB40
+CS

・Faujasite wasformed・Due to formation of
Faujasite(crystallization),
pores of size 10-1 to
100 [µm] increased.

・Faujasite wasformed・Due to formation of
Faujasite(crystallization),
pores of size 10-1 to
100 [µm] increased.

MB40
+IS

・C-S-H was formed
・Faujasite wasformed・Pores of size about
100 [µm] or moreincreased (cause
unknown)

・C-S-H was formed
・Faujasite wasformed・Pores of size about
100 [µm] or moreincreased (cause
unknown)

Humid
80 [℃]
・C-S-H was formed・Size of the pores increased due todrying, and number of pores that areof the size 10- 1 [µm] or less increased

・Faujasite was formed
・Sodalite was formed
・Due to formation of Sodalite(crystallization), pores of size 10-1[µm] or less decreased.
・C-S-H was formed
・Sodalite was formed
・Due to formation of Sodalite, pores ofsize 10-1 [µm] or less decreased
・Pores of size about 100 [µm] or moreincreased (cause unknown)・Size of the pores increased due todrying, and number of pores of size
about 10-1 [µm] increased.

20 [℃]

50 [℃]

80 [℃]

・C-S-H was formed

・C-S-H was formed

・C-S-H was formed
・Faujasite was formed

・Faujasite wasformed・Faujasite wasformed・Due to formation of
Faujasite(crystallization),
pores of size 10-1 to
100 [µm] increased.

・Faujasite was formed
・Due to formation of Faujasite(crystallization), pores of size 10-1 to
100 [µm] increased.
・Zeolite Na-P was formed

・C-S-H was formed
・Faujasite wasformed・Pores of size about
100 [µm] or moreincreased (cause
unknown)

・C-S-H was formed
・Faujasite wasformed・Pores of size about
100 [µm] or moreincreased (cause
unknown)

・C-S-H was formed
・Faujasite was formed
・Sodalite was formed
・Zeolite Na-P was formed
・Pores of size about 100 [µm] or moreincreases (cause unknown)✓ Due to heat curing, N-A-S-H in the matrix portion of the solidified substance crystallized, and zeolite minerals (Faujsite, Sodalite, Zeolite Na-P) were formed.
⇒ In some of the samples, the same zeolite minerals were found even at 20℃, due to which it is believed that the reaction got accelerated due to heat curing.
✓ The formed zeolite mineral differed depending on the compounding or curing conditions, and the change in pore structure also differed depending on theformed zeolite mineral.
⇒ It implies that the formed zeolite mineral changes depending on the composition of the solidified substance, and the difference in its formation processleads to change in pore structure.
Results of the heat curing test
✓ It was found in the heat curing test that the formation of mineral facies can get accelerated due to factors leading to change in properties(temperature and moisture content), compared to the basic conditions (factors leading to change in properties are not manipulated) with sealed
curing at 20℃.
©International Research Institute for Nuclear Decommissioning253

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Efflorescence test 1. Overview and test conditions -➢ Efflorescence test
Factors leading
to change inproperties

Water

Table 1 Overview of the efflorescence test

Evaluation items

Mass transfer due towater absorption

Efflorescence test
(Material age 28 days)

Dimensions of solidified
substance [mm]Measurement items

28 daysObservation of progress ofexternal appearance
Amount of precipitate
formation

50×50×20

Wind

Acceleration
period[weeks]Material age at the
beginning

1, 4, 8

・MIP
・XRD
・EPMA

Fig. 1

Evaluation of theamount ofprecipitate formed

Before the test

Solidified
substance

Solidified
substance

MIP measurement (Measured at 1, 4, 8
weeks)XRD measurement (Measured at 1, 4, 8
weeks)EPMA measurement (Measured at 8 weeks)

Testing conditions
・Temperature: 10°C
・Relative humidity: 70%
・Drying by means of ablower(Refer to past research)

Before coating

Measurement items pertaining to samples
after the tests are concluded

Status of efflorescence test

During the test
After the test (Collection(Formation of precipitate)of precipitate)
Precipitate
Solidified
Solidified
PrecipitateResinsubstance

substance

Resin

Resin

Water

Water

𝑅𝑎𝑡𝑒 𝑜𝑓 𝑝𝑟𝑒𝑐𝑖𝑝𝑖𝑡𝑎𝑡𝑒 𝑓𝑜𝑟𝑚𝑎𝑡𝑖𝑜𝑛 [%] =𝑄𝑢𝑎𝑛𝑡𝑖𝑡𝑦 𝑜𝑓 𝑝𝑟𝑒𝑐𝑖𝑝𝑖𝑡𝑎𝑡𝑒
× 100
𝑄𝑢𝑎𝑛𝑡𝑖𝑡𝑦 𝑜𝑓 𝑠𝑎𝑚𝑝𝑙𝑒 𝑏𝑒𝑓𝑜𝑟𝑒 𝑐𝑜𝑎𝑡𝑖𝑛𝑔

Schematic diagram of the change in weight of the solidified substanceFormation of precipitate on thesurface of the solidified substance

Cement + CS
(Acceleration period 1 week)
Fig. 2 Status of precipitate formation during
the prediction test (Photo showing AAM(M+CS)
under an actual microscope)(AAM(M) base material(Acceleration period 8 weeks)

AAM (M + IS)(Precipitate collection at 2 weeks)

AAM(MB40+CS)(Acceleration period 1 week)

Fig. 3 Example of status of formation of precipitate in this test
©International Research Institute for Nuclear Decommissioning254

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Efflorescence test 2. Test results (Cement based solidified substances) ➢ Results of the efflorescence test (Cement basedsolidified substances)
Table 1 Results of efflorescence test of cement based solidified
substances (TestMeasurement
period 8 weeks)
items
Samplename

Cementbasematerial

Cement +CS

Precipitate

CracksRate ofprecipitate
formation [%]Efflorescence

－

－

－

No major change

No major change

Ca: Decrease in concentration of top
surface 2 mm
Na: Increase in concentration of the top
surface 2 mm＋Heat curing
20℃ sealed－－

Not evaluatedNo major change

No major change

Not evaluatedNot evaluatedEfflorescence

＋

－

4.2

No major change

No major change

Na, Cl: Decrease in concentration (overall)(Due to formation of precipitate)＋＋＋

Heat curing
20℃ sealed－－

Not evaluatedNo major change

No major change

Not evaluatedNot evaluated＋Test name*1

External observation

Evaluation items

Changes and property changes from the initial sample
XRD (1, 4, 8 weeks)*2

MIP (1, 4, 8 weeks)*2

EPMA (Only 8 weeks)

Mass transfer due
to waterabsorption

Efflorescence

－

－

－

No major change

No major change

Na: Decrease in concentration in thesection that is 2 mm and deeper from the
top surface
Cl: Decrease in concentration at the top
surface and bottom surface

Heat curing
20℃ sealed－－

Not evaluatedNo major change

No major change

Not evaluatedNot evaluated＋: Formed－: Not formed＋: Formed－: Notformed－: Not formed－－

Top surface: dry surface
Bottom surface: immersed surface

Number of ＋:
Extent oftransferability

Cement +IS

Description of the signs andsymbols*1 The results of sealed curing at 20℃ during the heat curing test are provided as reference for comparison with efflorescence test.
*2 XRD and MIP measurement is carried out only at 4 and 8 weeks in the case of sealed curing at 20℃ during the heat curing test.

✓ In cement based solidified substances, precipitate (Thermonatrite, Halite, Gaylussite) formed only in the case of Cement + CS.
✓ The XRD and MIP results after the 8 week test period were not much different than the initial sample.
✓ According to the EPMA results, the concentration of Na and Cl changed (mass transfer) even in the case of base material and Cement + IS inwhich precipitate was not formed.
✓ Compared to sealed curing at 20℃ in the heat curing test in an environment without contact with water, since precipitate formed in the case of
Cement + CS, it is believed that the efflorescence test accelerates change in properties of cement based solidified substances.
©International Research Institute for Nuclear Decommissioning255

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Efflorescence test 3. Test results (AAM (M) based solidified substances) ➢ Results of efflorescence test (AAM(M) based solidified substances)
Table 1 Results of efflorescence test of AAM(M) based solidified substances (Test period 8 weeks)
Measurement items
Sample name

Precipitate

CracksRate ofprecipitate
formation [%]Efflorescence＋＋

Heat curing
20℃ sealed－

Efflorescence

Heat curing
20℃ sealed

Test name*1

External observation

Evaluation items

Changes and property changes from the initial sample

Mass transfer due towater absorption

XRD (1, 4, 8 weeks)*2

MIP (1, 4, 8 weeks)*2

EPMA (Only 8 weeks)

8.0

Formation and disappearance ofThermonatrite(Na dissolution in the basematerial)

Decrease in bulk density
Increase in porosity

Na: Decrease in concentration(overall)(Due to formation of precipitate)＋＋＋－

Not evaluatedNo major change

No major change

Not evaluatedNot evaluated*3＋

Not measured*3

Disappearance of Thermonatrite(Na dissolution in the base material
and in CS)

Decrease in bulk density
Increase in porosity

Na,
Cl:
Decreaseinconcentration (overall)(Due to formation of precipitate)

＋＋＋

－

－

Not evaluatedFaujasite was formed

Decrease in pores of size
about 100 [µm] formed due
to mixing of CS (cause
unknown)

Not evaluatedNot evaluatedSi, Al: Decrease in concentration(overall)Na,
Cl:
Decreaseinconcentration (overall)(Due to formation of precipitate)＋＋＋

AAM(M) basematerial

AAM(M+CS)

Efflorescence

＋

－

8.8

No major change

Decrease in bulk density
Increase in porosity

Heat curing
20℃ sealed－－

Not evaluatedNo major change

No major change

Not evaluatedNot evaluated＋: Formed－: Not formed＋: Formed－: Not formed－: Not formed－－

Top surface: dry surface
Bottom surface: immersed
surface

Number of ＋:
Extent oftransferability

AAM(M+IS)

Description of the signs and symbols*1 The results of sealed curing at 20℃ during the heat curing test are provided as reference for comparison with efflorescence test.
*2 XRD and MIP measurement is carried out only at 4 and 8 weeks in the case of sealed curing at 20℃ during the heat curing test.
*3 Since there was water up to the top surface of the solidified substance, precipitate was not found, but when the solidified substance is dried, a large quantity of
precipitate is formed.

✓ Precipitate was formed in all AAM(M) based samples. However, in the case of AAM(M+CS) precipitate was formed during the drying process after
the test was concluded.
(Thermonatrite precipitated in all AAM(M) based samples, further Gaylussite precipitated in AAM(M+CS) based samples and Halite precipitated in
AAM(M+IS) based samples.)✓ As compared to sealed curing at 20℃ in the heat curing test in an environment without contact with water, since it was found that precipitate wasformed in all samples, that Thermonatrite formed and disappeared, that porosity increased and bulk density decreased, it is believed that theefflorescence test accelerates change in properties of AAM(M) based solidified substances.
©International Research Institute for Nuclear Decommissioning256

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Efflorescence test 4. Test results (AAM (MB40) based solidified substances) ➢ Results of efflorescence test (AAM(MB40) based solidified substances)
Table 1 Results of efflorescence test of AAM(MB40) based solidified substances (Test period 8 weeks)
Measurement items
Samplename

AAM(MB40)
Basematerial

Precipitate

CracksRate ofprecipitate
formation [%]Efflorescence＋＋

Heat curing
20℃ sealed－

Efflorescence

Heat curing
20℃ sealed

Test name*1

External observation

Evaluation items

Changes and property changes from the initial sample

Mass transfer
due to water
absorption

XRD (1, 4, 8 weeks)*2

MIP (1, 4, 8 weeks)*2

EPMA(Only 8 weeks)

0.9

Formation of Aragonaite and Calcite(Formed due to Ca dissolution in the basematerial and carbonation)

No major change

Ca, Si, Al, Na: Decrease inconcentration＋＋－

Not evaluatedC-S-H

No major change

Not evaluatedNot evaluated

＋

＋

0.5

No major change

Increase in bulk density
Decrease in porosity－－

Not evaluatedFaujasite was formed

Due to formation(crystallization) of Faujasite,
number of pores of size 10-1
to 100 [µm] increased

Decrease in bulk density
Increase in porosity
Decrease in the number ofpores of size 10-2 to 101

AAM
(MB40
+CS)

Cl:
Decrease
concentration(Due to formation
precipitate)

in
＋＋

of

Not evaluatedNot evaluatedEfflorescence

＋

－

2.1

C-S-H (Formed due to the hydration reaction ofslag)
Calcite formation (Formed due to Ca dissolution
in the base material and carbonation)
Disappearance of Halite(Almost all the Cl in the solidified substance got
dissolved)

Heat curing
20℃ sealed－－

Not evaluated

C-S-H was formed
Faujasite was formed

Pores of size 100 [µm] or
more increased (cause
unknown)

Not evaluatedNot evaluated＋: Formed－: Notformed＋: Formed－: Notformed－: Not formed－－

Top surface: dry surface
Bottom surface: immersed
surface

Number of ＋:
Extent oftransferability

AAM
(MB40
+IS)

Description of the signs andsymbols

Na, Cl: Decrease
concentration(Due to formation
precipitate)

in
＋＋＋

of

*1 The results of sealed curing at 20℃ during the heat curing test are provided as reference for comparison with efflorescence test.
*F2 XRD and MIP measurement is carried out only at 4 and 8 weeks in the case of sealed curing at 20℃ during the heat curing test.
✓ It was found that precipitate formed in all samples as compared to sealed curing at 20℃ in the heat curing test in an environment without contact with water.
(Thermonatrite precipitated in all AAM(MB40) based samples, further Halite and Gaylussite precipitated in AAM(MB40+CS) based samples and Halite precipitated in
AAM(MB40+IS) based samples.)✓ Even though the generation phase changed and changes took place in the porosity and bulk density under sealed curing at 20℃ during the heat curing test,
formation of precipitate was not being confirmed. Hence, it is believed that the efflorescence test accelerated changes in properties of AAM(MB40) based solidified
substances.
©International Research Institute for Nuclear Decommissioning

257

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Efflorescence test 5. Test results (EPMA mapping images) ➢ Results of the efflorescence test (Example of EPMA mapping image)
1 sample each with large precipitate formation is selected from cement based, AAM(M) based and AAM(MB40) based solidified substances, and these are shown as some of the EPMA analysis result. (Table 1)
The initial samples were analyzed at a depth of 30 mm*, and the samples after an 8-week acceleration period were analyzed at a depth of 20 mm*2. The results are listed focusing on the Na and Cl contained in the
precipitates (Thermonatrite, Halite) found.

Table 1 Some of the EPMA mapping images and concentration distribution

Initial8w

Decrease in Na
concentration

Decrease in Cl concentration

8w

Initial

[Cl]

[Na]

Precipitate formed: Halite, Thermonatrite

EPMA results pertaining to Cement + CS

Decrease in Na concentration

Initial8w

Initial8w

Decrease in Cl concentration

[Cl]

[Na]

Precipitate formed: Halite, Thermonatrite

EPMA results pertaining to AAM(M)+IS

Decrease in Na concentration
Decrease in Cl concentration

Initial8w[Na]

Initial8w[Cl]

EPMA results pertaining to AAM(MB40)+IS
Precipitate formed: Halite, Thermonatrite*The initial sample was taken from a depth at which external impact is less (30 mm), whereas the sample after an 8 week acceleration period was takenfrom a depth at which changes appear in the EPMA mapping (20 mm).

©International Research Institute for Nuclear Decommissioning

258

(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Summary of accelerated testing Accelerated
testing
type

Freezing-thawing
test

Accelerated
carbonation test

Test results
Purpose of the tests
Cement
To observe changes inproperties due to changes
in the form of moisture in
the solidified substances
because of the effect offreezing and thawing, byassuming storage
environment of thesolidified substances to beinland, in a cold region,indoors where air
conditioning does not work,and outdoors.
To obtain information on
the solidified substanceproperty change behavior
in response to the impact
of carbon dioxide gas inatmosphere, assuming thatthe integrity of thecontainer in which solidified
substances are stored is
lost.

•
•

Heat curing test

Efflorescence
test

✓
✓

To observe the impact on
the solidified substances
when nuclides get
t r a n sf e r r e d t o t h e t o p
surface of the solidified
substances, assuming thatthe container in which thesolidified substances are
stored has lost its integrity
due to which the moisturein the solidified substances
gradually evaporates.

There was no major change inproperties due to freezing and thawing
in any of the cement based solidified
substances. Hence it is believed thatthe substances are frost damage
resistant.

••

Cracks formed on the AAM(M) basematerial in a short period. Hence its frost
resistance is low.
AAM(M) + CS showed a clear difference
in behavior than the other solidified
substances, but the cause was not clear.
It was presumed that AAM(M)+IS was
unlikely to be affected by freezing andthawing as compared to AAM(M) basematerial.

AAM(MB40)

•

•

•

•

The cement based solidified
substances were resistant tocarbonation even if those solidified
substances were mixed withsimulated carbonate slurry or
simulated iron co-precipitation slurry.

Cement hydrates containing Al
underwent phase change at 80℃
curing temperature.
This implies that there is interaction
between the simulated iron coprecipitation slurry and cement
hydrates in the solidified substances.
The change in mineral facies had
hardly any impact on the change inpore structure.

As the mineral composition or pore
structure of all of the cement basedsolidified substances remained almost
unchanged from the initial samples, it
is believed that there was little masstransfer inside the solidified
substances as compared to AAM(M)solidified substances and AAM(MB40)
solidified substance.

•

•
•
•

•

•

Applicability to evaluation of longterm property change behavior

AAM(MB40) base material was largely affected byfreezing and thawing because of its reduced
compressive strength and the behavior of themutual elastodynamic coefficient.
Although the impact on AAM(MB40)+CS andAAM(MB40)+IS was small as compared to
AAM(MB40), it is believed that they were affectedby freezing and thawing.

• Changes in properties can be
accelerated, and the cracking of thesolidified substance or decrease instrength can be evaluated.
• However, as water which is a medium
for diffusion gets solidified, it isbelieved that changes in the mineral
facies get suppressed.

Since carbonation due to cracks and decrease incompressive strength were conspicuous in
AAM(MB40) base material, it is believed that its
carbonation resistance is low.
The carbonation depth of AAM(MB40)+CS andAAM(MB40)+IS was 5mm or lower, and there wasno decrease in the compressive strength, it isbelieved that these solidified substances arecarbonation resistant.

• This implies the possibility of being
able to evaluate the strength ofsolidified substances and changes in
the mineral facies by quantitatively
accelerating change in properties(carbonation due to carbon dioxidegas). (It is necessary to check whether
it is the same as the status of change
in properties due to carbonation in
actual environment)

N-A-S-H crystallized due to heat curing, and zeolite minerals were formed (in some of the samples
zeolite minerals formed even at 20℃).
The formed zeolite minerals changed depending on the compounding or curing condition of thesolidified substance, and the change in pore structure also differed depending on the formed zeolite
mineral.
Multiple factors such as difference in composition of the matrix part, impact of waste, impact oftemperature, etc. compounded and gave rise to a complex change in the pore structure.

• It was verified by comparing with thereference (20℃ sealed curing) thatchange in properties (change inmineral facies) was accelerated due toheat curing.
• It was found that 2 factors, namely
temperature and moisture (vapor),contributed to the change in properties,

In the case of AAM(M) base materialand AAM(M)+IS, it was found that bulk
density decreased and porosity
increased due to the formation of
precipitate.
In the case of AAM(M)+CS, precipitate
did not form during the test, but it wasverified that precipitate formed when thesurface was dried.

• It was verified by comparing with thereference (20℃ sealed curing) thatchange in properties (change inmineral facies) was accelerated whenefflorescence test was conducted.
• It was found that moisture (pore water),
contributed to the change in properties.

•

•
•

•
To accelerate long termchange in properties of
A A M ma t e r i a l ( z e o l i t emineral formation, etc.)
through heating on a trial
basis, and observe thechange in properties ofsolidified substances.

AAM (M)

The carbonation depth of AAM(M)based solidified substances was small,and as major changes were not seen incompressive strength, it may be said
that these solidified substances arecarbonation resistant.

•

•

•
•

It was found the precipitate was formed in all
AAM(MB40) solidified substances.
The pore structure remained almost unchanged in
the AAM(MB40) base material.
Bulk density increased and porosity decreased in
the case of AAM (MB40)+CS, but in the case ofAAM(MB40)+IS bulk density decreased rapidly andporosity exponentially increased.

As it was confirmed during the accelerated carbonation test, heat curing test and efflorescence test that the mineral facies changed, there is apossibility of being able to accelerate long term change in mineral facies of the solidified substance and evaluate it.
It was found that change in mineral facies differed depending on the difference in factors (water, temperature) leading to change in properties or theway in which those factors influence the change in properties.
©International Research Institute for Nuclear Decommissioning

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(a) Establishment of selecting advance treatment methods [1] Low-temperature treatment techniqueiv. Investigation and evaluation of factors affecting long-term change in properties- Summary -Results so far✓ Accelerated testing that is expected to be applicable to the evaluation of long-term property change
behavior was investigated.
✓ With regards to the freezing-thawing test, changes in properties (freezing and thawing) can be accelerated
to bring about conditions equivalent to 48 years (288 times) in the Fukushima region, and the cracking of
the solidified substance or decrease in strength can be evaluated. However, as water which is a medium
for diffusion gets solidified, it is believed that changes in the mineral facies get suppressed.
✓ With regards to the accelerated carbonation test, it was implied that there is a possibility of being able toevaluate the strength of solidified substances and changes in the mineral facies by accelerating changes inproperties (carbonation due to carbon dioxide gas) 167 times (outdoors) and 50 times (indoors). (It isnecessary to check whether the status of change in properties in the case of AAM is the same as thatcaused by carbonation in actual environment)✓ With regards to heat curing test, it was verified that it was possible to accelerate changes in properties(change in mineral facies) by means of the factors leading to change in properties (temperature andvapour).
✓ With regards to efflorescence test, it was verified that it was possible to accelerate changes in properties(change in mineral facies) by means of the factors leading to change in properties (pore water).

Issues, etc.
✓ Methods for measuring the amount of moisture passing into the solidified substances need to bestudied.
✓ With regards to the efflorescence test and heat curing test, quantification of the extent ofacceleration needs to be studied by comparing the presence of acceleration and the extent ofdevelopment of property changes (change in mineral facies) over a long period of time.
©International Research Institute for Nuclear Decommissioning260

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment
technology
i. Organization of evaluation axis and acquired data for comparing technologies◼ Achievements up to FY2020• The concept of solidification treatment equipment was studied, data pertaining to economic efficiency such as facility configuration,
treatment efficiency, maintenance details, types of consumables and replacement frequency, types and quantity of secondary waste
generated from contaminated water treatment, etc. was acquired, and it was organized along with the data acquired on a trial basis,for evaluating waste body specifications of various treatment technologies.
• In doing so, a series of approaches were developed for ① Selecting the applicable solidification treatment based on thecharacteristics of waste, ② Selecting a solidification treatment technology that is more suitable to the requirements even if it is for thesame solidification treatment , ③ investigating the pro and cons of introducing the selected technology based on the technicalachievements of the selected solidification treatment technology, process information, and economic efficiency data.
• The following issue was identified in detailed development of the selection method.
✓ Expansion of data on an ongoing basis for evaluating the range of solid waste to which the treatment technologies areapplicable◼

Goal
To present methods for evaluating the treatment technologies that are likely to be applicable to various solid waste, and their scopeof application, by organizing the evaluation axis of the treatment technologies, their process data, etc. based on the variety ofacquired data, in order to evaluate the range of solid waste to which the treatment technologies are applicable.

Details of implementation
Constant updating of the evaluation methods (approaches)
Ongoing investigation of multi-faceted evaluation methods (approaches) for selecting the solidification technology forsecondary waste generated from contaminated water treatment
Reflection of data acquired this year
Updation by reflecting the findings from change in properties of solidified substances, long-term property change behavior,etc. due to heating, etc. acquired this year.

Indicators for the achievements of goal
Techniques (approach methods) for evaluating solidification technologies that are likely to be applicable tovarious solid waste are presented.
©International Research Institute for Nuclear Decommissioning

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[Investigation results up to FY2020][FY2021 Implementation items]• investigate of techniques for verifying the feasibility of solidification
by low-temperature treatment• Investigation of the change in properties of solidified material due toheating, etc.
• Evaluation of the relationship between the inventory of Cs, etc. andthe temperature of solidified substances• Investigation and evaluation of factors affecting long-term change
in properties• Investigation related to volume of Cs volatilization during hightemperature treatment and its control

➢

Eva
luat
ion
axis

Characteristics of solidified substances➢ Updating the FY2021 approach

Evaluation
axis(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
i. Organization of evaluation axis and acquired data for comparing technologies

ICV Joule heating(GeoMelt)

Evaluation sub-items

Plasma heating

Induction heating(Super-high frequency)

Cementin-drum

Cementout-drumAAM
in-drum

Has not been evaluated, but ispresumed to be highHas not been evaluated, but ispresumed to be highHas not been evaluated, but ispresumed to be highHomogeneity

There was no problem asviscosity at the treatment
temperature was controlled by
the composition (Except whenevaluating TiO2)

Melting solidified LLW was
found to be homogeneous.

There was stirring due to heat
convection andelectromagnetic force createdby electromagnetic induction.
After melting the entire
quantity, by securing about 1
hour or more of retention time,
the solidified substancebecame homogeneous.

Strength

・ Approx. 300Mpa （When it
was ensure that thesolidification material isborosilicate glass)Same as slag with CaO, MgO,
SiO2 as the main ingredients(about 20MPa or more)

Almost the same as the qualityof slag with CaO-MgO-SiO2
based components (SiO2 about
40wt%) or vitreous material

・ 7MPa (28 days, carbonate
slurry filling rate 30%)
・ 32MPa (28 days, iron coprecipitation slurry filling rate
20%)・ 7MPa (28 days, carbonate
slurry filling rate 30%)
・ 32MPa (28 days, iron coprecipitation slurry filling rate
20%)・ 5MPa or more (28 days,
carbonate slurry filling rate30%), (M, MB20, MB40)
・ 15MPa or more (28 days, ironco-precipitation slurry filling rate20%) (M, MB20, MB40)

Leaching resistance

NMLNa<0.2 g/m2/d
10－4 kg/m2/d order (In thecase of borosilicate glass)Same as slag with CaO, MgO,
SiO2 as the main ingredients(Estimated as 1E-7kg/m2/d(28d) or less (Cｓ, Sr))

Actual measurements are notavailable.
Estimated to be almost thesame as the quality of slag with
CaO-MgO-SiO2 based
components (SiO2 about
40wt%) or vitreous material.

Cs:95%, Sr:5%, Sn:≦0.1%.,
Ce:≦0.1% (Leaching rateevaluated assuming carbonate
slurry filling rate 30%,
ANS/ANSI-16.1)

Cs:95%, Sr:5%, Sn:≦0.1%.,
Ce:≦0.1% (Leaching rateevaluated assuming carbonate
slurry filling rate 30%,
ANS/ANSI-16.1)

Cs:23%, Sr:0.5%, Sn:5%,
Ce:0.6% (Leaching rateevaluated assuming carbonate
slurry filling rate 30% M
solidified substance,
ANS/ANSI-16.1)

Radiation resistance(Hydrogen G value
and property changeofsolidified
substance
organization)

Negligible

Estimated to be a level atwhich there is no particular
problem since it is a melting
solidification substance.

Estimated to be a level atwhich there is no particular
problem since it is a melting
solidification substance.

・ 0.16-0.17 (Carbonate slurry
simulated powder filling rate30%)

・ 0.16-0.17 (Carbonate slurry
simulated powder filling rate30%)

・ 0.13-0.28 (Carbonate slurry
simulated powder filling rate
30% M solidified substance)
・ 0.05-0.15 (Carbonate slurry
simulated powder filling rate
30% MB solidified substance)

Heat
resistance(Impact of thermalload on performance)

Negligible

Estimated to be a level atwhich there is no particular
problem since it is a melting
solidification substance.

Actual measurements are notavailable.
Heat resistance is estimated tobe high since it has the samecharacteristics as slag andglass.

Strength residual rate 90% ormore when dried for 28 days at
80℃

Strength residual rate 90% ormore when dried for 28 days at
80℃

Strength residual rate 20% orless when dried for 28 days at
80℃

Long-term stability

StableEstimated to be a level atwhich there is no particular
problem since it is a melting
solidification substance.

Almost the same as slag or
glass with that composition

Data on characteristics of solidified substances
obtained from this year’s implementation items isreflected as evaluation sub-items and theData acquired on solidified substances containing chemical admixtures was added as “Effect of adding chemical admixtures” to the evaluation sub-items of theapproaches are upgraded and updated.
section concerning “Characteristics of solidified substances” from among the 5 items mentioned below identified as the proposed evaluation axis for FY2021.
Evaluation items

Evaluation sub-items

Technical achievements

Development stage/trends; proven record of application to general waste, radioactive waste, and secondary waste generated from water treatment; and main
trouble case examples

Process performance

Number of processes, treatment temperature, treatment speed, maximum treatment unit, Cs volatilization rate, volatilization control measure, and required
specifications for solidification containers

Operability and safety

Process risk, maintenance details and frequency, remote operability, and ease of solidification process control

Economic efficiency

Main facility configuration, power consumption, consumables, facility lifetime, work area, volume reduction rate, secondary waste/amount, and the possibility oftreatment that mixes a different type of secondary waste generated from water treatment

Characteristics of solidified substances

Homogeneity, strength, leaching resistance, radiation resistance(hydrogen G value and property change of solidified substanceorganization), thermal resistance (the effects of a thermal load on the performance), long-term stability, effect of adding chemical admixtures©International Research Institute for Nuclear Decommissioning

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(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
i. Organization of evaluation axis and acquired data for comparing technologies➢ The evaluation method (approach) was improved by adding data acquired on solidified substances containing chemical
admixtures (Table 1).
Table 1 Details of the effect of adding chemical admixtures
Evaluation
items

Evaluation subitems

Cement in-drum/ Cement out-drum

AAM in-drum・ Verified chemical admixtures＞ Lignin sulfonic acid based＞ Melamine based＞ Naphthalene sulfonic acid based＞ Polycarboxylic acid based

・ Verified chemical admixtures＞ Lignin sulfonic acid based＞ Melamine based＞ Naphthalene sulfonic acid based＞ Polycarboxylic acid based
Characteristics
of solidified
substances✓✓[Addition]
Effect of adding
chemical
admixtures

・Lignin sulfonic acid based admixture was selected for carbonate slurry andpolycarboxylic acid based admixture was selected for iron co-precipitation
slurry.
・ The amount of carbonate slurry filled can be increased by adding chemical
admixtures.
＞ Scope of application：40%→45%・ In the case of iron co-precipitation slurry, since condensation gets delayedwhen filling rate increases, the scope of application cannot be expanded.
・ Irradiation characteristics of base material containing chemical admixtures＞ G value: 0.1 or less (irradiation of 3kGy Small cylinder of φ10x20mm)

・ M and MB40 indicated the same trend with respect to the verified chemical admixtures.
・ Solidified substances mixed with carbonate slurry showed a fluidity improvement
effect.
・ Fluidity of base material and all mixed systems reduced significantly in the case ofpolycarboxylic acid based admixture.
・ The scope of application of carbonate slurry can be expanded to some extent.
＞ M： By adjusting the Na concentration, Si concentration, the filling rate of carbonate
slurry can be improved from 30% up to 40%.
＞MB40: There are concerns about expanding the scope of application as there is asignificant drop in strength.
・ Iron co-precipitation slurry mixture system does not show fluidity improvement effect.

The effect of addition of chemical admixtures (4 types) to carbonate slurry and iron coprecipitation slurry from among the secondary waste generated from contaminated water
treatment, was added. The composition range that can contain secondary waste generated
from contaminated water treatment was extended for cement + carbonate slurry by addition ofchemical admixtures. (Fig. 1, Table 2)
The acquired data was organized as the evaluation axis and was reflected in the investigation
results for each solidification material.

Table 2 Scope of feasibility of solidification treatment when chemical admixtures are added (Cement + CS system)
Without chemical
admixtures

With chemical
admixtures

Carbonate slurry

15 to 40wt%

15 to 45wt%

Cement

30 to 55wt%

20 to 55wt%

Water

30 to 45wt%

30 to 45wt%

Fig. 1 Scope of application to cement +carbonate slurry©International Research Institute for Nuclear Decommissioning

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(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment
technology
i. Organization of evaluation axis and acquired data for comparing technologies- Summary -Results so far✓ Based on The investigation carried out in “①-i”, the evaluation sub-item “Effect of adding
chemical admixtures” was added to the section concerning “Characteristics of solidified
substances” in the proposed evaluation axis.
✓ The results of studies on secondary waste generated from contaminated water treatment(simulated carbonate slurry, simulated iron co-precipitation slurry) in the solidified substances
containing chemical admixtures were reflected, and the scope of application of the improved
low temperature solidification treatment technology items was presented.

Issues, etc.
✓ Establishing methods for acquiring data that is common for waste bodies with different
characteristics.

©International Research Institute for Nuclear Decommissioning

264

•

•

Achievement of the previous year

In order to more appropriately select the high temperature treatment
technology in which Cs volatilization becomes an issue, Cs volatilization andcontrol mechanism was investigated, the Cs volatilization characteristics wereconsolidated based on the treatment method and operating conditions, and in
addition, data related to the amount of volatilization was acquired andevaluated on a trial basis.
Based on the above-mentioned investigation, the following issues wereidentified in developing a detailed selection method.
✓ Expansion of data on an ongoing basis for evaluating the range of solid
waste to which the treatment technologies are applicable✓ Further acquisition of data under conditions considering actual operation➢ Goal• Cs volatilization is an issue in high temperature treatment
technologies and has a strong influence on the type or facility
configuration, etc. Hence, tests related to the Cs volatilization
control mechanism will be conducted at the laboratory level using
cold cap, top off frit, etc. regardless of the unique technology, toinvestigate the effect of Cs volatilization control.

Alumina tube
(outer diameter
3mm, inner diameter
2mm)

Air supply

Meteringpump

Drying column

Only during the
bubbling test

Air supply (carrier gas)
Drying column
Thermocouple(for internal use)

Dew point -73.3℃

Mass flow
meter
Thermocouple(for use on the surface)

Alumina pipe
(outer diameter
6mm, innerdiameter 4mm)
Off-gasSteel T pipe

Silica gelcolumn➢(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment
technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Details of implementation -Alumina Tammann
pipe (inner diameter
32mm, height
150mm)

Exhaust20mm (30mm at
the time of
bubbling)

0.5M nitric acid

0.5M nitric acid

Pump with flow control

Off-gas trap (3 or 6 stage)

CulletTube furnace

Sample volume : 80g
Internal temperature : 1195-1205℃

Fig. 1 Overview of the test apparatus

Details of implementation
Acquisition of data related to Cs volatilization control based on tests
Implementation of tests related to the Cs volatilization control mechanism at the laboratory level using cold cap, top off frit,etc. regardless of the unique technology, to investigate the effect of Cs volatilization control.

Indicators for the achievements of goal
Results of investigation of the volatilization control effect of various events are presented.

©International Research Institute for Nuclear Decommissioning265

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment
technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Test conditions ➢ Test Conditions (Table 1)• There are 2 loading formats as mentioned below.
Mixed sample: Loading format “Cs adsorbed zeolite + glass forming agent” for the sample being primarily assumed (Table 2).
Cullet: Mixed samples are melted to make glass cullets in order to reduce the volume of mixed samples due to testing related constraints (Nextpage Fig. 1)••

During test ⑤ (Top Off Frit (TOF)), 2 types of low melting point glass are selected and studied.
The maximum glass temperature is set at 1200℃ which is the maximum temperature of an electric furnace.
Table 2 Mixing ratio of zeolite andcomponents of glass forming agent

Table 1 Overview of the implemented test
No.

Test

①

Air supply flow ratedependency test②

Supply mode dependency
verification

Quantity of
Loading format*1 sample loaded[g]CulletMixed sample

100

30

Temperature dependency
verification

Cullet④

Cold cap

Cullet (Lower
layer)
Mixed sample(Upper layer)

80
20

⑤

Top off frit
(2 types of glass with alow melting point)

Cullet (Lower
layer)TOF (Upper layer)

40
40

③

100

⑥

Bubbling

Cullet100⑦

Cold cap+ Bubbling

Cullet (Lower
layer)
Mixed sample(Upper layer)

80
20

Heating
temperature[°C]

Surfacetemperature
[°C]

1200

1150

Bubbling flow rate
[mL / min]

About 2
About 2conditions
conditions
between 800 to between 800 to
1100
1100
About 2 to 3
About 2 to 3
conditions
conditions
between 800 to between 800 to
1200
1150
About 2conditions
between 1000 to
800-900
1200

1200

1200

1200

Component

Zeolite

53.88

(Cs)

(3.81)

Simulatedcarbonate slurrycomponent

CaCO3

10.05

MgO

6.95

Fe2O3

4.80

SiO2

4.21

B2O3

4.71

Na2CO3

2.23

LiOH・H2O

8.75

ZrO2

4.42

Total

100.00Glass forming
agent
component

1150

About 2 to 3
conditions
between 10 to 200

800

About 2 to 3
conditions
between 10 to 200

Weight added(%)

Zeolitecomponent

Simulated ironco-precipitation
slurry component

1150

Reagent
added*1: When the loading method involves lower layer + upper layer, if loading cannot be performed at 2 layers, the method of loading thesamples need to be re-examined.
©International Research Institute for Nuclear Decommissioning

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(a Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Test conditions ◼ Although the loading format for the samples being primarily assumed was supposed to be mixed samples, as the volume of the sample had to be reduced due tocircumstances pertaining to this test, the samples were supplied in the form of glass cullets.
➢ Preparation of sample to be loaded
Mixed sample made by adding glass forming agent to zeolite was heated at 1100℃ and glass cullets were prepared (Fig. 1). The condition at the time of heating is shown(Fig. 2).
Heated at
1100℃

Mixed sample

600°C

1000°C

1050℃

1100°C

1150°C

1200°C

After heating
The particles get fused but
there is almost no change in shape.

powder

The surface
becomes flat,and the sample
gets liquified

Along with increase in temperature,
the precipitation phase* reduces

Fig. 2 Melting behavior of the prepared glass cullets

Cullet (<2 [cm])Fig. 1 Preparation of the sample to be loaded

The shape of theparticles
disappears,
and the particles
fuse*Pollucite (nepheline) which does not easily dissolve at low temperatures and ZrO 2 were found in
the cullets, but it was found that these dissolve when heated at 1200℃.

➢ Preparation of the top off frit material
The top off frit materials (2 types) were heated under the following conditions respectively, to prepare top off frit. The condition at the time of heating is shown (Fig. 3).

•

Top off frit 1 (hereinafter, TOF-1)

600℃

700℃

800℃

950℃

1000℃

The mixed reagent is heated at
1200℃ and melted to test
manufacture the top off frit material
Melting point 950 to 1000℃ Particle fusion•

Top off frit 2 (hereinafter, TOF-2)

Disintegrationof the shapeof particles

Shape of the particles is lost,
they get fused,
and the surface becomes
unevenThe surface
becomes flat,and the sample
gets liquified

700℃

800℃

900℃

1000°C

Particle
separation

ParticlefusionDisintegrationof the shapeof particles

Unevensurface

1100°C

SiO2 is mixed with Top off frit 1, heated at
1300℃ and melted to prepare TOF-2
Melting point 1050 to 1100℃

Fig. 3 Melting behavior of the prepared TOF©International Research Institute for Nuclear Decommissioning

267

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Test results_1. ① Air supply flow rate dependency, ② Temperature dependency, ③ Supply mode dependency Test ① Air supply flow rate dependency test
• The air supply flow rate was changed and the volume of Cs volatilization was measured* (Fig. 1). Also, Cs was collected by means of a 6 stage offgas trap (Fig. 2).
*Cs volatilization volume: Amount of volatilized Cs collected from each site in the test system
6018100mL/min
500mL/min
1300mL/min

16

50

Cs volatilization (mg)

Cs volatilization (mg)

Cs volatilization

40
30
20
10

14
12
10
8
6
4
2

0

0

500

1000

0

1500

Tammann
Trap
22 ト
Trap
66
Trap
11 ト
Trap
55 ト
タンマン管 TT字管
ラップ
ラップ
ラップ
ラップ
ラップ
ラップ
pipe ト
Trap
33 ト
Trap
44 ト
pipe

Flow rate (mL/min)Fig. 1 Air supply flow rate and amount ofvolatilized Cs✓ With the increase in air supply flow rate, the amount ofvolatilized CS increased and the quantity that migrates
to the subsequent stages of the off-gas trap increased.
✓ When the air was supplied at 100 [mL/min], and 500[mL/min], the entire amount of volatilized Cs wascollected in the 6 stage off-gas trap.

Fig. 2 Amount of Cs collected from each site

✓ It was decided to reflect the conditions
when air was supplied at 500 [mL/min]in the test from then on.

Test ② + ③ Temperature dependency and supply mode dependency verification test
• The difference between the amount of volatilized Cs from the mixed sample obtained by adding glass forming agent to zeolite andthe cullets was acquired by changing the temperature (Table 1, Fig. 3).
Table 1 Relation between heating
temperature and surface temperature
Heating
temperature[℃]

Surface temperature [℃]
CulletMixed sample

1100

1053

1075

1000

955

977

850

819

811

Cs volatilization (mg)

12

カレット
CulletMixedsample
混合試料

10
8
6

Increase due todifference in surface
temperature

4
2
0

850˚C

1000˚C1100˚C

Fig. 3 Comparison of the volume of Cs volatilization incase of the cullets and the mixed sample✓ The amount of volatilized Cs was less in the case of
mixed sample as compared at cullets at heating
temperature 850℃ and 1000℃.
✓ At 1100℃ when both types of samples melted, theamount of volatilization increased in the case of themixed sample.
✓ Melting of samples can be considered to be an
impact of the difference in surface temperature(Table 1).
✓ In the cold cap test (Test ④) in which thesurface temperature is low, it was decided
that cullets would be used in the lower layer,
the mixed sample would be placed in theupper layer part where the temperature will be
1000℃ or lower, and the test would be
conducted.
©International Research Institute for Nuclear Decommissioning

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(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Test results_2.④ Cold cap ⑤ Top off Frit Test ④ Volatilization control effect of cold cap• The surface temperature of molten material was controlled to form a cold cap as illustrated in Fig. 1, and the volume of Cs volatilization under test
conditions (Table 1) was measured (Fig. 2).
Table 1 Test conditions(Heating temperature and measurement
temperature)
Test

Heating
temperature[℃]

Ref. 1200

1200

1205

1147

Ref. 850

850

-

819

1

1200

1185

803

2

1100

1094

806

3

1000

990

806

Mixed
sampleCulletFig. 1 Overview
of the sample tobe loaded✓ By letting a cold cap form on thesurface of the sample, the volume
of Cs volatilization was reduced to
1/50 or lower.
✓ It was confirmed that the volume
of Cs volatilization depends on thetemperature of the cold cap(sample surface temperature) andnot on the temperature of thesample inside.

BottomSurfacetemperature temperature[℃][℃]

Fig. 2 Relation between the temperature of the cold cap andthe volume of Cs volatilization

Test ⑤ Investigation of the volatilization behavior of Cs when Top off frit is used
• Top off frit material was loaded as shown in Fig. 3, and the volume of Cs volatilization respectively when 2 types of Top off frit material wereused under temperature conditions mentioned in Table 2 was measured (Fig. 4).

Table 2 Test conditions (Heating
temperature and measurement temperature)
Test
Top off
frit

Cullet4
5
6
7

Heating
BottomSurfaceTop oftemperature temperature temperature
frit
[℃]
[℃]
[℃]
TOF-1

TOF-2

1200

1198

1165

1200

1156

1028

1200

1201

1172

1200

1166

1035

Fig. 3 Overview of thesample to be loaded

Fig. 4 Relation between the presence of top off fritand the volume of Cs volatilization✓ When a top off frit with a low melting
point is used, the volume of Cs
volatilization is likely to actually
increase. Hence it is important to use
frits that have a melting point closer tothe heating temperature.
✓ By making sure that the surface
temperature is lower than the melting
point of the frits, high volatilization
control effect was seen, but the amountof frits loaded increased (increase involume of waste). Hence a balance
between the amount loaded and thevolume of volatilization is important.

©International Research Institute for Nuclear Decommissioning

269

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Test results_3.⑥ Bubbling, ⑦ Cold cap + bubbling Test ⑥ Investigation of the Cs volatilization behavior as a result of bubbling• Bubbling was performed and the volume of Cs volatilization was measured (Fig. 1, 2)Metering
pump

Bubbling gas line
Drying column

Gas feed line
Mass flow
meter

Drying column

Off-gas・・・

Cold cap (onlyTest ⑦)

Bubbling

Fig. 1 Overview diagram of the test (Same for test ⑥ and ⑦)✓ The moisture content from the air in the laboratory
was removed and the resulting gas was used as
bubbling gas.
✓ The volume of volatilization increased
approximately linearly with the increase inbubbling flow rate in Region A where individuallyindependent air bubbles were formed, but inRegion B where the air bubbles were continuous,
the volume of Cs volatilization showed a tendency
to remain constant regardless of the flow rate.
✓ The volume of Cs volatilization when bubbling
was performed at 10 [mL/min] (this is increased
4.9 times for in-furnace temperature) for melted
samples with a volume of approx. 32 [mL] was
2.1 times, and it was 4.0 times when the flow rate
Fig. 2 Dependency of the volume of Cs volatilization on
the bubbling flow rate
was 50 [mL/min].

Test ⑦ Cold cap + bubbling• Volume of Cs volatilization when bubbling is performed upon controlling volatilization by means of co ld cap, was measured (Table 1, Fig. 1, 3).
✓ By letting a cold cap form, the volume ofvolatilization at the time of bubbling was reduced
to 1/10 to 1/30.
✓ The volume of Cs volatilization tended to increase
along with an increase in bubbling flow rate, but it
was not significant.
✓ During Test 1 and Test 3, bubbles did not form at
the surface (formation of hot spot) other thanseveral occasions when they were observed in
the initial stages.

Table 1 Test conditions (Heating
temperature and measurement temperature)1

Heating
temperature
[℃]
1200

Bottomtemperature
[℃]
1130

Surfacetemperature
[℃]
914

2

1200

1170

803

10

3

1200

1140

891

20

4

1200

1139

919

50

Test

Bubbling[mL / min]10

Fig. 3 Results of measuring the volume
of Cs volatilization

Further, during the test white plume came up
from the surface. It is believed that the bubblinggas escaped from between the sample particlesof the cold cap.
©International Research Institute for Nuclear Decommissioning270

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Volume of Cs volatilization for each volatilization control effect and its evaluation ➢ Changes in the volume of Cs volatilization for each volatilization control effect• The volume of Cs volatilization was organized by measurement site in the test
equipment (Table 1).

Table 1 Volume of Cs volatilization by measurement site
Test ⑦

Test ④

Test ⑤

Test ⑥

Cold cap

Top off frit

Bubbling

1205

1185

1166

1198

Cold cap +bubbling
1139

Surface temperature [℃]

1147

803

1035

1184

919

Glass sample [%]99.09299.98399.95696.35399.798

Tammann pipe

0.338

0.006

0.015

1.167

0.029

T pipe

0.414

0.006

0.022

1.395

0.076

Off-gas trap 1

0.069

0.003

0.002

0.989

0.054

Off-gas trap 2

0.045

0.001

0.002

0.073

0.022

Off-gas trap 3

0.021

0.000

0.001

0.016

0.012

Off-gas trap 4

0.012

0.000

0.001

0.003

0.006

Off-gas trap 5

0.009

0.000

0.000

0.002

0.002

Off-gas trap 6

0.002

0.000

0.000

0.001

0.001

Total volatilization [%]

0.908

0.017

0.044

3.647

0.202

Total [%]

100.000

100.000

100.000

100.000

100.000

Location

Reference

Glass temperature [℃]Volatilized
substancecollection
site
[%]

✓

It was possible to reduce the volume of Cs volatilization to 1/50 or lower by letting a cold cap form with surface temperature approx. 800℃. It was
indicated that formation of a cold cap was effective in controlling volatilization.

✓

The volume of Cs volatilization increased to maximum 5 times as a result of bubbling, but by letting a cold cap form, the volume of volatilization
at the time of bubbling reduced to 1/10 - 1/30.

✓

It was indicated that it was possible to reduce the volume of Cs volatilization to 1/20 or lower by appropriately selecting and using the top off frit.

✓Cs was collected the most at the T pipe site. This suggested that Cs is likely to accumulate at low temperature sites inside the equipment.

©International Research Institute for Nuclear Decommissioning

271

(a) Establishment of selecting advance treatment methods [2] investigate of approach to assessing applicability for treatment
technology
ii. Investigation related to volume of Cs volatilization during high temperature treatment and its control- Summary -Results so far✓ Data related to the volume of volatilization for each volatilization control factor was acquired,the volatilization control effect was compared, and results were presented.
✓ It was possible to reduce the volume of Cs volatilization to 1/50 or lower by letting a cold cap
form with surface temperature approx. 800℃. It was indicated that formation of a cold cap
was effective in controlling volatilization.
✓ It was indicated that it was possible to reduce the volume of Cs volatilization to 1/20 or lower
by appropriately selecting and using the top off frit.
✓ The volume of Cs volatilization increased to maximum 5 times as a result of bubbling, but byletting a cold cap form, the volume of volatilization at the time of bubbling reduced to 1/10 1/30.

Issues, etc.
✓ The surface coverage of the cold cap is important for controlling the volume of Cs
volatilization. Hence the method of capturing visible images needs to be studied.

✓ The impact of bubbling is expected to change depending on the test system. Hence further
studies need to be carried out considering the difference with the actual equipment.
✓ The investigation on migration behavior of elements besides Cs at the time of melting needs
to be studied as well.
©International Research Institute for Nuclear Decommissioning

272

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods◼ Achievements up to FY2020
 1F waste was classified into 27 types, and waste information
was organized by type.
 Based on investigation of cases from Japan and overseas,
multiple dose evaluation conditions (scenario / model /parameters) were tentatively established while taking
uncertainties into consideration, and preliminary dose
evaluation was conducted for 27 types of 1F waste.
 The method for deriving the disposal classification of typical 1F
waste depending on the treatment options was pointed out
based on the results of the preliminary dose evaluation.

① Preliminary dose evaluation of eachtype of waste② Organization of waste information requiredfor investigating disposal concepts

Targeting 27
types of waste(Comprehensive
study)③ Evaluation of the characteristics ofeach type of waste<Scope of implementation in FY2019>• Preliminary dose evaluation of waste• Organization of waste information and evaluation ofcharacteristics• Disposal concepts and dose evaluation of the selected
waste<Scope of implementation in FY2020>• Review of the evaluation conditions based on the studies
conducted in FY2019• Study of disposal concepts and dose evaluation of theselected waste⑨ Selection of waste to be studied in
FY2020④ Selection of waste to be studied in
FY2019Narrowing down

► Information and characteristics of waste (I, III)► Disposal cases from Japan and overseas(Studied in 2017-2018)► Information and characteristics of waste (II, III)
► Radioactivity reviewed in FY2019► Disposal related good practices from Japan andoverseas (Studied in 2017-2018)► Evaluation conditions (Reviewed by means ofelicitation)
Disposal concept study process (Studied in 2017-2018)

Disposal concept study process (Studied in 2017-2018)◼ Goal To point out multiple disposal concepts for typical waste
streams and items required for their safety evaluation andstudy, based on the results of the latest characterization andtreatment studies.
◼ Details of implementation

Study of multiple disposal concepts for typical waste streams
based on the content of investigations related to disposal
concepts and safety evaluation methods in Japan andoverseas, as also information related to the properties of solid
waste. And, identification of items that need to be studied forestablishing the disposal concepts and safety evaluation
methods for each disposal concept based on treatment
technologies applicable to solid waste, and collection andorganization of the required information.

◼ Indicators for determining goal achievement
 Presentation of typical waste streams along with the reason for
their selection.
 Presentation of multiple disposal concepts for typical waste
streams
 Presentation of items that need to be studied for establishingthe disposal concepts and safety evaluation methods based on
the treatment technologies for typical waste streams
 Collection and organization of information on the items thatneed to be studied as mentioned above.

⑩ Study of the process until waste conditioning⑤ Study of the process until waste conditioning(Treatment, segregation, measurement, etc.)*(Treatment, segregation, measurement, etc.)*

Targeting theselected waste(Detailed study)⑥ Study of disposal
methods*⑦ Dose evaluation

Provision ofevaluation results
as feedback* The process until waste conditioning and thedisposal method are collectively called as“Disposal concept” in this study.
⑧ Consolidation of FY2019 accomplishments(Uncertainties and extent of influence of theparameters were evaluated)⑪ Study of disposal
methods*⑫ Dose evaluation

Provision ofevaluation results
as feedback

Output Image⚫
Proposed disposal concepts depending on the type of waste and theevaluation conditions⚫
Items such as dose evaluation method, issues pertaining toparameters, future requirements, etc.
⚫
Requirements pertaining to waste bodies depending on the form ofwaste disposal and the evaluation conditions

Fig. 1 Flow of study of the disposal concepts and safety evaluation methods
implemented in FY2019-2020

Leaching rate
Flash emission
Flash emission engineered barrier
present

Engineered barrier present

Engineered barrier present

Engineered barrier present

Fig. 2 Results of the FY2020 dose evaluation (Trench, groundwater)©International Research Institute for Nuclear Decommissioning

273

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

FY2021 Study Plan Overview
Accomplishments of government
R&D projects in FY2020
Characterization Inventory (FY2020 edition)
Properties ofAdvance treatment
solidified substances
methods
Materials with impact
on disposal

Sorption property

・・
・

・・
・

NUMO
JNFL
・・
・

Study of sensitivity
evaluation conditions

Consolidation of input
information

Amount and inventory of waste
Properties of waste bodies
Dose evaluation parameters

Domestic trend

Sensitivity evaluation

Consolidation of input
information

Sensitivity evaluation
conditions

Sensitivity evaluation

Comprehensive
Technical Report
Application for
permission for change

Sensitivity evaluation of disposal
depth, barrier composition, barrier material,
barrier thickness, leaching rate, etc.

・・・Study ofpreliminary
proposals for the
disposalconcepts
Feedback to The investigation ofcharacterization, treatment, etc.
Characterization: Important nuclides,
etc. for disposal
Treatment: Effect, etc. on disposal
due to the treatment method.

Sensitivity evaluation model

Study of typical waste
streams

Typical waste streams

Study of proposed
disposal concepts

Disposal concepts (Preliminary proposal)

Dose evaluation
Dose evaluation

Proposal of disposal concepts based on the
results of dose evaluation©International Research Institute for Nuclear Decommissioning274

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

FY2021 Study Plan Implementation items
Consolidation of input information••

Information on 1F waste studied until FY2020 (characterization, treatment method, materials with impact on
disposal, etc.) will be collected and consolidated, and parameters such as waste amount, inventory, waste bodyproperties, etc., used for evaluation will be established.
Dose evaluation parameters such as the distribution coefficient, etc. for each nuclide that might impact nuclide
migration will be established based on the latest domestic trend of the NUMO Comprehensive Technical Report,
etc.

Implementation of sensitivity evaluation•••

Models of multiple disposal concepts will be created with depths corresponding to trench disposal, pit disposal,
and medium depth disposal based on the results of studies on disposal concepts conducted up to FY2020.
Based on the results of studies on consolidation of input information, parameters for the sensitivity evaluationbase case and the parameters for the sensitivity evaluation will be established.
The models of multiple disposal concepts will be analyzed and evaluated by changing the parameters (barriermaterial, barrier thickness, distribution coefficient, etc.) of the facilities, and the sensitivity of parameters will be
evaluated.

Study of preliminary proposals for the disposal concepts•

Based on the results of sensitivity evaluation and the results of consolidating input information, preliminary
proposals for the disposal concepts suitable for the properties of 1F waste will be presented.

Dose evaluation•

Dose evaluation will be carried out with respect to the preliminary proposal of the disposal concepts, andrevisions will be made based on the results.
©International Research Institute for Nuclear Decommissioning275

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safetyassessment methods

Consolidation of input information (1/10)•

The amount and inventory used for investigating disposal concepts was updated based on the
results of the analytical inventory studied in FY2020 as part of “Characterization”.

•

Treatment methods were established for each waste based on the results of studies on treatments
conducted until FY2020 and the weight and amount when waste conditioning is carried out wascalculated.

•

The diffusion coefficient inside the solidified substance and the leaching rate of the solidified
substance were calculated for the leaching rate from glass, melt, cement, AAM solidified
substances, based on the outcome of studies on advance treatments up to FY2020.

•

Some of the set values of distribution coefficients of FY2020 based on the application for changing
the JNFL Low-level radioactive waste burial center, were reviewed while referring to the latest
domestic disposal trend (NUMO Comprehensive Technical Report, etc.).

©International Research Institute for Nuclear Decommissioning

276

(b) Proposal of disposal methods and development of safety evaluation methods① Consolidation of information to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (2/10)<Amount of waste (weight and volume)>
The amount of waste in which the outcome of characterization carried out in FY2020 was reflected, was used in the investigate
of sensitivity analysis and disposal concept proposals.
・The sum total of the net volume of metal and concrete respectively in the rubbles in FY2021 are listed.
⇒ The amount that takes waste conditioning into consideration is assumed to have an impact on the exposure dose mainlyin the nearing scenario.
FY2021/2020

Comparison of amount of waste

Amount of waste in FY2020

Amount of waste in FY2021

m3

ton

ton/ m3

m3

ton

ton/ m3

KURION

2847

11565

4.06

2877

11685

4.06

AREVA

597

597

1.00

597

597

1.00

SARRY

1196

5184

4.33

1274

5520

4.33

Slurry 1_iron co-precipitation

1833

1833

1.00

2001

2001

1.00

Slurry 2_Carbonate precipitation

7887

7887

1.00

8610

8610

1.00

ALPS-4_Ag impregnated activated carbon

222

184

0.83

181

181

1.00

ALPS-2_titanate

332

276

0.83

272

272

1.00

ALPS-3_ferrocyanide

222

184

0.83

181

181

1.00

ALPS-6_chelating resin

443

368

0.83

363

363

1.00

ALPS-5_titanium oxide

222

184

0.83

181

181

1.00

ALPS-7.1

46

57

1.25

68

68

1.00

Rubble 1

190000380000

2.00

34656

186873

5.39

Rubble 2

48000

96000

2.00

30140

177951

5.90

Rubble 3

26700

53400

2.00

6079

33705

5.54

Rubble 4

52000

104000

2.00

20729

103528

4.99

Rubble 5

36000

72000

2.00

19617

106797

5.44

Felled trees

18000

14400

0.80

17264

13811

0.80

Pressure vessel waste (equivalent to L1 before the accident)

230

216

0.94

231

222

0.96

Pressure vessel waste (equivalent to L2 before the accident)

185

1466

7.91

185

1467

7.91

Waste in the containment vessel (metal) (equivalent to L2 before the accident)

1164

873

0.75

1164

877

0.75

Waste in the containment vessel (metal) (equivalent to L3 before the accident)

1211

1831

1.51

1211

1829

1.51

Waste in the containment vessel (concrete) (equivalent to L2 before the accident)

431

328

0.76

431

329

0.76

Waste in the containment vessel (concrete) (equivalent to L3 before the accident)

449

682

1.52

449

679

1.51

Waste inside the building (metal) (equivalent to L3 before the accident)

23677

35677

1.51

23677

35677

1.51

Waste inside the building (metal) (equivalent to nonradioactive waste before theaccident)

6454

9725

1.51

6454

9725

1.51

Waste inside the building (concrete) (equivalent to L3 before the accident)

8766

13208

1.51

8766

13208

1.51

Waste inside the building (concrete) (equivalent to nonradioactive waste before theaccident)

3705865584031.51370586558403

1.51©International Research Institute for Nuclear Decommissioning

277

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (3/10)<Radioactivity inventory>
The inventory that serves as input for the investigate of sensitivity analysis and disposal concept proposals was established, based on theoutcome of characterization carried out in FY2020⚫
The radioactivity of 68 nuclides that have a major impact on disposal were established for 27 types of wastes (5 representative nuclides
are listed below). Considering the uncertainty of the inventory, evaluation was carried out in two ways, one using the single mostfrequent value (mode) (using some of the average values) and another using the 95%itle.
⚫
The amount of radioctivity from C-14 and I-129 from some of the secondary waste generated from contaminated water treatment(SARRY, KURION, AREVA, etc.) was changed substantially ⇒ It is assumed to mainly have an impact on exposure dose in the
groundwater scenario.
Radioactivity in 2021 and 2020 is less than 0.5
放射能の2021年/2020年が0.5以下
Radioactivity in 2021 and 2020 is more than 2
放射能の2021年/2020年が2以上

Legend:C-14
Amount of radioactivity
Comparison table of FY2020 and FY2021

Geomean_Mode2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

Sr-90

2021 Year(Bq)

95%ileComparison with theprevious year
(FY2021 / FY2020)

Geomean_Mode2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

I-129

95%ile2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

Cs-137

Pu-239

Geomean_Mode2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

95%ile2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

Geomean_Mode2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

95%ile2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

Geomean_Mode2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

95%ile2021 Year Comparison with the(Bq)previous year
(FY2021 / FY2020)

1
2
3
4
5

KURION
AREVA
SARRY
Slurry 1_Iron co-precipitation
Slurry 2_Carbonate precipitation

9.6E+11
6.0E+08
7.5E+112.7E+11
1.1E+11164.833
16.929180.367
32.652
1.685

1.4E+12+09.9E+2
1.1E+122.3E+12
1.4E+1290.20230.09695.3000.976
0.540

9.1E+14
1.5E+16
6.8E+16
3.9E+16
8.9E+150.056
51.4405.170
33.3741.881

2.0E+17
3.8E+16
1.6E+172.0E+17
1.1E+170.687
21.187
0.711
0.999
0.512

5.8E+10
3.7E+07+10.6E+4
4.4E+08
4.3E+08

162.530
16.696178.208
0.850
0.751

1.0E+11+08.7E+1
7.7E+10
1.3E+09
1.3E+09106.17630.162115.615
0.788
0.767

+17.7E+1
1.0E+15
1.3E+172.0E+12+11.9E+1

1.001
1.001
1.053
0.162
0.017

+17.7E+1
1.0E+15
1.3E+17+13.7E+1
4.7E+12

1.000
1.000
1.053
0.001
0.000

+10.9E+1
5.7E+08
1.4E+10
1.1E+09
5.4E+09117.334
6.167
122.963
0.495
0.653

8.0E+112.0E+11
4.9E+11
6.0E+11
7.9E+11245.389
20.349247.027
0.586
0.610

6

ALPS-4_Ag impregnated activated carbon

1.1E+06

0.370

9.3E+06

0.411

8.2E+102.021

5.6E+11

0.605

4.2E+102.505

1.5E+11

0.847

7.8E+08

0.005

7.3E+090.001

4.0E+04

0.476

2.3E+06

0.553

7
8
9
10
11
12
13
14
15
16
17

ALPS-2_titanate
ALPS-3_ferrocyanide
ALPS-6_chelating resin
ALPS-5_titanium oxide
ALPS-7, 1
Rubble 1
Rubble 2
Rubble 3
Rubble 4
Rubble 5Felled trees

1.1E+06
1.1E+06
1.0E+06
1.1E+06
1.1E+06
3.6E+08
1.1E+11
1.2E+11
1.8E+12
3.6E+122.1E+11

0.388
0.381
0.392
0.408
0.391
0.007
0.697
0.188
0.610
1.242
0.753

8.8E+06
9.0E+06
7.0E+06
8.7E+06
8.7E+06+09.9E+2
9.1E+119.8E+11
1.4E+13+13.9E+2
9.9E+11

0.403
0.390
0.313
0.386
0.383
0.012
1.200
0.334
4.756
9.680
0.714

8.8E+128.5E+077.8E+077.8E+09
2.4E+07
1.4E+10+12.5E+4
4.9E+127.1E+13
1.4E+145.8E+121.876
520.4481.993
84993.994
1424.382
0.015
1.436
0.387
0.405
0.466
1.102

1.1E+14
6.2E+08
4.8E+08
5.2E+102.2E+08
1.3E+11
4.1E+13
4.4E+13
6.5E+14
1.3E+15
4.7E+13

0.512
0.722
0.629
57.300
0.240
0.011
1.042
0.281
0.294
0.338
0.621

4.1E+05
4.1E+05
4.0E+05
4.2E+05
4.3E+05
6.0E+06+09.9E+1
2.1E+09
3.0E+10
6.1E+10
1.2E+10

0.762
0.761
0.784
0.816
0.807
0.002
0.210
0.057
0.190
0.387
0.714

1.3E+06
1.2E+06
1.2E+06
1.3E+06
1.3E+06+08.7E+1
5.4E+105.8E+108.5E+11+12.7E+1
6.2E+10

0.777
0.738
0.724
0.774
0.789
0.011
1.045
0.330
4.096
8.336
0.656

8.4E+08
2.2E+107.8E+05+07.7E+1
2.3E+05
1.1E+13
3.4E+15
3.7E+15
5.4E+16
1.1E+17
3.9E+15

0.006
0.002
0.006
124.363
1.772
0.017
1.678
0.452
0.474
0.544
1.041

8.2E+09+12.7E+1
7.5E+06
6.2E+08
3.6E+06
1.1E+13
3.5E+15
3.7E+15
5.4E+16
1.1E+17
3.9E+15

0.001
0.000
0.001
0.049
0.000
0.017
1.680
0.453
0.474
0.545
1.042

4.0E+04
40.E+04
2.2E+05
1.2E+04
1.2E+04
1.1E+06
3.4E+08
3.7E+08
5.4E+09
1.1E+10
3.3E+080.512
335.5160.295
1185.287
9815.208
0.008
0.817
0.220
0.231
0.265
0.439

2.4E+06
2.4E+06
2.3E+088.1E+05
9.2E+05+07.9E+1
6.0E+09
6.5E+09
9.4E+10+11.9E+1
6.3E+09

0.574
0.611
0.568
0.491
0.583
0.004
0.350
0.094
0.099
0.114
0.219

18

Pressure vessel waste (equivalent to L1
before the accident)+13.6E+4

1.000+13.6E+4

1.004

4.4E+16

1.790

4.8E+161.013

3.1E+070.258

3.2E+101.413

8.2E+16

1.246

8.7E+16

0.984

2.2E+141.502

2.3E+140.960

19Pressure vessel waste (equivalent to L2
before the accident)

5.4E+100.980

4.7E+11

1.386

4.0E+16

1.665

4.8E+16

1.024

1.4E+070.161

2.6E+10

1.417

6.6E+16

1.207

7.0E+16

0.984

2.1E+141.484

2.3E+14

0.965

20

Waste in the containment vessel (metal)(equivalent to L2 before the accident)

3.3E+10

0.824

4.9E+11

1.404

9.7E+15

1.441

1.2E+161.004

8.6E+070.170+10.9E+2

1.419

7.3E+16

1.351

7.8E+16

0.976

5.4E+13

1.455

5.8E+13

0.953

21

Waste in the containment vessel (metal)(equivalent to L3 before the accident)

6.9E+100.905

5.5E+111.377

1.1E+16

1.647

1.3E+16

1.015

8.8E+070.168

3.0E+10

1.419

7.7E+161.353

8.1E+16

0.976

5.5E+13

1.479

5.8E+13

0.968

22

Waste in the containment vessel(concrete) (equivalent to L2 before theaccident)

9.6E+08

0.275+11.7E+1

1.451

9.0E+15

1.424

1.2E+16

1.008

3.4E+070.177

1.1E+10

1.412

2.7E+161.278+16.9E+2

0.977

5.1E+13

1.378

5.8E+13

0.968

23

Waste in the containment vessel(concrete) (equivalent to L3 before theaccident)

1.4E+09

0.339

1.8E+11

1.449

1.0E+16

1.630

1.2E+16

1.003

3.5E+070.176

1.1E+101.415+16.9E+2

1.285

3.0E+16

0.976

4.9E+13

1.367

5.7E+13

0.971

24

Waste inside the building (metal)(equivalent to L3 before the accident)

1.3E+120.991

1.3E+120.976

8.5E+120.040

1.1E+150.182

3.0E+08

0.417

5.0E+08

0.199

9.3E+140.299

1.4E+15

0.257

2.8E+100.862

5.0E+11

0.344

25

Waste inside the building (metal)(equivalent to non-radioactive waste
before the accident)

1.2E+090.355

2.0E+090.182

8.1E+120.142

2.6E+140.168

6.3E+070.316

1.2E+08

0.184

2.4E+140.281

3.5E+140.240

5.4E+08

0.431

1.3E+110.300

26

Waste inside the building (concrete)(equivalent to L3 before the accident)

1.6E+100.846+10.7E+1

0.577

2.4E+14

3.194

3.1E+140.143

8.6E+070.314

1.4E+08

0.156

3.3E+140.282

4.3E+14

0.214

6.5E+08

0.380

1.6E+11

0.295

27

Waste inside the building (concrete)(equivalent to non-radioactive waste
before the accident)

6.7E+100.361

1.1E+11

0.174

6.6E+140.211

1.6E+16

0.172

4.7E+09

0.407

6.8E+09

0.175

1.3E+160.280+16.9E+1

0.229

2.7E+100.378

7.2E+120.346©International Research Institute for Nuclear Decommissioning

278

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (4/10)<Set-up for the treatment>
During the analysis and evaluation carried out in FY2021, treatment methods were established for eachwaste based on the results of studies on treatments conducted until FY2020 and the weight and amountwhen waste conditioning is carried out was calculated.

Up to FY2020: Waste was evaluated**: Treatment methods were established andevaluated in some of the studies on typical waste.

FY2021: Weight and amount were set up considering treatment
methods for each waste based on the results of studies ontreatments conducted until FY2020

Raw waste

Dehydration, etc.
Solidification, etc.

Weight and volume is the same as raw waste

Set by multiplying an appropriate coefficientto the weight and volume of raw waste©International Research Institute for Nuclear Decommissioning

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(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (5/10)<Values established for the volume and weight of waste>
Type of waste

Waste conditioning
volume ratio

Waste conditioning
weight ratio

Original volume
Original weight (t)(m 3 )

Waste body volume(m 3 )

Waste body weight (t)KURION

1

2

2,877

11,685

2,877

5,753

AREVA

0.03

0.04

597

597

18

24

SARRY

1

2

1,274

5,520

1,274

2,548

Slurry 1_Iron co-precipitation

0.3

0.5

2,001

2,001

600

1,000

Slurry 2_Carbonate precipitation

0.3

0.4

8,610

8,610

2,583

3,444

ALPS-4_Ag impregnated activated carbon

0.5

0.5

181

181

91

91

ALPS-2_titanate

0.5

0.5

272

272

136

136

ALPS-3_ferrocyanide

0.5

0.5

181

181

91

91

ALPS-6_chelating resin

0.5

0.5

363

363

181

181

ALPS-5_titanium oxide

0.5

0.5

181

181

91

91

ALPS-7, 1

0.5

0.5

68

68

34

34

Rubble 13.7

234,656

186,873

129,089373,747

Rubble 24.0

230,140177,951

120,113355,902

Rubble 3

3.8

2

6,079

33,705

23,11467,410

Rubble 4

3.5

220,729103,52873,019207,055

Rubble 53.8

219,617

106,797

73,588

213,594

Felled trees

1.3

3

17,264

13,811

22,44341,434

RPV waste (L1 before the accident)

1

2

231

222

231

443

RPV waste (L2 before the accident)

5

2

185

1,467

927

2,934

PCV waste (metal) (L2 before the accident)

1

2

1,164

877

1,164

1,753

PCV waste (metal) (L3 before the accident)

1

2

1,211

1,829

1,211

3,658

PCV waste (concrete) (L2 before the accident)

1

2

431

329

431

658

PCV waste (concrete) (L3 before the accident)

1

1

449

679

449

679

Waste inside the building (metal) (L3 before the accident)

1

2

23,67735,677

23,67771,353

Waste inside the building (metal) (non-radioactive waste
before the accident)

1

2

6,454

9,725

6,454

19,449Waste inside the building (concrete) (L3 before theaccident)

1

1

8,766

13,208

8,766

13,208Waste inside the building (concrete) (non-radioactive
waste before the accident)

1

1

370,586

558,403

370,586

558,403558,245

1,274,740

863,237

1,945,073

Total©International Research Institute for Nuclear Decommissioning280(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (6/10)

References<The approach while establishing the values for the volume and weight of waste>①KURION / SARRY⇒ The current weight has been calculated “including the radiation shielding container”.
(KURION: 2877m3, 11685t, SARRY: 1274m3, 5520t)⇒ It is assumed that volume does not change and the density becomes 2t/m 3, as an
assumption for solidified substances.
② AREVA / Iron co-precipitation slurry / Carbonate slurry✓

Assumed that Dehydration, etc. treatment → solidification would be conducted✓

Value obtained by calculating the volume ratio and weight ratio assuming thatthe filling rate for solidified substances is 30% for waste (solid) (20% for ironco-precipitation slurry) and 70% for filler material (80% for iron co-precipitation
slurry), based on the weight ratio and volume ratio of solid/liquid raw waste,
and rounding it up.
Waste conditioning
coefficient

AREVA

Volume ratio of raw waste

0.03

Iron coprecipitation
slurry0.3

Weight ratio of raw waste

0.04

0.5

Carbonate slurry

External appearance and structureof KURION adsorption tower

External appearance and structureof SARRY adsorption tower

Carbonate slurry

Iron co-precipitation slurry
Weight ratio of items stored in pits
Solid

LiquidVolume ratio of items stored in pits

Solid

LiquidCalculated based on materials from the 52nd
Secretariat meeting and the 73rd Monitoring and
Evaluation Committee meeting

Weight ratio
Solid

Weight ratio

Volume ratio
Solid

LiquidSolid

LiquidVolume ratio
LiquidExcerpts from materials from the 21st
Secretariat meeting

LiquidSolid

Excerpts from materials from the 34th
Secretariat meeting

0.3
0.4

③ ALPS 1 to 7
⇒ As analytical data on raw waste is not available, based on results of iron co-precipitation slurry and carbonate slurry, the weight ratio andvolume ratio of raw waste during waste conditioning was assumed to be 0.5.
④ Rubble 1 to 5
⇒ The filling rate of metal rubble was assumed to be 20% (that is, the waste conditioning volume ratio: 5) and the filling rate of concrete rubble wasassumed to be 50% (that is, the waste conditioning volume ratio: 2). Also, the waste conditioning weight ratio was considered to be 2 byconservatively (on the lower side) rounding up the weight ratio that includes the container and filler material.
The waste conditioning volume ratio of Rubble 1 to Rubble 5 in which concrete and metal are combined, changed in the range of 3.5 to 4 basedon an overall average, and changed in the range of 2 to 5 depending on the proportion of concrete and metal in individual waste bodies.

⑤ Felled trees⇒ ⇒ The amount of raw waste is the amount after incineration. Hence the filling rate of incineration ash was considered to be 30wt%
and the density of solidified substance was considered to be 2t/m3, the waste conditioning weight ratio was set at 3 and the wasteconditioning volume ratio was set at 1.3.
©International Research Institute for Nuclear Decommissioning

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(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

References

Consolidation of input information (7/10)<The approach while establishing the values for the volume and weight of waste>⑥ RPV / PCV waste and waste inside the building⇒ Density was calculated based on the volume and weight of raw waste and was organized from the following
perspectives.
(i) Metal waste✓ Density about 7.9t/m3 (Only RPV waste (L2 before the accident))⇒ Assuming a filling rate of 0.2 similar to that of metal rubble, the waste conditioning volume ratio was set at 5,and the waste conditioning weight ratio was set at 2.
✓ Density about 0.7 to 1.5t/m3 (Other than RPV waste (L2 before the accident))⇒ Assumed that it is stored in containers. The waste conditioning volume ratio was assumed to be 1, and thewaste conditioning weight ratio was assumed to be 2.

(ii) Concrete waste
The waste conditioning volume ratio was assumed to be 1, and when the density was 1 or less, the wasteconditioning weight ratio was considered to be 2, whereas when the density was 1 or more, the wasteconditioning weight ratio was considered to be 1.

©International Research Institute for Nuclear Decommissioning

282

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (8/10)<Leaching rate and diffusion coefficient inside the waste body>
The diffusion coefficient inside the solidified substance and the leaching rate of the solidified substance were
calculated for the leaching rate from glass, melt, cement, and AAM solidified substances, based on theoutcome of studies on advance treatments up to FY2020.
Table 1 (Example) Leaching rate of each nuclide in cement (OPC)
Diffusion dominated (When it is assumed that leaching is mainly due to diffusion)

Dissolution dominated**

Nuclides*Diffusion coefficient D (cm2/s) Leaching period (Years)
1
Carbonate slurry Cs

6.3E-08

Carbonate slurry Sr-10.7E+1

Carbonate slurry Sn, Ce

6.9E-14Iron co-precipitation
slurry Cs

4.4E-08

Iron co-precipitation
slurry Sr

1.6E-09

Iron co-precipitation
slurry Sn, Ce

6.9E-14*Corresponds to the solidification
mentioned in ② on p281

Leaching rate (1/y)

1

1.6E-01
Entire amount gets leached out
in 41 years
8.2E-03100

8.2E-04100

1

1.6E-04100

1.6E-0511

1.3E-01
Entire amount gets leached out
in 58 years
2.5E-02100

2.5E-031

1.6E-04100

1.6E-05

100

Leaching rate (1/y)

Leaching period for theentire amount (y)3.1E-01

3.2E+00

1.6E-02

6.1E+01

3.3E-04

3.0E+03

2.6E-01

3.8E+00

4.9E-02

2.0E+01

3.3E-04

3.0E+03** Since the results of a short (91 days) immersion test are extrapolated linearly, it is dissolution dominated.
The leaching rate (when it is assumed that leaching is mainly due to diffusion) becomes conservative.

Table 2 (Example) Leaching rate of glass solidified substance
Evaluation temperature (℃)

Leaching speed (kg/m2d)

Leaching rate (1/y)

Leaching period for the entire amount (y)10°C

1.0E-08

1.6E-08

6.4E+0725°C

5.0E-08

7.8E-08

1.3E+0760°C

1.0E-06

1.6E-06

6.4E+05

90°C

1.0E-04

1.6E-04

6.4E+03©International Research Institute for Nuclear Decommissioning

283

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Consolidation of input information (9/10)
The set values of some of the distribution coefficients of FY2020 based on the application for changing the JNFL Lowlevel radioactive waste burial center, were reviewed while referring to the latest domestic disposal trend (NUMO
Comprehensive Technical Report, etc.).
The values in red indicate the values
that were revised from last year

<Set values of distribution coefficient (Kd)>

Kd
H
Be
C
Cl
Ca
Co
Ni
Se
Rb
Sr
Zr
Nb
Mo
Tc
Pd
Ag
Cd
Sn
I
Cs

Cement
0
0.1
0.004
5E-04
0.009
0.1
0.009
0.01
0.01
0.009
0.1
0.1
0.01
0.0002
0.009
0.01
0.009
0.25
0.0001
0.01

Bentonite
0
0.004
0
0
0.01
0.004
0.004
0.01
0.1
0.01
0.1
0.1
0.0003
0
0.004
0.1
0.004
0.67
0
0.1

Rock/Soil00.01
0.0001
0
0.01
0.01
0.1
0.001
0.1
0.01
0.02
0.02
0.00010.0001
0.1
0.1
0.1
0.13
0
0.1

Farm soil
03
0.002
0.00025
0.11
0.99
1.1
1.8
0.67
0.15
7.3
2
0.027
0.0015
0.67
15
0.81
1.6
0.027
0.27

Kd
Ba
La
Pm
Sm
Eu
Tb
Ho
Pt
Pb
Po
Ra
Ac
Th
Pa
U
Np
Pu

Cement
0.009
0.009
0.009
0.009
0.009
0.009
0.009
0.009
0.009
0.009
0.009
0.1
0.4
0.4
0.02
0.009
0.4

Bentonite
0.01
0.004
0.004
0.004
0.004
0.004
0.004
0.004
0.004
0.004
0.01
1
0.03
0.03
0.009
0.004
0.03

Rock/Soil
0.01
0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.01
0.1
0.02
0.02
0.001
0.00090.02

Farm soil
0.06
0.65
1
3
3
0.65
3
0.09
22
6.6
2.4
5.4
89
6.6
0.4
1.2
1.8

Am
Cm

0.1
0.1

1
1

0.1
0.1

110
12

©International Research Institute for Nuclear Decommissioning284

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

References

Consolidation of input information (10/10)

The non-conservative set value was revised based on the NUMO Comprehensive Technical Report and
JAEA database.
FY2020 set
value

Status of setting the distribution coefficient Kd

The values in red indicate the values that wererevised from last year

NUMO Comprehensive Technical Report

JAEA

Neogene period Neogene period Neogene period Neogene period
sedimentary
sedimentary
sedimentary
sedimentary
rock
rock
rock
rock
Rocks, soil
Base caseBase case
Element(Based on theBase caseBase caseGr.3 （Up to
Gr.3 （Up to
JNFL application
HWL,
HWL,
1000 years after1000 years after
for change)
Gr.1，2，4
Gr.1，2，4
closing ofclosing of
Low Cl High Cl disposal site) disposal site)
Concentration Concentration
Low ClHigh ClConcentration Concentration
Co
0.1
3
0.5
3E-4
5E-5
Ni
0.1
3
0.5
3E-4
5E-5
Se
0.15
0.04
0.05
0.04
0.05
Sr
0.1
0.2
0.07
2E-4
7E-5
Nb
0.02
6
6
6
6
Sn
0.13
100
100
100
100
Cs
0.9
1
0.1
0.001
1E-4
Th
0.02
30
5
30
5
U
0.001
6
2E-6
6
2E-6
Np
0.0009
30
5
30
5
Pu
0.02
30
5
30
5
Am
0.1
200
20
200
20

Sandstone

Logarithmic
mean

0.08357
0.54220.028180.013611.774
2.433
0.1916
0.35660.22940.02128
1.019
1.124

Database

FY2021 set
value

Sandstone

Rocks, soil(Partiallyrevised based
Lower limiton the FY2020
value of the
JNFL
range with theapplication forlargest quantitychange)of data
0.01
0.1
0.01
0.01
0.1
0.1
0.1
0.1
0.01
0.001
0.1
1

0.01
0.1
0.001 *
0.01
0.02
0.13
0.1
0.02
0.001
0.00090.02
0.1*： The distribution coefficient for Se in rocks and soil is set conservatively at 0.001m 3/kg considering thatthe numerical values under oxidizing conditions for sandstone are distributed in the range of 0.001 to
0.01m3/kg according to the JAEA database.
©International Research Institute for Nuclear Decommissioning285

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of sensitivity evaluation (1/15)•

Multiple disposal concepts with depths corresponding to trench disposal, pit disposal, and medium
depth disposal were planned and models were created for sensitivity analysis. (Page 287 to 290)•

The parameters for sensitivity evaluation were set for each disposal concept. (Page 287 to 290)•

For the sensitivity evaluation, it was assumed that 1GBq nuclides are distributed evenly in the waste
filler material, the outlet nuclide migration rate at each barrier (filler material, low diffusion layer, lowpermeability layer, natural barrier) was compared, and the attenuation in the nuclide migration rate
once the nuclides have passed each barrier was evaluated. (Page 291 to 295)•

The sensitivity of the barrier composition and the sensitivity of the barrier parameters were organized.
(Page 296 to 300)

©International Research Institute for Nuclear Decommissioning286

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of sensitivity evaluation (2/15)<Overview of sensitivity evaluation>
The outlet nuclide migration rate at each barrier (filler material, low diffusion layer, low permeability layer, naturalbarrier) considered in groundwater migration in the case of trench disposal, pit disposal and medium depth disposal in
the groundwater scenario, was compared, and the attenuation in nuclide migration rate once the nuclides have passed
each barrier was understood. (Refer to Fig. 2)
An indicator for discussing the necessity of the abilities
required of barriers while evaluating each waste.

Low permeability layer
・ Setting the amount of water penetration
The reduction in the amount of water penetration through thebottom barrier layer was considered as well while setting

Evaluation
point ①

Evaluation
point ①

Source term

• Even distribution of radioactivity• The filler material is either sand Kd: C-14(0.0001) Cs137(0.1) or cement based material Kd: C-14(0.004)
Cs-137(0.01)

Low diffusion layer
Evaluation
point ②• Cement based material Kd : C-14(0.004) Cs-137(0.01)• Effective diffusion coefficient De=1E-11m2/sec)

Bq/y_C-14(L3)Waste / filling material(1GBq nuclide)

Evaluation
point ②

Evaluation point ③

Evaluation
point ④

Bottom barrier layer
Evaluation
point ③• Bentonite Kd : C-14(0) Cs-137(0.1)• Effective diffusion coefficient De=1E-10m2/sec)
Evaluation point ④

Natural barrier

Fig. 1Image of evaluation modelRiverst (year)Fig. 2 Image of nuclide migration rate at each evaluation point
The peak nuclide migration rate after the nuclides pass through each barrier was evaluated.
→ As there is delay during migration, the peak nuclide migration rate decreases due tobroadening and attenuation(Note) The barrier composition illustrated in this diagram is an example.
Models will be created according to the disposal concept (Refer to the following page onwards)©International Research Institute for Nuclear Decommissioning

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(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of sensitivity evaluation (3/15)＜Study of sensitivity analysis cases (Disposal concept equivalent to trench disposal)>Trench disposal evaluation model (Fig. 1), disposal concept (Fig. 2), and sensitivity analysis
parameters (Table 1) were established.
Drainage layer water
shielding sheet

Amount of water penetration:

Soil covering (Low permeation of water as compared tosurrounding ares)(Or water shielding structure as shown in the above figure)・Setting the amount of water penetration
The reduction in the amount of water
penetration through the bottom barrier layer
was considered as well while setting

Waste / fillingmaterial
Evaluation
point ①

Evaluation point②

Source term

• Even distribution of radioactivity• Filler material was sand
• Kd: C-14(0.0001) Cs-137(0.1)

Bottom barrier layer

Low permeability layer

Earth's surface

Low permeability layer

Evaluation point ①
Evaluation point ②

If a bottom
barrier layer is
added• Bentonite Kd : C-14(0) Cs-137(0.1)• Effective diffusion coefficient De=1E10m2/sec)

The effect of theupper barrier was
Waste layer
evaluated based on
the fluctuation in(Bottom barrier layer)
amount of water
Flow of groundwater
Evaluation point ③ After migrating
penetrationfrom the natural barrier

Fig. 2 Trench disposal concept
Table 1

List of sensitivity analysis parameters
Initial value
Facility configuration

Natural barrierRivers
Evaluation
point ③Fig. 1 Trench disposal evaluation model

100

W(m)

100

H(m)

5
1

x 10

x 0.1

x 0.01

Nuclide leaching rate

m 3/y/m 2
1/y

1

0.01

0.001

0.0001

Confinement period

y

0

100

1000

10000

Damage period

y

0

100

1000

10000

x 100

x 10

x 0.1

Amount of water penetration
Waste

Filler material

Filler material ingredients
Distribution coefficient Kd

Bottom barrier layer

Natural barrier

Sensitivity parameter/magnification

L(m)

Sand
m3/kg

Material

Set value

x 0.01

bentonite

Thickness

m

0.5

0

1

2

Distribution coefficient Kd

Set value

x 100

x 10

x 0.1

x 0.01

Migration distance

m3/kg
m

100

30

300

Distribution coefficient Kd

m3/kg

Set value

x 100

x 10

x 0.1

x 0.01©International Research Institute for Nuclear Decommissioning

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(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of sensitivity evaluation (4/15)<Study of sensitivity analysis cases (Disposal concept equivalent to pit disposal)>Pit disposal evaluation model (Fig. 1), disposal concept (Fig. 2), and sensitivity analysis
parameters (Table 1) was established.
Amount of water
penetration:

Amount of water
penetration:
Soilcovering

Low permeability layer

Groundwater levelLow permeability layer(Bentonite)

Filler mortar

・Setting the amount of water penetration
The reduction in the amount of water penetration
through the bottom barrier layer was considered
as well while setting

Filler mortar

Evaluation point ①

Evaluation point ①

Waste / filling material

Source term

Low permeability layer(Bentonite)
Waste
body

Waste
body

(Porous concrete)

Groundwater level

Soilcovering

Bedrock(Internal partition)

Concrete pit(Porous concrete)

Evaluation point ③ After migrating from the natural barrier

Bottom barrier
Bedrock(Internal partition)

Evaluation point ②
Concrete pit

Evaluation point ③ After migrating from the natural barrier

Evaluation • Even distribution of radioactivity
point ①
• Filler material was cement based material
Kd: C-14(0.004) Cs-137(0.01)

Bottom barrier layer

Fig. 2 Pit disposal concept (left: without bottom barrier, right: with bottom barrier)If a bottom
barrier layer is
added

Table 1

Initial value

Evaluation • Bentonite Kd : C-14(0) Cs-137(0.1)
point ②
• Effective diffusion coefficient De=1E10m2/sec)

Facility configuration

Waste

RiversNatural barrier

List of sensitivity analysis parameters

Filler material
Bottom barrier layer

Fig. 1 Pit disposal evaluation model
Natural barrier

100

W(m)

100

Sensitivity parameter/magnification

H(m)

5

Amount of water penetration

m 3/y/m 2

0.1, 0.05

x 10

x 0.1

x 0.01

Nuclide leaching rate

1/y

1

1.E-02

1.E-03

1.E-04

Confinement period

y

0

100

1000

10000

Damage period

y

0

100

1000

10000

x 100

x 10

x 0.1

Filler material ingredients
Distribution coefficient Kd

Evaluation point ③L(m)Cement typem 3/kgMaterial

Set value

x 0.01

bentonite

Thickness

m

1

0

0.5

2

Distribution coefficient Kd

m 3/kgSet value

x 100

x 10

x 0.1

x 0.01

Migration distance

m

100

30

300

m 3/kgSet value

x 100

x 10

x 0.1

x 0.01

Materialrock

Distribution coefficient Kd©International Research Institute for Nuclear Decommissioning289

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Implementation of sensitivity evaluation (5/15)<Study of sensitivity analysis cases (Disposal concept equivalent to medium depth disposal)>Medium depth disposal evaluation model (Fig. 1), disposal concept (Fig. 2), and sensitivity analysis parameters(Table 1) were established.
Amount of water penetration:

Low permeability layer
Backfill・Setting the amount of water penetration
The reduction in the amount of water that seeps
through the bottom barrier layer is considered as wellwhile setting

Waste / fillingmaterial
Evaluation
point ①

(Low permeability layer)(Low diffusion layer)

Flow of groundwater

Source term

Waste body /Filler mortar

Backfill• Even distribution of radioactivity• Filler material was cement based material

Evaluation
point ①
Evaluation
point ②
Evaluation
point ③
Backfill

Concrete pit(Low permeability layer)

Low diffusion layer
Evaluation
point ②(Invert)• Cement based material Kd : C-14(0.004) Cs-137(0.01)• Effective diffusion coefficient De=1E-11m2/sec)

Evaluation
point ③Fig. 2 Medium depth disposal concept
Table 1

Bottom barrier layer

List of sensitivity analysis parameters
Initial value
Facility configuration• Bentonite Kd : C-14(0) Cs-137(0.1)• Effective diffusion coefficient De=1E10m2/sec)
Waste

Natural barrier

Evaluation point ④ After migratingfrom the natural barrierRiversFiller material

Evaluation point ④

Fig. 1 Medium depth disposal evaluation model

Bottom barrier layer

Natural barrier

100

W(m)

100

Sensitivity parameter/magnification

H(m)

5

Amount of water penetration

m 3/y/m 2

0.16, 3.2E-03

x 10

x 0.1

x 0.01

Nuclide leaching rate

1/y

1

1.E-02

1.E-03

1.E-04

Confinement period

y

0

100

1000

10000

Damage period

y

0

100

1000

10000

x 100

x 10

x 0.1

Filler material ingredients
Distribution coefficient Kd

Low diffusion layer

L(m)Cement typem 3/kgMaterial

Set value

x 0.01

Cement type

Thicknessm

0.5

Diffusion coefficientm 2 /y1E-11

Material

0

bentonite

Thickness

m

1

0

0.5

2

Distribution coefficient Kd

3

m /kg

Set value

x 100

x 10

x 0.1

x 0.01

Migration distance

m

300

100

900

m 3/kgSet value

x 100

x 10

x 0.1

x 0.01

Material
Distribution coefficient Kd

rock©International Research Institute for Nuclear Decommissioning290(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods
Implementation of sensitivity evaluation (6/15)<Sensitivity evaluation of barrier structure>
Sensitivity of each nuclide with respect to impact of barrier structure was evaluated for typical nuclides (result of studies
from last year) that contribute to the dose from the groundwater evaluation picked out in Table 1. The nuclides were
evaluated by grouping them as shown below.
⚫ C-14, Tc-99, I-129: Long half-life (up to 5000y), delayed attenuation due to sorption is not expected as much
⚫ Cs-135, Pu-239: Long half-life (up to 5000y), delayed attenuation due to sorption is expected to some extent⚫ Sr-90, Cs-137: Short half-life, delayed attenuation due to sorption is expected to some extent

Table 1

Nuclides
C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239Distribution coefficient of each nuclide for each barrier material Kd

Half-life (y)5700
2921110015700000
2300000
30
24110

Distribution coefficient Kd (m3/kg)
CementBentonite
Rocks, soil, sand0.004
00.0001
0.009
0.01
0.01
0.000200.00010.0001
0
0
0.01
0.1
0.1
0.01
0.1
0.1
0.4
0.03
0.02

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Implementation of sensitivity evaluation (7/15)
Trench disposal<Evaluation of the impact of barrier structure in trench disposal*>
The sensitivity to leaching rate, and controlling the amount of water penetration
was evaluated, and it was verified that C-14, Tc-99, I-129 reduce due toleaching rate and because of controlling the amount of water penetration, butthat there is almost no delay during migration (Fig. 2)

Low permeability layer

Earth's surface

Soil covering (Low permeation of water as compared tosurrounding areas)(Or water shielding structure as shown in the above figure)

Waste layer

Evaluation point ①
Evaluation point ②(Bottom barrier layer)
Flow of groundwater*: In this document only trench disposal is introduced as a typical disposal method.

Delay has only aminor impact

Leaching
Evaluation point ①
rate

If a bottom
barrier layer is
added

Evaluation point ②

Evaluation point ①

Evaluation point ②

Evaluation point ③

Evaluation point ① Evaluation point ② Evaluation point ③

Evaluation
point ③
Outlet nuclide
migration rate

Natural barrier

Reduced due to theleaching rate andbecause the amountof water penetration iscontrolled

Outlet nuclide
migration rate

Source term

Outlet nuclide
migration rate

Leaching
rate

Bottombarrier layer

Evaluation point ③ After migratingfrom the natural barrier

C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239Amount of water
penetration

Amount of water penetration

Waste fillingmaterial

Drainage layer
Leak-isolation
sheets

Red: Parameters that werecaused to fluctuateFig. 1 Trench disposal evaluation model

Reduced due to the leaching rate and because theamount of water penetration is controlled
There is almost no delay during migration
Evaluation point ① Evaluation point ②

Fig. 2

Evaluation point ③

Results of evaluation of sensitivity to amount of water penetration and leaching rate
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Implementation of sensitivity evaluation (8/15)<Evaluation of the impact of barrier structure in trench disposal*>
The sensitivity to leaching rate, and controlling the amount of water penetration
was evaluated, and it was verified that C-135, Pu-239 reduce due to leachingrate and because of controlling the amount of water penetration, and in addition,that there is attenuation due to sorption (In particular, the waste filled layer has amajor impact) (Fig. 2).

Trench disposal
Low permeability layer

Earth's surface

Drainage layer
Leak-isolation
sheetsSoil covering (Low permeation of water as compared tosurrounding areas)(Or water shielding structure as shown in the above figure)

Waste layer

Evaluation point ①
Evaluation point ②(Bottom barrier layer)
Flow of groundwater

Evaluation point ③ After migratingfrom the natural barrier

C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239*: In this document only trench disposal is introduced as a typical disposal method.
Amount of water penetration

Waste fillingmaterial

Leaching
rate

Leaching
rate

Bottombarrier layer
Evaluation point②

Natural barrier

Evaluation point ①

Evaluation point ②

Evaluation point ③

Nuclide migration rate reduces at the filler material outlet→ Large impact of sorption in the filler material
If a bottom
barrier layer is
added

Evaluation
point ③

Red: Parameters that were caused to fluctuateFig. 1 Trench disposal evaluation model

Outlet nuclide
migration rate

Evaluation point①

Sourceterm

Outlet nuclide
migration rate

Amount of water
penetration

Evaluation point ①

Evaluation point ②

Evaluation point ③

Fig. 2 Results of evaluation of sensitivity to amount of water penetrationand leaching rate
Reduced due to the leaching rate and because the amount of water penetration is controlled.
It was confirmed that there was attenuation due to sorption (In particular, the waste filled later
has a major impact)©International Research Institute for Nuclear Decommissioning293

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Implementation of sensitivity evaluation (9/15)<Evaluation of the impact of barrier structure in trench disposal*>
The sensitivity to leaching rate, and controlling the amount of water penetration
was evaluated, and it was verified that Sr-90, Cs -137 reduce due to leachingrate and controlling the amount of water penetration, and that the peak largely
reduces at each barrier due to delayed attenuation. (Fig. 2)*: In this document only trench disposal is introduced as a typical disposal method.

Trench disposal
Low permeability layer

Earth's surface

Drainage layer
Leak-isolation
sheetsSoil covering (Low permeation of water as compared tosurrounding areas)(Or water shielding structure as shown in the above figure)

Waste layer

Evaluation point ①
Evaluation point ②(Bottom barrier layer)

Flow of groundwater

C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239Evaluation point ③ After migratingfrom the natural barrier

Amount of water penetration
Amount of water
penetration

Leaching
rate

Source term
Evaluation point ①

Leaching
rate

Outlet nuclide
migration rate

Waste fillingmaterial

Reduced due to sorption
into the filler material
Evaluation point ①

Bottombarrier layer

If a bottom
barrier layer is
added

Natural barrier

Evaluation point ③

Peak largely reduces at eachbarrier due to delayed
attenuation
Outlet nuclide
migration rate

Evaluation point ②

Evaluation point ②

Evaluation point ③
Evaluation point ①

Red: Parameters that were caused to fluctuate

Evaluation point ②

Evaluation point ③

Fig. 2 Results of evaluation of sensitivity to amount of water
penetration and leaching rate

Fig. 1 Trench disposal evaluation model©International Research Institute for Nuclear Decommissioning294

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Implementation of sensitivity evaluation (10/15)

Summary of evaluating sensitivity to barrier structure○ Nuclides with low sorption property and long half-life (C-14, Tc-99, I-129)➢ The engineered barrier is effective in controlling leaching and controlling the amount ofwater penetration.
○ Nuclides with sorption property and long half-life (Cs-135, Pu-239)➢ The engineered barrier is effective in controlling leaching and controlling the amount ofwater seepage.
➢ The engineered barrier and natural barrier are expected to lead to delayed attenuation toa certain extent.
○ Nuclides with sorption property and short half-life (Sr-90, Cs-137)➢ Leaching and amount of water penetration from waste is expected to be controlled, but
the effect of delayed attenuation due to the engineered barrier and natural barrier isdominant.

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Implementation of sensitivity evaluation (11/15)<Evaluation of sensitivity to the parameters of the barrier>Sensitivity of each nuclide with respect to the parameters of the barrier (amount of water penetration, leaching rate, confinement
period, Kd of each barrier, migration distance) was evaluated for typical nuclides (result of studies from last year) that contribute to
the dose from the groundwater evaluation picked out in Table 1.
The nuclides typical to each group were evaluated as shown below.
⚫ C-14 : Low sorption property, long half-life nuclide. Half life is long (up to 5000y), not much delayed attenuation is expected
due to sorption.
⚫

Cs-135: Sorption property, long half-life nuclide. Half life is long (up to 5000y), some delayed attenuation is expected due tosorption.

⚫

Sr-90: Sorption, short half-life nuclide. Half-life is short, some delayed attenuation is expected due to sorption.

Table 1

Distribution coefficient of each nuclide for each barrier material Kd

NuclidesDistribution coefficient Kd (m3/kg)

Half-life (y)

C-14
Sr-90
Tc-99
I-129
Cs-1355700
2921110015700000
2300000

Cement0.004
0.009
0.00020.00010.01

Cs-137
Pu-239

30
24110

0.01
0.4

Bentonite
0
0.01
0
0
0.1

Rocks, soil
0.00010.01
0.0001
0
0.1

0.1
0.03

0.1
0.02

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Implementation of sensitivity evaluation (12/15)
Earth's surface<Evaluation of the impact of barrier parameters in trench disposal*>
The sensitivity to barrier parameters was evaluated, and it was foundthat amount of water penetration, filler material Kd, leaching rate,
damage period have a major impact on C-14. (Fig. 2)

Drainage layer
Leak-isolation
sheets

C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239Fig. 1 Conceptual diagram of
trench disposal
Low permeability layer

Soil covering (Low permeation of water as compared tosurrounding areas)(Or water shielding structure as shown in the above figure)

Waste layer

Evaluation point ①
Evaluation point ②(Bottom barrier layer)

Flow of groundwater

Evaluation point ③ After migratingfrom the natural barrier*: In this document typical examples of trench disposal and medium depth
disposal are introduced.

◆ is the reference case
Nuclidemigration rate

Evaluation of the natural barrier outlet(Evaluation point ③)(Refer to Fig. 1)

Amount of water
penetration

Bottom barrier
thickness

Natural barrier migration
distance

Table 1 List of sensitivity analysis parameters(Repeated)
Initial value
Facility configuration

Waste

Filler material
Kd

Bottom barrier Kd

Natural barrier Kd
Filler material

100

W(m)

100

H(m)

5

Amount of water penetration

m3/y/m2

1

x 10

x 0.1

x 0.01

Nuclide leaching rate

1/y

1

0.01

0.001

0.0001

Confinement period

y

0

100

1000

10000

Damage period

y

0

100

1000

10000

x 100

x 10

x 0.1

Filler material ingredients
Distribution coefficient Kd

Bottom barrier layer

Natural barrier

Leaching rate

Confinement period

Damage period

Sensitivity parameter/magnification

L(m)

Sand
m3/kg

Material

Set value

x 0.01

bentonite

Thickness

m

0.5

0

1

2

Distribution coefficient Kd

m3/kg

Set value

x 100

x 10

x 0.1

x 0.01

Migration distance

m

100

30

300

Distribution coefficient Kd

m3/kg

Set value

x 100

x 10

x 0.1

x 0.01

・・・Parameters to which thenuclides are highly sensitiveFig. 2 Results of evaluating the sensitivity to each parameter©International Research Institute for Nuclear Decommissioning

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Implementation of sensitivity evaluation (13/15)

C-14
Sr-90
Tc-99
I-129
Evaluation point ①
Cs-135
Evaluation point ②
Cs-137
Pu-239Fig. 1 Conceptual diagram of trench disposal<Evaluation of the impact of barrier parameters in trench disposal*>
The sensitivity to barrier parameters was evaluated, and it was found thatamount of water penetration, leaching rate, leaching rate, confinement period
damage period have a major impact on Sr-90. (Fig. 2)

Drainagelayer
Leak-isolation
sheetsLow permeability layer

Earth's surface

Soil covering (Low permeation of water as compared tosurrounding areas)(Or water shielding structure as shown in the above figure)

Waste layer(Bottom barrier layer)

Flow of groundwater

Evaluation point ③ After migratingfrom the natural barrier*: In this document typical examples of trench disposal and medium depth
disposal are introduced.

◆ is the reference case
Nuclidemigration rate

Evaluation of the natural barrier outlet(Evaluation point ③)(Refer to Fig. 1)

Amount of water
penetration

Bottom barrier thickness

Natural barrier migration
distance

Table 1 List of sensitivity analysis parameters(Repeated)
Initial value
Facility configuration

Waste

Filler material Kd

Bottom barrier Kd

Natural barrier Kd
Filler material

100

W(m)

100

H(m)

5

Amount of water penetration

m3/y/m2

1

x 10

x 0.1

x 0.01

Nuclide leaching rate

1/y

1

0.01

0.001

0.0001

Confinement period

y

0

100

1000

10000

Damage period

y

0

100

1000

10000

x 100

x 10

x 0.1

Filler material ingredients
Distribution coefficient Kd

Bottom barrier layer

Natural barrier

Leaching rate

Confinement period

Damage period

Sensitivity parameter/magnification

L(m)

Sand
m3/kg

Material

Set value

x 0.01

bentonite

Thickness

m

0.5

0

1

2

Distribution coefficient Kd

m3/kg

Set value

x 100

x 10

x 0.1

x 0.01

Migration distance

m

100

30

300

Distribution coefficient Kd

m3/kg

Set value

x 100

x 10

x 0.1

x 0.01

・・・Parameters to which thenuclides are highly sensitiveFig. 2 Results of evaluating the sensitivity to each parameter©International Research Institute for Nuclear Decommissioning

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Implementation of sensitivity evaluation (14/15)

C-14
Sr-90
Tc-99
I-129
Cs-135
Cs-137
Pu-239Fig. 1 Conceptual diagram of
medium depth disposal
Backfill<Evaluation of the impact of barrier parameters in medium depth
disposal*>
The sensitivity to barrier parameters was evaluated, and it was foundthat amount of water penetration and filler material Kd, have a major
impact on C-14. (Fig. 2)(Low permeability layer)(Low diffusion layer)

Flow of groundwater

Evaluation point ①

Evaluation point ②

Backfill

Waste body /Filler mortar

Evaluation point ③

Backfill

Concrete pit(Low permeability layer)
Invert*: In this document typical examples of trench disposal and medium depth
disposal are introduced.

Evaluation point ④ Aftermigrating from the naturalbarrier

Nuclidemigration rate

Evaluation of the natural barrier outlet(Evaluation point ④) (Refer to Fig. 1)◆ is the reference caseAmount ofwater
penetration

Bottom barrier thickness

Natural barrier migration
distance●： Cases that do not consider controlling water
penetration(The permeability coefficient (1E-7) for the areaaround the facility is considered to be the sameas that for rocks.)
Table 1 List of sensitivity analysis parameters (Repeated)
Initial value

Filler material
Kd

Facility configuration

Bottom barrier Kd

Natural barrier Kd

100

H(m)

Filler material

Amount of water penetration

m /y/mNuclide leaching rate
Confinement period
Damage period

Confinement period

Damage period

Bottom barrier layer

Natural barrier

Fig. 2 Results of evaluating the sensitivity to each parameter

520.16, 3.2E-03

x 10

x 0.1

x 0.011/y

1

1.E-02

1.E-03

1.E-04

y

0

100

1000

10000

y

0

100

1000

10000

x 100

x 10

x 0.1

Filler material ingredients
Distribution coefficient Kd

Low diffusion layer

3

Sensitivity parameter/magnification

100

W(m)

Waste

Leaching rateL(m)Cement type
3

m /kg

Material

Set value

x 0.01

Cement type

Thicknessm

Diffusion coefficientm 2/yMaterial

0.5

0

1E-11
bentonite

Thickness

m

1

0

0.5

2

Distribution coefficient Kd

m 3/kgSet value

x 100

x 10

x 0.1

x 0.01

Migration distance

m

300

100

900

x 100

x 10

x 0.1

x 0.01

Material
Distribution coefficient Kdrockm 3/kgSet value

・・・Parameters to which the nuclides are highly sensitive©International Research Institute for Nuclear Decommissioning

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Implementation of sensitivity evaluation (15/15)
Summary of evaluating sensitivity to barrier parameters
The sensitivity of each nuclide to the barrier parameters was evaluated, and the parameters to which
each nuclide is highly sensitive were identified. (Table 1)

Table 1

Parameters to which the nuclides are highly sensitive
TrenchPit

Medium depth

Nuclides with low sorption
property and long half-life(C-14 was evaluated as a
typical nuclide)Amount of water
penetration
Filler material Kd
Leaching rate
Damage period

Amount of water
penetration
Filler material Kd
Leaching rate
Damage period

Amount of water
penetration
Filler material Kd

Nuclides with sorption
property and long half-life(Cs-135 was evaluated asa typical nuclide)Amount of water
penetration
Filler material Kd
Leaching rate

Amount of water
penetration―(Does not become thecontrolling nucide even if
there is no engineered
barrier)

Nuclides with sorption
property and short half-life(Sr-90 was evaluated as a
typical nuclide)Amount of water
penetration
Leaching rate
Confinement period
Damage timing―(Does not become thecontrolling nucide even if
there is no engineered
barrier)―(Does not become thecontrolling nucide even if
there is no engineered
barrier)©International Research Institute for Nuclear Decommissioning300

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Preliminary proposals of disposal concepts (1/8)•

The method for deriving the disposal category for each waste based on the dose
obtained through sensitivity evaluation was studied. (Page 302 to 303)•

Preliminary proposals of disposal concepts for each waste were consolidated onreceiving the results of sensitivity analysis.
(Page 304 to 308)

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Preliminary proposals of disposal concepts (2/8)
The disposal category for each waste was studied using the dose evaluation methods studied until last year (In thetable below, there is no engineered barrier and the inventory is the MODE value). The 12 wastes selected until lastyear were studied.
In the case of waste for which the disposal category is determined based on the nearing scenario, even if anengineered barrier is installed, it is difficult to revise the disposal category. Hence, wastes for which disposal category
is determined based on the groundwater scenario when there is no engineered barrier, were identified. (Refer toTable 1)
Table 1Results of investigating disposal categories
Nearing scenario(Unit: μSv/y)Groundwater scenario (without barrier)Waste for which the disposal
category is determined bygroundwater evaluation

Control period
Waste name

400y

400y

Disposal category

L3

L2

100,000y
L1

Disposal category

L3

L2

L1

KURION

L1

2.1E+05

8.8E+04

1.5E+03

L1-strata

2.3E+03

6.9E+01

9.3E+00

AREVA

L1

6.9E+06

2.6E+06

4.3E+03L2

7.5E+03

4.3E-02

5.8E-03

✓

SARRY

L1

7.2E+05

3.0E+05

2.7E+03

L1-strata

3.3E+04

5.4E+01

7.3E+00✓

Slurry 1_iron co-precipitation

L1

4.7E+05+05.9E+1

6.9E+02

L1-strata+04.9E+1+01.9E+1

2.6E+00✓

Slurry 2_Carbonate precipitation

L1

2.7E+04

1.0E+04

4.8E+01

L2-strata

4.3E+03

7.9E+00

1.1E+00✓

Rubble 1

L3

3.0E-01

1.4E-01

8.4E-03L3

8.7E-01

2.6E-02

3.5E-03

Rubble 2

L3

1.0E+02

4.7E+01+00.9E+2

L2 - strata

2.8E+028.2E+00

1.1E+00✓

Rubble 5L1

5.3E+03

2.5E+03

1.5E+02

Strata8.8E+03

2.6E+02

3.5E+01✓

RPV waste(L1 before the accident)

Strata9.1E+073.8E+07

4.0E+06

Strata1.1E+05

3.3E+03

4.4E+02

PCV waste (metal)(L2 before the accident)

Strata+06.6E+4

1.8E+06

2.0E+05

L1

4.7E+03

3.1E+01

6.5E-01

Waste inside the building (metal)(L3 before the accident)

L2

7.0E+02

3.3E+02

3.8E+00

Strata3.2E+03

9.4E+01

1.3E+01✓

Waste inside the building (Concrete)(Non-radioactive waste before the accident)

L3+02.9E+1

7.2E+01

8.0E+00

L2 - L1

3.2E+02

4.8E+00

6.5E-01✓

Dose criteria in the nearing scenario

Exceeds the dose criteria

Dose criteria in the groundwater scenario

For waste that does not exceed the dose criteria in the groundwater
scenario, considering the weight of other nuclides and waste, thewaste that exceeds the approximate dose of 1μSv/h are color coded
red.

*: Disposal category with the least depth wherein dose criteria is satisfied in the nearing scenario and approximate dose is satisfied in the
groundwater scenario.

©International Research Institute for Nuclear Decommissioning302

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Preliminary proposals of disposal concepts (3/8)
If there is no engineered barrier, in the case of waste for which the disposal category is determined based on groundwater scenario,
barriers or barrier performance that can control the dose up to approximate dose: 1μSv/y should be selected. The method forselecting barriers or barrier performance for Rubble 5 as an example of relatively low dose rubble, and for KURION as an exampleof high dose waste generated from contaminated water treatment is provided from next page onwards.

Table 1

Dose and main nuclides in each disposal concept(Unit: μSv/y)Groundwater scenario Person using rivers Without barrier

Nearing
scenario

L1

L2

L3

Waste name

Disposal category

Nuclides1/10 <

Dose

Nuclides1/10 <

Dose

NuclidesKURION

L1

C14

Sr90

2.3E+03

C14

I129

6.9E+01

C14

AREVA

L1

Sr90

7.5E+03

C14

Se79, I129

4.3E-02

C14

I129

1/10 <

Dose
9.3E+00

Se79

5.8E-03

SARRY

L1

Sr90

3.3E+04

C14

5.4E+01C14

7.3E+00

Slurry 1_Iron co-precipitation

L1

Sr90

+04.9E+1

C14+01.9E+1

C14

2.6E+00

Slurry 2_Carbonate precipitation

L1

Sr90

4.3E+03

C14

7.9E+00

C14

1.1E+00

Rubble 1

L3

C14

8.7E-01C14

2.6E-02

C14

3.5E-03

Rubble 2

L3

C14

2.8E+02C14

8.2E+00

C14

1.1E+00

Rubble 5

L1

C14

8.8E+03C14

2.6E+02

C14

3.5E+01

RPV waste(L1 before the accident)

StrataC14

Sr90

1.1E+03C14

3.3E+03C14

4.4E+02

PCV waste (metal)(L2 before the accident)

Strata

Sr90

Pu240, Pu239
Pu238, Am241

4.7E+03

Pu239

3.1E+01

Pu239

Waste inside the building (metal)(L3 before the accident)

L2

C14

3.2E+03C14

9.4E+01C14

1.3E+01

Waste inside the building(concrete)(non-radioactive waste before theaccident)

L3

Sr90

3.2E+02

C14

4.8E+00

C14

6.5E-01

C14

Pu240

I129

Np237, C14
U234, Po210

6.5E-01

Disposal concept in the groundwater scenario that exceeds approximate dose: 1μSv/y from among the disposal concepts inwhich disposal is carried out at a depth that is shallower than the nearing scenario©International Research Institute for Nuclear Decommissioning303

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Preliminary proposals of disposal concepts (4/8)<Application of sensitivity evaluation to KURION>In the nearing scenario the disposal category of KURION is medium depth disposal, but if disposal category
is medium depth disposal in the groundwater scenario, the approximate dose (1μSv/y or lower) is exceeded.
(Refer to Table1)
The dominant nuclide in the groundwater scenario is C-14. The barrier or barrier performance for reducing
the dose of 9.3μSv/y to the approximate dose or lower, should be studied based on the sensitivity evaluation
results. (Refer to Fig. 1)
Table1

Dose for each disposal concept in the groundwater scenario
Groundwater scenario Person using rivers (Without barrier) Target dose 1μsV/yNearing
scenario

L3

KURION

L1

Disposal category

Nuclides1/10 <

Dose

Nuclides1/10 <

Dose

Nuclides

L1

C14

Sr90

2.30E+03

C14

I129

6.90E+01

C14

Kurion"Dominant nuclides ingroundwater scenario: C-14""Dose in groundwater scenario: 9.3μSv/y (C-14)"

1/10 <

Dose
9.30E+00

Exposure dose whenriver water is used

Waste name

L2

Fig. 1 Time dependence of dose when there is no L1 barrier©International Research Institute for Nuclear Decommissioning304

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Preliminary proposals of disposal concepts (5/8)<Application of sensitivity evaluation to KURION>In the case of KURION, in order to reduce the dose of 9.3μSv/y to the approximate dose: 1μSv/y or lower, based on
the results of sensitivity evaluation, the amount of water penetration at the bentonite layer needs to be kept at about
3.2E-3(㎥/y/㎡) or lower. (Fig. 1) (Water penetration control barrier is essential.) Since even the sensitivity ofconcentration reduction at the filler material due to sorption is comparatively high, it is determined that it is importantto take actions that would not have an adverse effect on filler Kd, such as limiting the materials with impact on
disposal, etc. (Fig. 2)
Cases that do not consider controlling water
penetrationis the reference

Table 1

Nuclidemigration rate

Low diffusion
layer absent

Nuclide migration
rate equivalent to1μSv/yLow diffusion
layer present

Amount of water penetration

Nuclidemigration rate

Fig. 1 Dependence on amount of water penetration

Waste and filler material Kd

Change in dose due to change in the amount of water
penetration and nuclide migration rate
Nobarrier

Reference

TargetAmount ofwater
penetration(m3/y/m2)

0.16

0.00320.0032

Nuclidemigrationrate(Bq/y)

6.5E+5

7.0E+4

7.0E+4Dose(μSv/year)

9.3

1.0

1.0

In order to keep the dose equivalent to 1μSv/year or
lower (approx. 1/10 or lower), the amount of water
penetration at the bentonite layer needs to be kept at
about 3.2E-3 (m3/y/m2) (permeability coefficient 1E10m/s) or lower.

Fig. 2 Dependence on waste and filler material Kd©International Research Institute for Nuclear Decommissioning305

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Preliminary proposals of disposal concepts (6/8)<Application of sensitivity evaluation to Rubble 5>
In the nearing scenario the disposal category of Rubble 5 is medium depth disposal, but if disposal category is
medium depth disposal in the groundwater scenario, the approximate dose (1μSv/y or lower) is exceeded. (Refer toTable 1)The dominant nuclide in the groundwater scenario is C-14. The barrier or barrier performance for reducing the doseof 35μSv/y to the approximate dose: 1 μSv/y or lower, should be studied based on the sensitivity evaluation results.
(Refer to Fig. 1)
Table1

Dose for each disposal concept in the groundwater scenario
Groundwater scenario

Nearing
scenario
Waste nameRubble 5

Person using rivers

L3

Disposal category

NuclidesL1

C14

1/10 <(Without barrier)Target dose 1μSv/y

L2
Dose

Nuclides

8.80E+03

C14

L1

1/10 <

Dose

Nuclides

2.60E+02

C14

1/10 <

Dose
3.50E+01

Exposure dose whenriver water is used

Rubble 5
Dominant nuclides in groundwater scenario: C-14
Dose in groundwater scenario: 35μSv/y (L1)

Fig. 1 Time dependence of dose when there is no L1 barrier©International Research Institute for Nuclear Decommissioning

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Preliminary proposals of disposal concepts (7/8)<Application of sensitivity evaluation to Rubble 5>
In the case of Rubble 5, in order to reduce the dose of 35μSv/y to the approximate dose: 1μSv/y or lower, based on
the results of sensitivity evaluation, the amount of water penetration at the bentonite layer needs to be kept at about
1E-3(㎥/y/㎡) or lower (Fig. 1) (Water penetration control barrier is essential). Since even the sensitivity ofconcentration reduction at the filler material due to sorption is comparatively high, it is determined that it is importantto take actions that would not have an adverse effect on filler Kd, such as limiting the materials with impact on
disposal, etc. (Fig. 2)
Table 1 Change in dose due to change in the amount of water penetrationand nuclide migration rateis the reference

Cases that do not consider
controlling water penetration

No barrier Reference

Nuclidemigration rate

Low diffusion
layer absent(Low diffusion
barrierpresent)

Nuclidemigration rate

Nuclide migration
rate equivalent to1μSv/yLow diffusion
layer present

Amount of water penetration

Dependence on amount of water
penetration

Dose(μSv/year)

Controlling the amount of water penetration has amajor dose reduction effect.

Nuclidemigration rate

Fig. 1

Amount of water
penetration

Targetis the reference case

Waste and filler material Kd

Fig. 2 Dependence on waste and filler material Kd©International Research Institute for Nuclear Decommissioning307

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Preliminary proposals of disposal concepts (Summary) (8/8)
Preliminary proposals of disposal concepts for each waste on receiving the results of sensitivity
analysis are indicated in Table 1.

Table 1 Preliminary proposals of disposal concepts for each waste
Waste name

Disposal
category
preliminary
proposal

Scenario for
determining
the disposal
category

Dominant nuclides innearing scenario

Dominant nuclides ingroundwater scenario

Waterpenetration
controlbarrier

Amount of water
penetration(m3/y/m2)

Filler material Kd (*2)(m3/kg)KURION

L1

Ground water

I-129, Cs-135 (>10%)

C-14, I-129 (>10%)

Required

3E-3 or less0.004 or more

AREVA

L1

Nearing

Se-79, Tc-99 (>10%)

C-14, Se-79 (>10%)

Not required－

0.004 or more

SARRY

L1

Ground water

I-129, Cs-135 (>10%)

C-14, I-129 (>10%)

Required

3E-3 or less0.004 or more

Slurry 1_iron co-precipitation

L1

Ground water

Se-79, Tc-99 (>10%)

C-14, I-129 (>1%)

Required

3E-3 or less0.004 or more

Slurry 2_Carbonate precipitation

L1

Ground water

Se-79, Tc-99 (>10%)

C-14, I-129 (>1%)

Required

3E-3 or less0.004 or more

Rubble 1

L3

Nearing/
GroundwaterCs-137, Sr-90 (<1%)

C-14, I-129 (>1%)

Not required－

0.004 or more

Rubble 2

L2

GroundwaterCs-137, C-14 (>1%)

C-14, I-129 (>1%)

Required

1E-2 or less0.004 or more

Rubble 5

L1

Groundwater

Tc-99, Cl-36 (>10%)

C-14, I-129 (>1%)

Required

1E-3 or less0.004 or more

Tc-99, Ra-226 (>10%)

C-14, TRU (up to 10%)

Required

1E-4 or less0.004 or more

Tc-99, Ra-226 (>10%)

TRU

Not required－－

RPV waste (L1 before the accident)

Strata (*1)

PCV waste (Metal) (L2 before the accident)

Strata (*1)

Nearing

Nearing

Waste inside the building (Metal)(L3 before the accident)

L1

Ground water

C-14, Cs-137 (>10%)

C-14, I-129 (<1%)

Required

3E-3 or less0.004 or more

Waste inside the building (Concrete)(Non-radioactive waste before the accident)

L2

Nearing/
GroundwaterCs-137, Tc-99 (>10%)

C-14, I-129 (>10%)

Required

1E-2 or less0.004 or more

*1: In this evaluation, waste that cannot be disposed at L1 is tentatively assumed to be disposed geologically. L1 values are provided for dominant nuclides in
the groundwater scenario.
*2: Filler material Kd for dominant nuclides in groundwater scenario©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (1/14)•

Dose evaluation was carried out for the disposal concept proposals obtained as a result of sensitivity
evaluation. (Only groundwater scenario) (Page 310 to 318)•

Dose evaluation was conducted using 2 models, namely, “Conservative conceptual model (Analysis model
similar to the models that have a domestic track record)” and “Cautiously realistic conceptual model” using thesame parameters as the sensitivity evaluation, and the results were compared. (Page 310 to 318)•

Based on the results of dose evaluation using the conservative model (Analysis model similar to models thathave a domestic track record), it was projected that the disposal concept that was determined throughsensitivity evaluation was generally satisfactory when its dose was evaluated. (Page 319 to 321)•

According to the results of dose evaluation using the cautiously realistic conceptual model, the maximum dose
showed a decline. It was verified that the components of the disposal concepts have a major impact on dose
evaluation, in particular when diffusion of waste / backfill and ageing degradation are considered. (Page 317 to318)©International Research Institute for Nuclear Decommissioning

309

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Implementation of dose evaluation (2/14)
・ Dose evaluation was carried out for the disposal concept proposals obtained as a result of sensitivity evaluation.
(Only groundwater scenario)
・ Dose evaluation was conducted using 2 models, namely, “Conservative conceptual model (Analysis model
similar to the models that have a domestic track record)” and “Cautiously realistic conceptual model” using
the same parameters as the sensitivity evaluation, and the results were compared. (Refer to Fig. 1)
Disposal concept proposals(Created based on the results ofsensitivity analysis)

Dose evaluation
results

Dose evaluation model
Conservative conceptual model

Inventory/ Parameters

Cautiously realistic conceptual modelFig. 1 Comparison of the broad outline of the dose evaluation models©International Research Institute for Nuclear Decommissioning310

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Implementation of dose evaluation (3/14)
・Based on the results of investigation of overseas disposal cases carried out from FY2017 - FY2019, it was decided to use the cautiously
realistic conceptual model for dose evaluation on a trial basis.
・The cautiously realistic conceptual model adopts an approach that is based on safety cases from the UK. It consists of “Observation”,“Process model”, “Numerical model”, and “Data elicitation” and is updated in accordance with updates in safety cases based on latest
findings. (Refer to Fig. 1)

Observation

Process model

Numerical model

Cautiously
realistic
conceptual
model

Data elicitation

Parameters of the cautiously realistic model at the waste
layer
Grout part: Cement (80vol%) + water (20vol%)Grout part diffusion coefficient: 1E-11m2/s
Interval in cracks: 1m, Width of crack: 0.001mm
Flow inside the cracks: Cross-sectional area of facility (m2) x
Amount of water penetration (m/s)
Saturation: Over the water table 0.5, under the water table
1.0

Fig. 1

Outline of the cautiously realistic conceptual model©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (4/14)

Concerns

・A conservative evaluationthat includes uncertainties
is possible
・Compliance with domestic
regulations is believed tobe higher

・Increase in load on disposal
depth and facility structure due
to excessive dose resulting from
evaluation.
(Excessive conservativeness)

Cautiously realistic conceptual model

Advantages

Concerns

・Reduction in load on
disposal depth and facility
structure by eliminatingexcessiveconservativeness・Due to uncertainties in data, settingof parameters and ensuring validityof the calculation results is difficult.
・Compliance with domestic
regulations is an issue

Even if the integral dose (area) over the entire
duration of emission is the same, the maximum
dose (peak) reduces when there is a delay inemission time.

Conservative model→t
Cautiously realistic conceptual model

Dose →

Advantages

Diffusion control→t

Dose →

Conservative model

Dose →

When dose evaluation is conducted, while a conservative evaluation that includes uncertainties can be conducted if aconservative set of parameters and a conservative model are used, the resulting dose is likely to be excessive. (Likely to havean impact on the required disposal depth and concept) (Refer to Fig. 1)・Hence, the evaluation was carried out using both, the conservative model and the cautiously realistic model, and the impact
due to the differences in the models was compared.

Diffusion control+ Barrier degradation over time→tFig. 1 Impact of the difference in evaluation models on the
results of dose evaluation (Image)©International Research Institute for Nuclear Decommissioning312

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Implementation of dose evaluation (5/14)
・Conservative model and cautiously realistic model were arranged respectively for each component (lowpermeability layer, waste bodies, external wall of the facility, etc.) of the disposal facilities conducting dose
evaluation.
・Dose evaluation was carried out for multiple cases in accordance with the combination of models for eachcomponent (barrier).
Dose evaluation cases (Example of L2 disposal facility) (Refer to Fig. 1)
・Setting all components in the conservative model・Setting the low permeability layer in the cautiously realistic model and all the other components in theconservative model・Setting the waste layer in the cautiously realistic model and all the other components in the conservative model・Setting all components in the cautiously realistic conceptual model
Etc.
Soil covering

Ground waterlevel(Upper part) Low
permeability layer

Filler mortar

Waste body

(Porous concrete)
BedrockFig. 1(Internal partition)

Concrete pit(Outer wall of facility)

Example of L2 disposal facility©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (6/14)
・The difference in the conservative model and cautiously realistic model in the case of the low permeability layer is shown in Fig. 1.
・The changes in infiltration ratio due to ageing degradation of the low permeability layer are taken into consideration in the cautiously
realistic conceptual model.

Conservative model・Condition with large amount of water
penetration (0.05 m/yr) isconservatively set throughout the entire
duration.

Infiltration ratio

Conservativemodel

Cautiously realistic
conceptual modelFig. 1

Temporal changes in the infiltration ratio of
the low permeability layer

Cautiously realistic conceptual model・Changes in amount of water
penetration due to ageing degradation
are assumed
・Initially a condition in which theamount of water penetration is thelowest at 1.6E-5 m/yr(=1E-11 m/s) is
set.
・Assuming ageing degradation over a
period of 5000 years, it is set at 0.05
m/yr for after 5000 years.

(The value used forthe cautiously realistic
conceptual model thistime is set based on
the UK LLWR safetycases)©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (7/14)
・The difference in the conservative model and cautiously realistic model in the case of waste bodies is shown
below.
・Ageing degradation of the container and delay in diffusion of each nuclide in the waste body are incorporated in
the cautiously realistic conceptual model.
(Refer to Fig. 1 and Fig. 2)

Conservative model(Simple flash emission model)
Instant mixing ofwater and waste
Advection ofdissolved nuclides

Cautiously realistic conceptual model(Time dependent emission model)
Delay in diffusionof radioactive
nuclides in thewaste bodies

Diffusiveness of
dissolved radioactive
nuclides in thecracksRadioactive nuclide aqueous
distribution consideringsolubility/sorption

Fig. 1 Comparison of the condition of waste bodies in the conservative modeland cautiously realistic conceptual model・Instant degradation of container
・Instant mixing of water and waste
・Rapid emission of radioactive nuclides
・Continuous and constant speed process

・Ageing degradation of container
・Consideration of cracks in cement
・Sequential emission of radioactive nuclides
・Time dependent process

Fig. 2 Characteristics of each model©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (8/14)
・The impact of differences in the models (conservative model / cautiously realistic conceptual model) on the dose evaluation
results is indicated below.
・Decline in dose was seen at Slurry-1 at L2 when time dependency of the low permeability layer was considered. (Table 1)
Table 1Results of dose evaluation with L2 as the disposal concept proposal and Slurry-1 as the target waste (iron co-precipitation)

Conservative model
Considering the time dependency of the low permeability layer

Impact of using the cautiously realistic conceptual model for the low
permeability layer(In the case of the scenario in which river is used)

・Decline in maximum dose is seen
・The time period when the dose is maximum is delayed about 1700
years

Fig. 1 Temporal changes in the infiltration
ratio of the low permeability layer

Physical delay has a major impact(The period of 5000 years by when degradation of the low permeability
layer is assumed to happen, is equivalent to the half-life of the main
nuclide C-14 (Fig. 1)©International Research Institute for Nuclear Decommissioning

316

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Implementation of dose evaluation (9/14)
・As a result of dose evaluation, the highest dose reduction effect was seen when the cautiously realistic conceptual model
was used for the waste bodies and backfill.
→ The evaluation of waste body performance with respect to Slurry-1 (iron co-precipitation) is likely to have a major impact on
the disposal safety evaluation.
(Further, in the case in Fig. 1, the analysis was conducted using the same value as Kd (conservative value, refer to P.284))
Cautiously realistic conceptual
model(Pb: Waste bodies, backfill)

Dose rate

Dose rate

Conservative model (Pb Ref)

Cautiously realistic conceptualmodel(Pb5: Low permeability layer )Dose rate

Dose rate

Cautiously realistic conceptual
model(Pb6: Concrete pit)

Table 1Results of dose evaluation with L2 as the disposal concept proposal and Slurry-1
as the target waste (iron co-precipitation)©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (10/14)

All barriers
Waste body
Backfill
Naturalbarrier

H-3

C-14

Sr-90

I-129 Cs-137 Pu-240

Cautiously realistic conceptual model(Pb5: Low permeability layer)

All barriers
Waste body
Backfill
Concrete
Bentonite
Naturalbarrier

H-3

C-14 Sr-90
Table 1

Ratio of inventory decayed in each barrier

Conservative model (Pb Ref)

Ratio of inventory decayed in each barrier

Ratio of inventory decayed in each barrier

Ratio of inventory decayed in each barrier・Confinement of C-14 and Pu-240 into the waste body is seen when diffusion inside the waste body is considered.
・Confinement of I-129 into the backfill is seen due to the effect of delayed diffusion.
(Further, in the case in Fig. 1, the analysis was conducted using the same value as Kd (conservative value, refer to P.284))
Cautiously realistic conceptual model (Pb2: Waste
body and backfill)

All barriers
Waste body
Backfill
Naturalbarrier

H-3

C-14

Sr-90

I-129

Cs-137 Pu-240

Cautiously realistic conceptual model (Pb6:
Concrete pit)

All barriers
Waste body
Backfill
Concrete
Naturalbarrier

H-3 C-14 Sr-90 I-129 Cs-137 Pu-240
I-129 Cs-137 Pu-240
Results of dose evaluation with L2 as the disposal concept proposal and Slurry-1
as the target waste (iron co-precipitation)©International Research Institute for Nuclear Decommissioning

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Implementation of dose evaluation (11/14)
・It was projected that the disposal concept proposal L3 would be suitable for “Rubble 1” even when the evaluation was carriedout using the conservative model. (Refer to Table 1)
・In the case of “Waste inside the building (Concrete) (Non-radioactive waste before the accident)”, the maximum dose washigh when the conservative model was used. Hence it is believed that dose evaluation needs to be carried out using a model inwhich ageing degradation of the cap, etc. is appropriately evaluated. (Refer to Fig. 1)
・Results can be said to be the same as The investigation of preliminary proposals for the disposal concepts
Table 1Cap
(Upper barrier layer)

Low permeability layer

Earth's surface

Soil covering (Low permeation of water as compared tosurrounding ares)(Or water shielding structure as shown in the above figure)

Waste layer

Evaluation point ①
Evaluation point ②

Summary of dose evaluation with respect to the disposal concept
proposal L3
Scenario in which river is used

Waste

Case

Model conditions

Maximumdose(μSv/y)Time period
when the doseis the maximum
(y)

Main
nuclidesDrainage layer
Leak-isolation
sheetsWaste inside thebuilding(Concrete)(Non-radioactive
waste before theaccident)

T_Ref

Conservative model (Without cap)

7.37E+02

128

Sr-90

T_2

Conservative model (With cap)

3.74E+01

135

Sr-90

T_6

Cautiously realistic conceptual
model(Considering ageing degradationof the cap)

3.95E+00

347

C-14

Rubble 1

T_Ref

Conservative model (Without cap)

1.50E+00

43.6

C-14

T_2

Conservative model (With cap)

1.61E-01

46.4

C-14

T_6

Cautiously realistic conceptual
model(Considering ageing degradationof the cap)

1.99E-02

349

C-14

(Bottom barrier layer)

Flow of groundwater
Evaluation point ③ After migratingfrom the natural barrier

Fig. 1
Disposalconcept proposal L3

L3 is tough as it is close
to the dose criteria©International Research Institute for Nuclear Decommissioning

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(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of dose evaluation (12/14)
・It was projected that the disposal concept proposal L2 would be suitable for “AREVA” and “Rubble 2” even when the evaluation was
carried out using the conservative model. (Refer to Table 1)・As “Rubble 2” is close to the dose criteria in the Ref case, it is believed that the amount of water penetration and the leaching at thewaste body layer need to be controlled.
・The maximum dose is high when the conservative model is used for the “Waste inside the building (Metal) (L3 before the accident)” and“Slurry 1_iron co-precipitation”. L2 is difficult to accomplish even in the preliminary proposal evaluation. Hence an L1 concept that
includes barriers is necessary.
・In order to dispose these at L2, it is believed that dose evaluation needs to be carried out using a model that appropriately evaluatesageing degradation and diffusion, etc. of the waste bodies and backfill.
Table 1
Summary of dose evaluation with respect to the disposal concept proposal L2
Scenario in which river is used
Waste

Soil covering

Filler mortar

Ground water level

AREVA*

Case

Bedrock(Internal
partition)

Concrete pit

Main nuclides

3.00E-02

108

C-14

Pa_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

2.48E-04

97000

Cs-135

Waste inside the
Pa_Refbuilding(Metal)(L3 before the accident) Pa_2

Conservative model

6.20E+01

119

C-14

Cautiously realistic conceptual
model(Waste bodies / Backfill)

4.38E-02

6630

C-14

Slurry-1
Iron co-precipitation*

Pa_RefConservative model

1.27E+01

118

C-14

Pa_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

9.21E-03

6220

C-14

Pa_RefConservative model

5.42E+00

116

C-14

Pa_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

6.62E-03

2350

I-129

Evaluation point ③ After migrating from the natural barrier

Fig. 1 Disposal concept proposal L2

Time period when
the dose is the
maximum (y)Conservative model

Evaluation point ①(Porous concrete)

Maximumdose(μSv/y)Pa_RefLow permeability layer (Bentonite)

Waste body

Model conditions

Rubble 2*: Waste for which disposal concepts for disposal deeper than L1 are used in the nearing scenario©International Research Institute for Nuclear Decommissioning320

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of dose evaluation (13/14)
・It was projected that the disposal concept
proposal L1 which includes engineered
barriers can be used for “SARRY”, “Slurry2” and “PCV waste (metal)(L2 before theaccident)” even when the evaluation was
carried out using the conservative model.
(Refer to Table 1)
・The maximum dose in “KURION”,“Rubble-5” and “RPV waste (L1 before theaccident)” is high when penetration is notcontrolled. Hence it is believed that dose
evaluation needs to be carried out using amodel that appropriately evaluates ageing
degradation and diffusion, etc. of the low
permeability layer, waste bodies andbackfill.

Table 1 Summary of dose evaluation with respect to the disposal concept proposal L1
Scenario in which river is used
Waste

KURION

SARRY

Slurry-2

Rubble 5

Case

Model conditions

Low permeability layer(Low diffusion layer)

Backfill

Waste / Filler
mortar

Backfill

Concrete pit(Low permeability layer)(Invert)Fig. 1 Disposal concept
proposal L1

Time period when
the dose is the
maximum (y)Main nuclides

I_Ref**

Conservative model

1.15E+01

1380

C-14

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

1.42E-01

2740

I-129

I_Ref**

Conservative model

9.06E+00

1380

C-14

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

1.11E-01

2750

I-129

I_Ref**

Conservative model

1.30E+00

1390

C-14

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)3.31E-03

6700

C-14

I_Ref**

Conservative model

4.31E+01

1390

C-14

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

1.88E-01

3250

I-129

Medium depth disposal
Backfill

Maximumdose(μSv/y)PCV waste(Metal)(L2 before theaccident)*

I_Ref**

Conservative model

1.73E+00

5760

Np-237

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)

8.70E-02

180000Ra-226***

RPV waste(L1 before theaccident)*

I_Ref**

Conservative model

5.40E+02

1390

C-14

I_2

Cautiously realistic conceptual
model(Waste bodies / Backfill)1.32E+00

8030

C-14

*: Waste for which disposal concepts with disposal at L1 are difficult in the nearing scenario as well**: Controlling the amount of water penetration through the low permeability layer is not considered in I＿Ref***: Ra-226 has been formed through migration of U. Hence the contribution of Rn-222 is not considered.

©International Research Institute for Nuclear Decommissioning321(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Implementation of dose evaluation (14/14)

Summary of dose evaluation・Based on the results of dose evaluation of the disposal concepts that were determined throughsensitivity evaluation, it was projected that the disposal concepts were generally satisfactory.
Further, the barrier compositions and parameters for some of the waste in which the dose criteria
was exceeded, need to be carefully decided if overlap with other waste is considered.
・Dose evaluation was carried out using both, the conservative model and the cautiously realistic
conceptual model, considering a number of barrier compositions.
As a result, it became clear that by taking into consideration the decline in amount of water
penetration due to the low permeability layer, diffusion control, and ageing degradation of barriermaterial, the maximum dose can be reduced. In the evaluation using parameters that are based on
the cautiously realistic model that takes diffusion in the waste layer into consideration and the safetycases from UK, a comparatively large decline in dose similar to the decline in leaching rate wasseen.
・Since it became clear when the cautiously realistic conceptual model was used that creating models
for the waste body layer and considering changes due to ageing can be expected to contribute todose reduction, developing parameters and incorporating them in the future is presumable.
・In the waste stream that was studied, even though determining the disposal category differeddepending on the waste, C-14 was dominant most of the times in the groundwater scenario and I129 followed.
©International Research Institute for Nuclear Decommissioning

322

(b) Provision of disposal methods and development of safety assessment methods[1] Information organization to investigate the disposal concepts and to establish safety assessment methods

Summary
Results so far✓ Parameters concerning dose evaluation were developed based on the outcomes of the studied conducted by IRID
until FY2020 (characterization, treatment, etc.) and the latest domestic disposal trends.

✓ Sensitivity evaluation of the barrier composition and barrier parameters was conducted for the disposal depths ofeach disposal method, namely, trench disposal, pit disposal, and medium depth disposal, and the parameters thatwere sensitive were identified for each nuclide.
✓ The method for deriving the disposal category for each waste based on the dose obtained through sensitivity
evaluation was studied. Methods for deriving the barrier composition and required performance of the barrier using
the results of sensitivity analysis were proposed, and preliminary proposals of disposal concepts for each waste
were consolidated.
✓ Based on the results of dose evaluation of the preliminary proposals of disposal concepts, it was projected that thedisposal concepts determined through sensitivity evaluation would generally meet the dose criteria.

Issues, etc.
✓

Since it became clear when the cautiously realistic conceptual model was used that creating models for the waste
body layer and considering changes due to ageing can be expected to contribute to dose reduction, developing
parameters and incorporating them in the future is presumable.

✓

In the waste stream that was studied, even though determining the disposal category differed depending on thewaste, C-14 was dominant most of the times in the groundwater scenario and I-129 followed.

©International Research Institute for Nuclear Decommissioning323

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal◼ Achievements until the previous year
 Nuclide migration parameters, barrier material, and affecting substances (6substances) that have a high priority in impact assessment were selectedbased on the investigation of affecting substances likely to be contained inwaste or the assumed disposal concepts.
 The idea behind the method of evaluating the impact on nuclide sorption
parameters depending on the affecting processes of the affecting
substances, the level of understanding the processes and the amount of dataon the processes, was developed. (Fig. 1)
 Trial runs and derivations were carried out concerning the sorption reduction
effect due to direct impact of the 6 substances, and sorption reduction effectincluding their indirect impact under some of the conditions that have acomparatively large amount of information contributing to the investigation.
(Table 1)
 Issues✓ Lack of findings on change in solid phase of borate, ferrocyanide, etc. inparticular✓ Need for a selective response while evaluating the extent of contribution to
exposure dose in the case of the disposal concepts being investigated◼ Goal To organize the set of parameters to be applied to dose evaluationconsidering direct as well as indirect impact and the idea behind them, basedon the findings so far and data collected on cases pertaining to which
sufficient findings are not available.
◼ Implementation items
a. Expansion of findings and data that are insufficient for evaluating direct andindirect impact of nuclide migration, and quantitative evaluation of the impact.
b. investigation of the impact of the sorption effect of the affecting substanceson dose evaluation.

Method①: Sets sorption reduction factors
手法①：溶解度上昇等の影響
based
on impact data such as rise in
データからの収着低減係数の設
solubility
定Effective in circumstances wherein semi溶解度や錯形成に関する半定量
quantitative and limited information can be
的かつ限定的な情報が得られる
obtained related to solubility and complex
状況において有効
formation
Method②: Sets sorption reduction factors
手法②：熱力学データによる錯
based
on quantitative information on
形成の定量情報に基づく収着低
complexformation based on
減係数設定
thermodynamicdata
錯形成の定量的な評価が可能な
Effectivein circumstances wherein complex
状況において有効
formation can be evaluated quantitatively

手法③：影響物質共存下での収Method③: Sets sorption reduction factors
着データに基づく収着低減係数
based
on data on sorption in the presenceofの設定
affecting substancesThe
most reliable assessment technique.
最も信頼性の高い評価手法であThegoal is to eventually use this method
り、最終的にはこの手法での設
for
setting the factors.
定が目標

Type and amount of affecting substances such as
廃棄物成分や共存物質等の影響物質の種類と量
waste components and coexisting substances(有機物,ホウ酸,塩分,セメント,…等)(Organic
matter, boric acid, salt, cement, etc.)
A) Direct impact assessment
A) 直接影響の評価手法：
method:B) Indirect impact assessmentB) 間接影響の評価手法：
method:Thresholdfor impact
影響が生じる閾値
+
Sorption＋
reduction
収着低減係数
factor[Method ①, ②, ③]【手法①,②,③】

Impact assessment method
塩濃度等の環境条件
forpH,
environmental
conditionsの影響評価手法
such as pH, salt
concentration,etc.
【条件変換手法，[Condition
conversion method
収着モデル】and sorption model]

＋不確実性
KKdd+
Uncertainty

KdKd+＋不確実性
Uncertainty

Reference conditions
レファレンス条件(Barrier
material, pore
(バリア材,間隙水化学)
water chemistry)

Directimpact such as rise in solubility and Changes
in the chemical conditions of the
影響物質成分と核種との錯形成
影響物質によるTHMC変化（特にreduction in sorption due to complex
barrier material and pore water due to等による溶解度上昇，収着低減化学）によるバリア材や間隙水
formation between affecting substance
THMC changes (chemical in particular)
等の直接的な影響の化学的条件の変化
components and nuclides
caused by affecting substances.

レファレンスK
Reference Kd +dUncertainty
＋不確実性

A)
+ B)直接＋間接影響：
Direct + Indirect Impacts:
A+B)
Consideration of both impacts by combining assessments A
AとBの評価の組合せによる双方の影響の考慮
andB

Fig. 1 Overview of methods for assessing the impacts of affecting
substances on nuclide sorption
Table 1
Organization of sorption reduction factors
for barrier materials against affecting substances
Element group

Representat
ive element

Alkali metal
Alkali earth metal
Divalent transition metal
Tetravalent transition metal
Pentavalent transition metal
Trivalent actinide
Tetravalent actinide
Pentavalent actinide
Hexavalent actinide
HalogenAnionic species

Cs
Sr
Ni
Sn
Nb
Am
Th
Np
U
I
Se

Concentration for the onset of impact (threshold value) and the sorption reduction factors (SRF)
Organic matter (ISA)
Sea-water
Boric acid
Ferrocyanide
Threshold value
SRF
SRF*3
Threshold value
SRF
Threshold value
SRF
ー
1
2
ー
1
ー
1
-2 *1
10
8
ー
1
ー
1
1×10 M
ー
1
1
ー
1
ー
1
100
10
ー
1
Issues exist
Issues exist
1×10-4M
100
10Issues exist
Issues exist
Issues exist
Issues exist
1×10-4M
10
10
ー
1
ー
1
1×10-4M-4100
10Issues exist
Issues exist
Issues exist
Issues exist
1×10 M
-4
-3
10
10
ー
1
3
1×10 M
1×10 M
1010500
3
5×10-4M
1×10-2M
1×10-3M
*2
*2
-4
1
1.3
ー
1
ー
ー
1×10 M*2*2
1
Issues exist
Issues exist
Issues exist
Issues exist
ーー(Sorption reduction factor to cement caused by affecting substances (direct impact))◼ Indicators for the achievements of goala. Expansion of data on the combination of impact affecting substances andpH
nuclides, which is required for impact assessment, and presentation ofState IState II
State IIIa
State IIIb
impact assessment techniques
Affectingb. Results of analyzing the impact of various types of parameters on dose
Absent
0.1
0.03
0.1
0.1
evaluation including the impact of decline in the nuclide sorption parameters,
ISA
0.021M
0.01
1.6×10-3
0.01
0.01
and identification of important parameters⇒ b. Will be implemented as part of “[1] Information organization to investigation
the disposal concepts and to establish safety assessment methods”.

Boric acid - 0.4M
Boric acid - 0.04M
Sea-water - 0.6M
Sea-water - 6M

0.025
0.1
0.017
1.7×10-3

7.5×10-3
0.03
5×10-2
5×10-4

0.025
0.1
0.017
1.7×10-3

0.025
0.1
0.017
1.7×10-3(Sorption distribution factor to cement caused by the affecting substance Sr (including indirect impact))©International Research Institute for Nuclear Decommissioning324

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Flow of investigation (In the case of direct impact assessment) Table 1 List of affecting substances(Example of clarification of relation of
impact)

影響物質

陽イオン

NaOH
Ca(OH)2

Na+,Ca2+

Alkali metal
Alkali earth metal
Divalent transition
metal
Tetravalenttransition metal
Pentavalenttransition metal
Trivalent actinide

陰イオン

OH-

固相表面収着サイト状態変化

核種スペシエーション
変化
核種-OH, -CO3錯体
濃度変化

液相pH変化
固相溶解

陽イオン交換
容量変化

二次鉱物生成固相表面収着エッジ量変化②陽イオン交換サイト量の変化による分配係数の減少もし
くは増加

Fig. 1 Example of clarification of relationshipbetween information on affecting processes
and nuclide interaction④陰イオン交換サイトへの収着競合による分配係数低減

液相陰イオン
の増加

Investigation and organization of information on
target substances, and clarification of relationships
among processes, mechanisms and interactions
that affect behavior such as sorption
核種-OH, -CO3錯体
濃度変化

液相pH変化

固相溶解

陰イオン交換
容量変化

二次鉱物生成固相表面収着エッジ量変化⑤陰イオン交換サイト量の変化による分配係数の減少もし
くは増加⑥表面収着エッジ(表面錯体)
への収着量の変化による分配係数の低減もしくは増加⑦核種-影響物質の錯体形成に伴う、液相中核種-影響物質
錯体の濃度上昇による分配係数の低減(+核種-影響物質
錯体の固相への収着がある
場合は低減量が減少)

核種-影響物質の錯体濃度変化

中性分子

収着挙動への影響①陽イオン交換サイトへの収着競合による分配係数低減③表面収着エッジ(表面錯体)
への収着量の変化による分配係数の低減もしくは増加e-

液相中性イオン
組成変化

核種-影響物質の錯体濃度変化

液相酸化還元電位変化

核種原子価・スペ
シエーション変化⑧液相中の核種化学形の変化により、分配係数が低減もしくは増加➢ Comprehensive investigation of information
on clarified affecting substances andclarification of their general relationships

RepreseSorption reduction factor (SRF)
ntative Substan Substan Substan Substan Substan Substan
element
ce A
ce B
ce C
ce D
ce E
ce F
Cs
1
2
2
1
2
46
Sr

1

7

3

10

8

31

Ni(II)

1

10

36

1

12

1

12

109

Tetravalent actinide

Presentation of impact
Zr(IV)
11010103
assessment
dataondisposal-affecting
Nb(V)
1
10
10
1
substancesthat 10contribute
Am(III)1
10102
to1 safety
Th(IV)
10 assessment
10
106

12

1

Pentavalent actinide

Np(V)

1

100

10

102

12

1

Hexavalent actinide

U(VI)

1

100

10

102

12

1

Halogen

I

–

–

–

1

–

–

Anionic species

Se

–

–

–

–

–

–

12

10

12

1

Sample chart of organization of sorption reduction factors for barrier materials against affecting factors➢ Acquisition of sorption impact data, including mainly data on substances that isunavailable or insufficient➢ Investigation and evaluation of evaluation methods appropriate to the volume of data

1.E+06
6

Investigation of evaluation

Complementation of
insufficient data on
SRF
substancesand nuclides
DagradationIIand investigationofevaluation methods basedon the complemented data

10 methods for substances and

SRF

nuclides that cannot be
10 4
evaluated using data alone
手法①
10 3
3
1.E+03

1.E+0510 5

1.E+04
10 4

10

手法②

1.E+02
10 2

Kd (m3/kg)

Element group

液相陽イオン
の増加

間隙水へ
の溶解

Identification ofsubstances forinvestigation➢ Target substances: 6 substances identified(Organic matter, seawater component, boric acid,ferrocyanide, sulfate, and carbonate)

核種の収着への影響メカニズム

液相・固相の状態変化

SRF
Factors)
SRF (Sorption Reduction

共存物質/廃棄体影響化処理追加成分
ホウ酸水
・ホウ酸イオンは核種と錯体を形成し、溶解度及び分配係数に影響を及ぼす可能性がある。
・ホウ酸塩として存在する場合、そのカウンターイオン放出により間隙水イオン強度が増加する可能性がある。
海水成分
・海水成分の陽イオンは、核種の収着において競合する可能性があるほか、イオ
ン強度増加により溶解度等に影響を及ぼす可能性がある。
・海水成分の溶出によって間隙水のイオン強度が上昇し、溶解度等に影響を及ぼす可能性がある。
・海水成分によって、セメント中での二次鉱物生成や、ベントナイトのイライト化が生じる可能性がある。
油分・核種によっては、油分の有機相に核種が取り込まれ、油分とともに移行する可能性がある。
・油分の分解により放出される酸によって、コンクリートの劣化が生じる可能性がある。
シリカ系
・無機系の飛散防止剤（ケイ酸塩）が溶出してシリカを放出した場合、ベントナイト（無機系飛散防止 のセメンテーションやセメント中C-S-H相形成が生じる可能性がある。
剤）
有機物
・有機物により、その分解生成物と核種での錯形成が生じる可能性がある。
（草木、有機系飛散防止剤）
金属
・溶出した鉄イオンによって、ベントナイトの変質や鉄系コロイドの形成が生じる可能性がある。
・腐食時に周辺間隙水のEh変化や腐食生成物によるコロイド形成が生じる可能性がある。
セメント系材料
・高pH間隙水を形成し、ベントナイトの変質や、核種溶解度の変化が生じる可能性がある。
・セメント水和物起源のコロイドに核種が収着する可能性がある。

影響プロセス
影響物質の溶解

1.E+0110 1

10 2

影響発現の閾値
手法③
10 1

1.E+00
10 0

収着低減
係数

-1

10
1.E-01
1.E-0610 -6-5
1.E-0510-4
1.E-0410

1.E-030-3

10
10ISAconcentration
concentration [M] (M)ISA
Case example of impact
assessment by means of multiple 10 -1
10 -6
methods appropriate to thevolume of information
Sorption Reduction Factor of Am
to cement in a system whereorganic matter (ISA) coexists-2
1.E-02

10

10 -5

10 -4

10 -3

10 -2

10 -1

10 0

ISA concentration (M)Cases of evaluation based on existing data
Organic matter effect of Am on Kd©International Research Institute for Nuclear Decommissioning325(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Assessment of sorption impact of affecting substances = Direct impacts + Indirect impacts ◼

There are multiple methods to assess the sorption impacts of affecting
substances, and they differ depending on what impacts the affecting
substances have on the target barrier system.
Case 1. Direct impact: In this case, affecting substances have direct
impact only on the behavior of radionuclides without changing any
chemical conditions or any solid phase properties in the system. An
example of this case is a situation in which in a system where pH is
kept constant, such as cement, organic ligands impact only the
behavior of radionuclides (Curve a in Fig.1).
Case 2. Indirect impact: In this case, affecting substances have an
impact on liquid phase chemistry or solid phase properties in thesystem. An example of this case is a situation in which in addition to
direct impact due to complex formation between affecting
substances and radionuclides, sorption impact occurs indirectly as aresult of environmental transition such as changes in the pH of porewater (Curve b-c in Fig. 1).

If there is a direct impact, a sorption reduction factor (SRF) is set,
and the effect of reduction of the sorption distribution coefficient Kd
is quantified according to the concentration of coexisting substances.



If there is an indirect impact, the sorption impact is quantified on thebasis of condition conversion methods, according to the affecting
processes to be considered and the knowledge of the processes.
Indirect impacts need to be assessed in combination with direct
impacts.

Concentration of
有機配位子等
affectingsubstances
ligandの影響
suchas organic
物質濃度
ligandconcentrationcpH

Environmental
transition in cement and evaluation of nuclide sorption
セメント中の環境変遷と核種収着評価
aaChanges
in sorption
distribution coefficient (Kd) caused by the affecting
安定したpH
条件での影響物質による収着分配係
substances under stable pH conditions [Direct disposal]
数（Kd）の変化【直接処分】
b Changes in pore water pH caused by cement transition, and changes in
セメント変遷による間隙水pHの変化と，それにKbd associatedwith it
伴うKdinの変化
c Changes
Kd caused by the affecting substances under conditions
whereinthe pore water pH has changed due to cement transition [Indirectc セメント変遷による間隙水pHが変化した条件でimpact]の影響物資によるK の変化【間接影響】
dFig. 1 Overview of direct and indirect sorption impacts
Type and amount of affecting substances such as
廃棄物成分や共存物質等の影響物質の種類と量
waste components and coexisting substances（有機物，ホウ酸，塩分，セメント，・・・等）(Organic
matter, boric acid, salt, cement, etc.)

method:Basic approach to sorption impact assessment (Fig. 2)abA) Direct
impact assessment
A)
直接影響の評価手法：◼

Kd

Reference conditions
レファレンス条件(Barrier material, pore water(バリア材，間隙水化学）
chemistry)B) Indirect impact assessment

method:B) 間接影響の評価手法：

Direct impact such as rise in solubility
影響物質成分と核種との錯形成
and reduction in sorption due to等による溶解度上昇，収着低減
complex formation between affecting
等の直接的な影響
substance components and nuclides

Changes in the salt concentration and
影響物質によるTHMC変化（特pHof pore water in barrier materials due
to
THMC (especially chemical) changesに化学）によるバリア材間隙水
caused by affecting substancesのpHや塩濃度の変化

Thresholdfor impact
影響が生じる閾値
+＋
Sorption reduction
収着低減係数
factor【手法①，②，③】[Methods①, ②, ③]

Impactassessment method
pH，塩濃度等の環境条件
for environmental conditionsの影響評価手法
such as pH, salt
concentration,etc.
【条件変換手法，[Condition conversion method
収着モデル】and sorption model]+ Uncertainty
KKdd＋不確実性+ Uncertainty
KKdd＋不確実性

レファレンスK
Reference
Kd +dUncertainty＋不確実性A) + B) Direct + Indirect Impacts:
A+B) 直接＋間接影響：
Consideration of both impacts by combining
AとBの評価の組み合わせによる双方の影響の考慮
assessments
A and B

Fig. 2 Overview of methods for assessing the impacts ofaffecting substances on nuclide sorption©International Research Institute for Nuclear Decommissioning326

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Direct impact assessment: Setting of data acquisition targets and conditions for setting sorption reduction factors -◼

Elements pertaining to which the acquired data was not sufficiently
accurate (Table 1 Red frame) and elements pertaining to which some ofthe data was insufficient (Table 1 Green frame), were identified.

◼

The consistency, etc. of the factors due to which distribution coefficient thatserves as the basic data for calculating the sorption reduction factor thatwas unable to be acquired, and the assumed geological environmental
conditions was verified against the elements that were identified.

◼

The conditions such as the nuclides used, liquid to solid ratio, etc. werechanged, and test condition was set wherein the concentration of affecting
substances and the pH conditions were considered to be parameters.
(Table 2, Accuracy improvement: Red, Sufficiency of data: Green)
Table 2 List of test conditions for the elements to be retested this year

Cement based

Table 1 Elements to be re-tested this year

The relationship between test conditions from last year and the acquired
data was analyzed.

Bentonite based◼

Element group

Representat
ive element

Alkali metal
Alkali earth metal
Divalent transition metal
Tetravalent transition metal
Pentavalent transition metal
Trivalent actinide
Tetravalent actinide
Pentavalent actinide
Hexavalent actinide
HalogenAnionic species

Cs
Sr
Ni
Sn
Nb
Am
Th
Np
U
I
Se

Concentration for the onset of impact (threshold value) and the sorption reduction factors (SRF)
Organic matter (ISA)
Sea-water
Boric acid
Ferrocyanide
Threshold value
SRF
SRF*3
Threshold value
SRF
Threshold value
SRF
ー
1
2
ー
1
ー
1
-2 *1
10
8
ー
1
ー
1
1×10 M
ー
1
1
ー
1
ー
1
100
10
ー
1
Issues exist
Issues exist
1×10-4M-4100
10Issues exist
Issues exist
Issues exist
Issues exist
1×10 M
10
10
ー
1
ー
1
1×10-4M-4100
10Issues exist
Issues exist
Issues exist
Issues exist
1×10 M
-4
-3
10
10
ー
1
3
1×10 M
1×10 M
1010500
3
5×10-4M
1×10-2M
1×10-3M
*2
*2
-4
1
1.3
ー
1
ー
ー
1×10 M*2*2
1
Issuesexist
Issuesexist
Issuesexist
Issuesexist
ーー

Element group

Representative
element

Alkali metal
Alkali earth metal
Divalent transition metal
Tetravalent transition metal
Pentavalent transition metal
Trivalent actinide
Tetravalent actinide
Pentavalent actinide
Hexavalent actinide
HalogenAnionic species

Cs
Sr
Ni
Sn
Nb
Am
Th
Np
U
I
Se

Boric acid

Concentration for the onset of impact (threshold value) and the sorption reduction factors (SRF)
Boric acid
Ferrocyanide
Threshold value
SRF
Threshold value
SRF
1
1
1
1
1
Issues exist
Issues exist
-2
-3
100
2.7
1×10 M
1×10 M
2.6
1
1×10-4M-4
1.3
1
1×10 M
Issues exist
Issues exist
Issues exist
Issues exist
14
1×10-3M
1
2
1×10-3M
Issues exist
Issues exist
1-4Issues exist
Issues exist
2.0
1×10 M

Ferrocyanide

Bentonite
NaCl

CementNaCl

Bentonite
NaCl

CementNaCl

7, 9.3

Cement equilibrium w ater (Approx.
pH12.5)

4-12

Cement equilibrium w ater
(Approx. pH12.5)

Boric acid concentration
Sn , Se
Sn*1 , Nb*3 , Se*3
*1

*2

Ferrocyanide concentration*5*4

Sr , Ni , Sn*1, U*4

Ni *3, Sn *1, Nb *3, Am (Eu)*4, Se *3

The behavior of each target nuclide such as container sorption, sedimentation, etc. was verified and evaluated through blank tests, etc. that do not include a solid phase.
*1: The data collected so far was collected under conditions wherein the initial concentration was high, and the disposal environment was not sufficiently simulated. Hence data will be collected using
radionuclides under conditions wherein the initial concentration is low.
*2: Since the difference between the initial concentration and the concentration after sorption equilibrium was small, distribution coefficient was unable to be calculated. Hence data will be acquired
once again by changing the liquid to solid ratio.
*3: Since there was large amount of sorption and the concentration after sorption equilibrium was lower than the lower limit of quantification of the equipment, the distribution coefficient was unable tobe calculated. Hence, data will be acquired once again by changing the liquid to solid ratio.
*4: Since data under conditions wherein ferrocyanide that is to be compared did not coexist, was not available, the dependency of distribution coefficient on the ferrocyanide concentration was
unable to be sufficiently evaluated. Hence data will be acquired once again.
*5: Since data on the circum-neutral range, which is required for ascertaining the disposal environment is not available, it will be acquired once again.

©International Research Institute for Nuclear Decommissioning

327

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal
the Sorption- Direct impact assessment: Assessment and investigation of boric acid impactsEvaluating-◼

The distribution coefficient to cement (OPC) declined underconditions wherein the Na concentration was 0.05 mol/l and the boric
acid concentration was high (0.1 mol/l）.

◼

The distribution coefficient to bentonite declined under conditions
wherein the boric acid concentration was high and the pH was either
7 or 9.3 (Fig. 1).
As there was large amount of sorption to cement (OPC), theconcentration after the test was below the lower limit of quantification.
Hence the distribution coefficient was unable to be evaluated.

Nb

◼

Se

◼

The distribution coefficient to cement (OPC) declined underconditions wherein the boric acid concentration was high (Fig. 2).

◼

The distribution coefficient to bentonite declined under conditions
wherein boric acid coexisted and the pH was either 7 or 9.3 When the
pH was 9.3, if boric acid coexisted, the volume of sorption reduced.
Hence the distribution coefficient was unable to be calculated.

 The sorption reduction factors (SRFs) were derived for the
combination of barrier material and nuclides pertaining to which thedependency of the distribution coefficient on boric acid concentration
was verified, SRFs were newly evaluated or were updated to highly
reliable values.
Newly evaluating the SRFs (mentioned hereinafter in the format of“Element - target solid phase: SRF value”)
Se - OPC: 32, Se - Bentonite: 1.5
Updating the SRFs with highly reliable values

Sn - OPC： 1 → 2, Sn - Bentonite: 100 → 2.0

Boric acid concentration

Fig. 1 Results of the sorption test of Sn tobentonite in the presence of boric acid
Evaluating the Sorption
reduction factor (SRF) basedon the slope of the solid line

Sorption distribution
coefficientSn

Sorption distribution
coefficient

reduction factor (SRF)
based on the slope ofthe solid line

Se (Na concentration 0.05 mol/l)Se (Na concentration 0.5 mol/l)Boric acid concentration

Fig. 2 Results of the sorption test of Se to cement (OPC) in thepresence of boric acid
©International Research Institute for Nuclear Decommissioning

328

Sr

◼

The distribution coefficient to bentonite declined when ferrocyanide
concentration was high (0.1 mol/l) in all pH conditions (pH4-12).

Sn◼

The distribution coefficient to cement (OPC) declined when the
ferrocyanide concentration was high (Fig. 1).
The distribution coefficient to bentonite declined when the
ferrocyanide concentration was high and the pH was 4-10. No
dependency when pH was 12.

◼

Am
(Eu)

◼

Sorption
distribution
coefficient(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Direct impact assessment: Assessment and investigation of ferrocyanide impacts -The distribution coefficient to cement (OPC) did not appear to dependon ferrocyanide concentration (Fig. 1).
The distribution coefficient to bentonite declined due to theferrocyanide concentration when the pH was 8 (Fig. 2).

Se◼

The distribution coefficient to cement (OPC) did not appear to dependon ferrocyanide concentration.

Updating the SRFs with highly reliable values
Sr – Bentonite: 1 → 2.9, Sn - Bentonite: 2.7 → 100
U – Bentonite: 2 → 2.1

Sorption
distribution
coefficient◼

Newly evaluating the SRFs (mentioned hereinafter in the format of“Element - target solid phase: SRF value”)
Sn – Cement (OPC): 1.8, Se – Cement (OPC): 1

Ferrocyanide concentration
Fig. 1 Results of the sorption test of Eu and Sn tocement (OPC) in the presence of ferrocyanideU The sorption reduction factors (SRFs) were newly evaluated based on thetest results, or the values were updated to highly reliable values.

Evaluating the Sorption
reduction factor (SRF)
based on the slope ofthe solid line

Evaluating the Sorption
reduction factor (SRF)
based on the slope ofthe solid line

Ferrocyanide concentration
Fig. 2 Results of the sorption test of U to
Bentonite in the presence of ferrocyanide©International Research Institute for Nuclear Decommissioning329

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Direct impact assessment: Summary of the results of the trials on the threshold value for onset of
impact and the sorption reduction factors -Bentonite based

Cement based

Element group

Alkali metal
Alkali earth metal

Representa
tive
element

Concentration for the onset of impact (threshold value) and the sorption reduction factors (SRF)
Organic matter (ISA)

Sea-water

Boric acid

Ferrocyanide

Threshold value

SRF

SRF*3

Threshold value

SRF

Threshold value

SRF

Cs

－

1

2

－

1

－

1

Sr

1×10-2 M*1

10

8

－

1

－

1

Divalent transition metal

Ni

1

1

－

1

－

1

Tetravalent transition metalSn

1×10-4 M

100

10

1×10-2 M

2

1×10-2 M

1.8

Pentavalent transition metalNb

1×10-4 M

100

10Issues exist

Issues exist(During analysis)

Trivalent actinides

Am/Eu

1×10-4 M

10

10

－

1

－

(Duringanalysis)1

Tetravalent actinides

Th

1×10-4 M

100

10Issues exist

Issues exist

Issues exist

Issues exist

Np

1×10-4 M

10－1

1×10-3 M

3Hexavalent actinides

U

5×10-4 M

1010

1×10-2 M

500

1×10-3 M

3Halogen

I

－*2

1

－*2

1×10-4 M

1.3

－

1

Anionic species

Se

－*2

1

－*2

1×10-2 M

32－

1

Element group

Representa
tive
element

Pentavalent actinides

10

Concentration for the onset of impact (threshold value) and the sorption
reduction factors (SRF)
Boric acid

Ferrocyanide

Threshold value

SRF

Threshold value

SRF
1

Alkali metal

Cs

－

1

－

Alkali earth metal

Sr

－

1

1×10-2 M

2.9

Divalent transition metal

Ni

－

1

(During analysis)(Duringanalysis)

Tetravalent transition metalSn

1×10-2 M

2.2

1×10-2 M

100

Pentavalent transition metalNb

1×10-4 M

2.6

－

1

Trivalent actinides

Am/Eu

1×10-4 M

1.3

－

1

Tetravalent actinides

Th

Issues exist

Issues exist

Issues exist

Issues exist

Pentavalent actinides

Np

－

1

1×10-3 M

4Hexavalent actinides

U

－

1

1×10-4 M

2.1

HalogenI

Issues exist

Issues exist

－

1

Se

1×10-3 M

1.5

1×10-4 M

2.0

Anionic species*1 M：mol/L*2 Kd = 0 (m3 kg-1)*3 SRF between precipitation conditions andseawater conditions

Set based on past information
Set based on data obtained throughsorption tests
Red: Data that was updated based on resultsof tests conducted this year

* Findings from complex formation in thepast suggested that neither sulfate nor
carbonate had a significant impact andhence these have not been mentioned.

©International Research Institute for Nuclear Decommissioning330

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Methods of assessing indirect impact based on past disposal research

◼





Concept of evaluation methods for sorption distribution coefficients
under various environmental conditions for performance evaluation ofgeological disposal (NEA, 2005; Tachi et al., 2017)
 Increase in uncertainty associated with environmental conditions or
scenarios assumed in performance evaluation◼

If there is a change in the pore water pH or solid
phase properties of cement or bentonite due tocoexisting substances, their indirect impact needs tobe considered.
For assessment of indirect impact, the methods forsetting sorption distribution coefficients can be utilized
which were established through the past
investigations of geological disposal and areadaptable to various environmental conditions andtheir changes.
The setting methods include the following: athermodynamic sorption model that predicts changes
in conditions based on a theoretical model; acondition conversion method that provides corrections
based on the relationships between changes inenvironmental conditions and sorption distribution
coefficients; and experts’ judgments based on limited
actual measurement data or knowledge of sorption
mechanisms. The applicability of these methods
depends on the processes in which affecting
substances exert an impact on the properties ofbarrier materials or depends on the level ofunderstanding of the actual measurement data or themechanisms.

・性能評価で想定する環境条件やシナリオに係る不確実性の増加
 Increase in uncertainty associated with the expansion of the scope・条件変換で取り扱う条件範囲の拡大に伴う不確実性の増加
of conditions used in condition conversionなど
・直接的な根拠となる実験データ等の減少
direct evidence, etc.
 Decrease in experimental data that serves as

◼



Data
acquisition /experimental
データ取得/
experimental
conditions
実験条件
conditions

(exp)
KK
d d(exp)

Actual measurement data under various❑ conditions
多様な条件の実測データ
 Actual measurement data under site
❑ conditions
サイト条件での実測データ
Error in experimental data
実験データの誤差 Analysisof tendency
Condition ❑傾向性の分析
条件変換
conversion (Experts' judgments)
 （専門家による判断）
Condition conversion method Thermodynamic❑
条件変換手法 sorption❑ model
熱力学的収着モデル

Performance
PA reference
性能評価
evaluation
レファレンス条件
conditionconditions
Reference(ref)KKdd(ref)Error
in experimental data
実験データの誤差+＋
Uncertainty
associated with
条件変換に伴う不確実性
condition conversion
Condition
条件変換
conversion

Other performance evaluation
その他の性能評価条件
conditions•Changesin geological /・地質/地下水条件の変動
groundwater conditions
・変動シナリオ•Changescenarios,など
etc.

Kd (A, B, ・・・)

Error
in experimental data
実験データの誤差+ Uncertainty associated with＋ 条件変換に伴う不確実性
condition conversion
条件変動に係る不確実性+＋
Uncertainty
associated withcondition changes

Indirect impact assessment for bentonite: The condition conversion method that has been investigated during the geological disposal
evaluations conducted so far, for setting the sorption parameters considering environmental transition impact on bentonite, isapplicable.
Indirect impact assessment for cement: It is difficult to assess indirect impact of cement based materials using the method used forbentonite, because of the properties of its solid phase, the transition process, and the complexity of their relationship with theradionuclide sorption and retention mechanisms. Hence an estimation and assessment method based on the relationship between thesolid phase transition process and its impact on nuclide sorption is required.
©International Research Institute for Nuclear Decommissioning331

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Approach towards concretization of the assessment of indirect impact on cement Approach towards investigating the assessment of indirect impact on cement◼

In the past, the method that considers the gradual transition process of the pore water and mineral facies of cement, and the associatedchanges in the properties of nuclide sorption has been considered for establishing the parameters for nuclide sorption to cement. Whilefollowing the methods for setting sorption that were used so far, a method of assessing the indirect impact was presented wherein thechanges in mineral facies of cement caused by affecting substances in the accident wastes, and their impact on nuclide sorption, wereadditionally considered.

◼

In that assessment method (Fig. 1), the transitions in constituent minerals of cement are consolidated (Step ①) based on the pastevaluations of sorption in cement during radioactive waste disposal, and thereafter the constituent minerals of cement that dominantly
contribute to the sorption of nuclides being evaluated are specified (②). Then, the changes taking place in the mineral facies of cement
caused by the affecting substances are identified (③), and their impact on nuclide sorption is qualitatively estimated based on therelationship between the properties of nuclide sorption and the cement minerals specified in steps ① and ②. Further, based on theseestimation results, parameters can be conservatively set considering the indirect impacts, depending on the type and contained amount of
the affecting substances.
Past
evaluations
従来の放射性of sorption of
廃棄物処分にnuclides to
おけるセメント
cementduring中の核種収着radioactive waste
disposal評価

Assessment of
福島事故廃棄indirect
impact of
物中の影響物
affectingsubstancesin the
質がセメント中Fukushimaの核種収着にaccident
waste on
及ぼす間接影nuclide sorption
響評価
in cement①①セメントの変遷過程におけるSpecification of the changes in
the composition of minerals during
鉱物組成変化の特定
the cement transition process②
Specification of the mineral②放射性核種の収着を支配す
faciesthat dominate sorption of
る鉱物相の特定radionuclides③ Analysis of impact of affecting
substanceson the mineral composition of③影響物質によるセメントのcement④ Assessment of impact on nuclide
sorptionassociated with mineral changes④影響物質による鉱物変化に
causedby affecting substances
伴う核種収着への影響評価

セメントの水和反応等を考慮した
Consolidationof the cement mineral
セメント鉱物組成や間隙水化学
composition
considering cement
及びそれらの変遷過程を整理
hydration,
the pore water chemistry,
and their transition process

鉱物組成への影響分析

Analysis and organization of the impact ofaffectingsubstances in the Fukushima
福島事故廃棄物の影響物質によaccident
wasteson the constituent
るセメント構成鉱物やその変遷状
minerals of cement and their transition
態への影響を分析・整理
states

評価対象の放射性核種のセメント
Consolidationof mineral facies thatに対する収着を支配あるいは寄contributeto or dominate sorption to
与する鉱物相の整理
cementof radionuclides being
evaluated

Assessment of impact of changes in
影響物質によるセメント鉱物相のcementmineral facies caused by affecting
substances,on nuclide sorption, based on
変化が核種収着に及ぼす影響を
informationfrom① - ③.
①～③の情報に基づき評価

Fig. 1 Approach towards assessment of indirect impact of affecting substances on nuclide sorption in cement (Steps ① - ④)©International Research Institute for Nuclear Decommissioning

332

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Assessment of indirect impact on cement (1) Specification of the changes in the composition of constituent minerals
during the cement transition process (Step ①)◼ Cement, which is represented by OPC, passes through severaltransition states due to interactions such as long-term
hydration reaction with groundwater components, etc. (Fig. 1)
state II, pH = 12.5
state I, pH > 12.5

◼ The transition of the mineral facies of cement has been
understood through experiments and models. The relative
importance of the constituent mineral facies differs dependingon the transition states respectively. The changes in the mainmineral compositions can be organized depending on the
category of the transition state of cement as shown in Table 1(Berner, 1990; Taylor, 1997; Ochs et al., 2016).

state III,12.5 > pH > 10

state IV,
pH = incoming water

Fig. 1 Transition process (State I to IV) of the pore water of OPC (Ochs et al., 2016)
Table 1 Changes in
the composition of the Transition state (State)main constituent
Pore water pH
minerals during thePortlandite
cement (OPC)
transition process

Roughly 20 wt%
Roughly 45 wt%

AFt phase /Ettringite

Roughly 10 wt%

AFm phase
Hydrotalcite

Roughly 10 wt%

CalciteNon-hydrated clinker minerals

Roughly 2 wt%
Roughly 1 wt%
Roughly 10 wt%

Can exist in all states(Gray: Main mineral composition during the transition process)©International Research Institute for Nuclear Decommissioning333

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Assessment of indirect impact on cement (2) Specification of the mineral facies that dominates sorption of radionuclides (Step ②)◼

During the past safety evaluations of waste disposal, nuclide sorption evaluations that considered the transition in state of mineral
facies and pore water of cement, were taken into consideration (Wieland, 2014; Ochs et al., 2016). Specifically, the sorption
distribution coefficients were set on the basis of data obtained by actual measurement during the respective transitions, and by taking
the relation between the cement mineral facies and nuclide sorption properties into consideration.

◼

In order to investigate the impact of affecting substances in the accident wastes on cement mineral facies and nuclide sorption, therelationship between the cement mineral facies and the sorption properties of a variety of nuclides were investigated based on casesthat examined setting of sorption distribution coefficient to cement so far, and in addition, the relationships were consolidated asshown in Table 1.
Table 1 Relationship between the constituent minerals of cement and the sorption properties of radionuclides (extent of sorption: Sorption
present +++ > ++ > +, sorption absent -, unclear*)

Cement constituent
minerals

Alkali andalkali earth

Transition metal

Actinides and lanthanides

Anion

Portlandite

AFt phase /Ettringite
AFm phase
Hydrotalcite
Calcite©International Research Institute for Nuclear Decommissioning334

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Assessment of indirect impacts on cement (3) Assessment of impact of affecting substances on the mineral composition of cement (Step ③)◼ By investigating and analyzing the past findings related to each of the 6 affecting substances, the impacts of seawatercomponent were organized in the form of an impact flow as shown in the example (Fig. 1), and from among the indirect
impact processes to be considered, the possibility of transition of mineral facies in particular was specified.

◼ For example, as an indirect impact of seawater components on cement, in addition to the changes in the main constituent
minerals such as CHS, AFt phase (Ettringite), AFm phase, and hydrotalcite as shown in Fig. 1, there may be possibilities ofsolid phase transition such as formation of brucite due to Mg components, formation of ettringite and gypsum due to sulfuric
acid components, formation of Friedel's salt due to Cl, etc.
影響プロセス
Impactprocess
Change液相・固相の状態変化
in the state of liquid / solid phases

Dissolutionof affecting
影響物質の溶解substances
Affecting
影響物質
substance

Mechanismof impact on sorption of nuclides
核種の収着への影響メカニズム

Risein concentration of
間隙水中の影響affecting substances in
物質濃度の上昇pore water

Change in nuclide
核種スペシエーション
speciation
変化

Changein the収着サイト
state of
状態変化
sorption site

Directimpact*
直接影響※

Seawater海水成分
component

Cl-, Na+, Mg2+,
Ca2+, K+, SO42-,
HCO3-substancなど
ePossibility of change in properties of the main
cement constituent minerals
・CSH: Due to the impact of seawatercomponents, CSH phase dissolves dependingon the pore water pH and ionic strength.
・AFt phase (Ettringite) and calcite: Depending on
the seawater components, there is a possibilityof precipitation in the secondary solid phase.
・AFm phase and hydrotalcite phase are also
likely to dissolve depending on pH just like CSH.

収着挙動への影響
Impacton sorption behavior①①
Reductionin distribution coefficient due
陽/陰イオン交換サイト
to sorption competition between sorptionなどの収着サイトへの収着
sites
such as cation/anion exchange sites,
etc.
競合による分配係数低減②②
Reductionin distribution coefficient due
核種-影響物質の錯体形
to rise in the concentration of complexes
betweennuclides and affecting substances
成に伴う、液相中核種-影響
in liquid phases as a result of complex
物質錯体の濃度上昇による
formation
between nuclides and affecting
substances(The extent of the reduction is
分配係数の低減(+核種-影響
moderated if there is sorption of complexes
betweennuclides and affecting substances
物質錯体の固相への収着がto solid phases.)

Changein the concentration
核種-影響物質のof complexes between
錯体濃度変化
nuclidesand affecting
substances

ある場合は低減量が減少)

Indirect
impact*
間接影響※

Change in the
液相pH変化
liquidphase pH

Change in the concentration
核種-OH, -CO3錯
of complexes between
体濃度変化
nuclidesand OH or CO3

Reduction or increase in③③表面収着エッジへの収着
distribution coefficient due to
量の変化による分配係数のchangein the quantity of sorption tosurfacesorption edge
低減もしくは増加

Changein oxidation液相酸化還元
reduction
potential in
電位変化liquid phases

Change in nuclide
核種原子価・スペ
valence/speciation
シエーション変化

Reduction or increase in④④液相中の核種化学形の変
distribution coefficient due to化により、分配係数が低減
changein the chemical forms of
nuclidesin liquid phases
もしくは増加Solid-phase
固相溶解・
dissolution/second
ary mineral
二次鉱物生成formation

Fig. 1 Flow of impact of seawater components on
sorption distribution coefficients for cement or
bentonite (direct and indirect impact)※セメント系材料及びベントナイト系材料

Change in
陽/陰イオン交
cation/anion
換容量変化
exchange
capacity⑤陽/陰イオン交換サイト
Reduction or increase in⑤
distribution coefficient due to
量の変化による分配係数のchange in the quantity of
cation/anion
exchange sites
減少もしくは増加

Changein the quantity
固相表面収着of surface sorption
エッジ量変化edge
in solid phases

Reduction or increase in⑥⑥
表面収着エッジ量の変化
distribution coefficient due toによる分配係数の低減もしchangein the quantity of surface
sorptionedge
くは増加©International Research Institute for Nuclear Decommissioning335

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Indirect impact assessment: Assessment of indirect impacts on cement (4) Assessment of impact on nuclide sorption associated with mineral changes caused by affecting
substances (Step ④)◼

The impact of respective affecting substances on nuclide sorption through the intermediary of transition of the cement mineral facies was
qualitatively evaluated based on the impact of affecting substances investigated in Step ③ above, on cement mineral facies (dissolution /sedimentation), and the relationship between cement mineral facies and properties of nuclide sorption investigated in Steps ① and ②.

◼

For example, as an impact of seawater component, indirect impact evaluation is possible as shown in Table 1. Hence the possibility ofsorption reduction of a variety of nuclides due to CSH dissolution, increase in sorption of anions due to Aft phase (Ettringite)
sedimentation, etc. needs to be considered.

Table 1 Changes in cement constituent minerals caused by the impact of seawater component, and results ofassessment of the impact on nuclide sorption associated with it
Changes in cement transition state (State)
Cement constituent
minerals

Alkali andalkali earth

Impact on nuclide sorption
Transition metal Actinides andlanthanides

Anion

Portlandite

AFt phase /Ettringite
AFm phase
Hydrotalcite
Calcite(Legend) Gray: Main mineral composition during the transition process, Red/-： Dissolution of minerals or decline in nuclide sorption, Blue/+：
Sedimentation of minerals, increase in nuclide sorption, Green/±： Unclear impacts©International Research Institute for Nuclear Decommissioning336

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Current status of the sorption impact assessment (direct and indirect impact) considering the affecting
substances and the issues thereof ◼ The current status of assessment of sorption impact of the 6 affecting substances namely, organic matter, seawatercomponents, boric acid, ferrocyanide, sulfates and carbonates, including both direct and indirect impact, and the futureissues have been summarized below based on the results of investigations and studies conducted so far.
Affectingsubstance
Organic matter

Current assessment of sorption impacts during disposal

Future issues

・ Past findings on the impact of organic matter are comparatively substantial. It is realistic to use the SRF set values based on actual・ The direct impact of organic matter due to complex formation is highly likely to become evident

measured data on the sorption impact of isosaccharinic acid (ISA) which is believed to have the largest impact in particular.
・ The method for estimating the SRFs for ISA and EDTA based on thermodynamic data was investigated. It was believed that it
would be possible to evaluate the estimation of SRF if highly reliable thermodyamic data can be used. However, thermodynamic
data on organic matter cannot be considered sufficient.

compared to that of other affecting substances. Therefore it is necessary to investigate more
realistic impact assessment by assuming the organic matter likely to be actually contained in
accident wastes and its concentration.
・ In particular, the investigation of applicability of impact assessment methods to a variety oforganic matters, and the investigation of the systematicity evaluation methods, etc. between
organic matters is important.

Seawatercomponent• Regarding the impacts of seawater components on nuclide sorption, there is a relatively large amount of knowledge, so it ispossible to some extent to quantitatively evaluate sorption reduction (direct impact) based on the understanding of sorption
mechanisms such as sorption competition with seawater components.
・Regarding the assessment of indirect impact of high concentration seawater components, the sorption impact based on thetransition of cement solid phase can be qualitatively assessed.

Sulfates・ It was presumed based on the actual measured data obtained from this project and from the assessment based on thermodynamic
data that the direct impact of the formation of sulfate complex on sorption is not conspicuous.
・ Meanwhile, the indirect impact of transitioning of the cement mineral facies assumed to have occurred due to the impact caused by

・ The impact of seawater components is highly likely to become evident. Hence the settings need
to be revised further based on the latest findings on seawater component impact.
・ Also, a more realistic indirect impact assessment based on the more realistic analysis andevaluations of long term transition of barrier material such as cement, etc. as a result of seawatercomponents is required.

・ Regarding the solid phase transition in the presence of sulfates, in addition to understanding thattransition process, it is important to understand the impact on barrier performance other than
nuclide migration. Hence further investigations including analytical assessment are required.

the presence of sulfates, can be assessed based on the qualitative and conservative methods proposed in this project.
Carbonates

・ The impacts of carbonates can be assessed using methods that have been investigated in the past, because carbonates arecomponents originally contained in natural rock or groundwater environments.

・ The indirect impact of transitioning of the cement mineral facies assumed to have occurred due to the impact caused by the・ As mentioned in the column on the left, although carbonates have been included in the disposal
evaluation conducted so far as well, including their direct and indirect impacts, investigations

based on the status of carbonates contained in accident wastes are necessary.

presence of carbonates, can be assessed based on the qualitative and conservative methods proposed in this project.
Boric acid・ Regarding the impact of boric acid, data on sorption impact was obtained and based on those results, it can be assessed that boric

・ It is believed that the presence of high concentration boric acid will be concretized whileacid has little impact on alkali metals and alkali earth metals, whereas in the case of actinides, if the presence of carbonic acid isinvestigating waste conditioning considering the waste conditioning process and theconsidered where boric acid has an impact to some extent, that impact is not prominent.

concentration limit as boron is a substance subject to environmental standards.

・ When the waste has high concentration of boric acid, the impact of solid phase transition needs to be considered as well, but thereis lack of information on this point.
Ferrocyanide

・ Regarding ferrocyanide, data on sorption in the presence of ferrocyanide was obtained which suggested that even though it has aslight impact on some nuclides, that impact is not significant.
・ When there is high concentration of ferrocyanide, the impact of solid phase transition is likely to occur, but there is lack of
information on this point.

・ Along with the fact that total cyanide is a substance subject to environmental standards, thebehavior of ferrocyanide inside the waste body or in the disposal environment needs to beevaluated, and investigations need to be conducted including measures to be taken at the time ofthe pretreatment©International Research Institute for Nuclear Decommissioning

337

(b) Provision of disposal methods and development of safety assessment methods[2] Development of techniques for assessing impact of affecting substances, etc. on disposal- Summary Current results✓The impact processes of the impacts of affecting substances (6 main substances) that are likely to be present in accident
wastes, on nuclide sorption to the engineered barriers (cement, bentonite) which are used in common in a variety of disposal
concepts were identified.
✓The affecting substances were categorized into those that have a “direct impact” due to formation of complexes with thenuclides, and those that have an “indirect impact” by giving rise to changes in the solid phase or pore water of engineered
barriers, and the methods for evaluating the respective impacts were proposed.
✓For direct impact, a method of deriving sorption reduction factors (SRF) depending on the status of data was presented, and in
addition, by investigating existing information and expanding data that was insufficient, the SRFs and concentration at which theimpact appears (threshold) were organized respectively for the 6 main affecting substances - the engineered barriers (cement,
bentonite) - the main elements.
✓With regards to indirect impacts, the method of qualitatively assessing the impact of affecting substances on nuclide sorption asa result of changes in the cement mineral facies, was presented, whereas for bentonite, using the conventional method was
proposed.
✓The possibility that the impact of organic matter and seawater components would be relatively important in the case of direct
impacts, whereas the impact of seawater components and sulfates would be relatively important in the case of indirect impacts,
was presumed.
Future issues✓The method for evaluating direct impact of important organic matters and other affecting substances included in accident
wastes considering their diversity and concentration needs to be concretized.
✓Assessment of direct and indirect impacts considering the concentration of affecting substances in actual disposal systems,
their spread, etc. is required.
✓It is important to investigate impact assessment and identification of issues in conjunction with the analysis of nuclide migrationand the analysis of uncertainties.
✓Investigations need to be carried out including processing measures considering perspectives such as impact at the time ofwaste conditioning, substances subject to environmental standards, etc.

©International Research Institute for Nuclear Decommissioning338

2. Project details (No. 3)
c. Characterization of waste(a) Efficient characterization[1] Investigation on analytical inventory estimation methods[2] Proposal for mid-and-long-term analysis plan(b) Development of sampling technology©International Research Institute for Nuclear Decommissioning339

c. Characterization of waste
Overview of analysis and R&D with data from analysis
Drafting the plan for this year

Collection of rubbles, secondary waste generated
from contaminated water treatment, etc.

Investigation of plan
directed towards thenext milestone
Analysis planMid-and-long-term Plan

Development of analysis and planning method
Development of sampling technology

Sampling
Investigation of method of
combining Bayesian statistics
with the DQO process

Collection from cesium adsorption vessel
Development of analysis methods

Analysis

DQO: Data Objectives Quality
Information that would be
required in the future(Feedback)Analytical data
Analysis by JAEA,
NDC, and NFD
Database enhancement
Continued use of
FRAnDLi
Development of inventory estimation methods

Reduction of
uncertainties of theinventory

Estimated inventory
Investigation of disposal concepts, etc.

©International Research Institute for Nuclear Decommissioning340

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis◼ Achievements up to FY2020
 Collection, management and analysis of analysis samples requiredfor characterization of solid waste continued to be carried out.
 Rubbles, secondary waste generated from contaminated water
treatment, contaminated water, plants, and soil were analyzed byassigning the task of analysis to 5 facilities (JAEA NSRI / NCL /Oarai, NDC and NFD) located in Ibaraki prefecture.
 The obtained analyzed data is being used for decommissioning(Fig. 1) and in addition, it has been included in the database so thatit can be used by concerned parties (Fig.2).
◼ Goal To accumulate analyzed data on solid waste so as to contribute to
the development of analytical inventory estimation methods.

4①

α核種分析

試験水に対し、段階的なフィルタを設け、各フィルタでの回収物とろ液に対し分析を実施。
フィルタ径の選定にあたっては、ALPSで使用しているクロスフローフィルタが0.02μmであることか
ら本試験でも0.02μmまでを採用することとした。

試験水

Bq/L

粒径
10μmフィルタ

ろ液
1μmフィルタ

U-235

U-238

Am-241

Cm-244

Cm-242

Pu-238> 10 µm

7.2E-01

5.7E+00

1.7E+04

1.3E+04

5.6E+01

5.2E+03

1.8E+0310～1 µm<6.0E-04

1.3E-03<2.0E+00 <2.0E+00 <2.0E+00 <6.0E-01<6.0E-01

1～0.1 µm

<6.0E-04

1.7E-03<2.0E+00 <2.0E+00 <2.0E+00 <5.0E-01<6.0E-01

0.1～0.02 µm

3.0E-03

2.4E-02<1.0E+00 <2.0E+00 <2.0E+00 <6.0E-01<9.0E-01< 0.02 µm（ろ液）<9.0E-04

1.9E-03

7.7E-01<5.0E-01<6.0E-01

1.4E+00

【参考】

ろ液

Bq/L

粒径
0.1μmフィルタろ液
0.02μmフィルタ

ろ液<5.0E-01

Pu-239+240

全αCs-134

Cs-137> 10 µm

3.7E+04

1.7E+06

3.2E+07

1.7E+06

10～1 µm<2.0E+00

2.2E+04

4.4E+05<8.0E+02

1～0.1 µmCo-60

<2.0E+00 <7.0E+02

3.2E+03<5.0E+02

0.1～0.02 µm<2.0E+00

5.9E+03

1.1E+05

5.6E+02< 0.02 µm（ろ液）

2.2E+00

7.0E+07

1.4E+09

5.5E+04

Uを除くデータは廃炉・汚染水対策事業による成果

α核種の粒径として、概ね数μm以上のものと推測され、同程度のフィル
タを設置することにより、告示濃度(4Bq/L)を満足できると考える。

フィルタ設備のメッシュ径の設計に反映

5

Fig. 1 Results of analyzing the water from Unit 2 Torus room (Referenced from TEPCO) *1◼ Details of implementation
 Obtaining the cooperation of TEPCO, collecting analysis samples
and storing them inside the power station with the idea of continuing
to analyze the solid waste in the mid-and-long-term.
 Transporting the analysis samples to analytical facilities such as
JAEA, etc. located in Ibaraki prefecture, and conducting analysis
 Recording the obtained analyzed data into the database.
“Fukushima Daiichi Radwaste Analytical Data Library (FRAnDLi)”,and making it available for general use (Fig. 2) Also, investigating
databases for managing analysis samples.
◼ Indicators for the achievements of goal Continued disclosure of analytical data.
Fig. 2 Analysis database that continues to be usedhttps://frandli-db.jaea.go.jp/FRAnDLi/

341

Research
Institute for Nuclear Decommissioning
th session),*1 Secretariat of the Team for Countermeasures for Decommissioning and Contaminated Water©International
Treatment (86Document 3-1, January 28, 2021

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis plan)◼
This year, the plan for conducting analysis focusing on rubble and secondary waste generated from contaminated water treatment is being drafted. (Table 1)
 In order to reduce the uncertainties in estimating inventories, contamination mechanism and waste that contributes to accumulating statistical data will be selected andanalyzed.
Table 1 Plan for this year’s analysis and status of implementation
Collection

Location (Off-gas system)

3

Completed

Same as above

1

Completed

Location (inside the building), Material (rubbles)

8

Completed

This year

Location (Basement floor)

To bedetermined

Pending

Planned for FY2022

Contamination mechanism
Same as above

2

Completed

Received on August 19

Completed

JAEA

1

Completed

Received on August 19

Completed

JAEA

Same as above

1

Completed

Completed

Completed

NFD

Actual waste data

4

Completed

Completed

Completed

JAEA

Same as above

6

Completed

Completed

JAEA

Same as above

4

Completed

This year

First timeCompleted

NFD

Same as above

2

Completed

This year

First timeCompleted

NFD

Same as above

5

Completed

Completed

Needs to beinvestigatedAnalysis inprocess

JAEA

Chelating resin 2

Same as above

1

Completed

Completed

InvestigatedCompleted

JAEA

Chelating resin 1

Same as above

1

Completed

Completed

InvestigatedCompleted

JAEA

Silver zeolite

Same as above

2

Completed

Completed

Analysis inprocess

JAEA

Same as above

2

Completed

Completed

Needs to beinvestigatedNeeds to beinvestigatedSame as above

1

Completed

Completed

Focus points for inventory estimation

SampleThis report
Dismantling
Waste
RubbleStatusSampleNumber ofsamplesClassification
Units 1, 2 exhaust stack core

Smear solution in Unit 1 SGTS piping
Unit 1 Reactor Building, 1-3 Radwaste Building

Unit 1 Reactor Building stagnant water (containing sludge)

This report
Unit 3 Reactor Building accumulated water (containing sludge)
Water from contaminated water storage tank (containing sludge)

Location (residual water in the tank)Transport
This year

Analysis
methodFirst timeCompleted
First time

Analyzed byAnalysisCompleted

NDC

Completed

NFD

Completed

NDC

JAEA

Contamination mechanism
Unit 3 Turbine Building sludge
Secondarywaste
generatedfromcontaminatedwater
treatment

Iron co-precipitation slurry

Interim Report

Carbonate precipitation slurry
Carbonate precipitation slurry
Concentrated liquid waste slurry
Cerium oxide

Titanium oxide

This report

This report

Ferrocyanide
Titanate

Same as above

2

Completed

Titanate 2Same as above

3

Completed

Silver zeolite

Same as above

1

Completed

Same as above

2

Completed

Same as above

To bedetermined

Completed

Same as above

4

Completed

Same as above

1

Completed

Cerium oxide
Activated carbon
Sub-drain adsorbent
Activated carbon sandbag

Interim ReportCompleted

VerifiedCompleted

JAEA

InvestigatedCompleted

JAEA

Completed

InvestigatedAnalysis inprocess

JAEA

This year
Secondtime
This year
Secondtime
This year
Secondtime

Needs to beinvestigatedNeeds to beinvestigatedNeeds to beinvestigated

JAEA
JAEA
JAEA

To be determined
This year

First timeCompleted

JAEA
Completed

NDC

Completed

JAEA©International Research Institute for Nuclear Decommissioning342

(a) Efficient characterization ① Investigation on analytical inventory estimation
methodsi. Implementation of analysis (Analysis plan)◼
◼

Technology is being developed for collecting adsorbent from the used cesium adsorption vessel ((b) Development of sampling technology).
The dose rate of the adsorbent is extremely high and is expected to reach the Sv order every hour with about 10 cm3 of adsorbent . Hence,
considering safety, specific procedures right up to analysis need to be investigated.
◼ Impacts to be considered at each stage of storage and transportation were identified and technical constraints were investigated (Table 1)
Further, the handling method was investigated considering 2 facilities of the JAEA as examples of analysis facilities.
 Since the operation of chemical separation is carried out in the glove box or the hood where contamination is controlled, radioactive Cs
needs to be removed or fractionated in advance in the cell. Investigations continue to be carried out.

Table 1 Impact of the properties of the sample on the handling method
Collection
Process

Mass*(g)Dimensions (cm)Collection• Measurementwork• Transportation ofheavy weight
objects• Dimensions ofsampling head

• Exposure during
work such as
sampling, etc.

Storage

Storage• Transportation ofheavy weight
objects• Dimensions of
the canister• Storage
location• Conformancewith the control
value• Exposure during
handling work

Transport

Transport• Specifications oftransport cask
• Transportation ofheavy weight
objects• Type oftransport cask

• Transportation ofheavy weight
objects• Dimensions of
the service
entrance

Analysis

Radioactivity (Bq)

KURIONSARRY

Dose rate (mSv/h)• A2 value• Formation ofhydrogen• Conformancewith theregulatory value• Exposure during
work such asmeasurement,
packing, etc.

• Quantity
permitted• Limit of fixed
quantity• Conformancewith the control
value• Exposure during
work

Spare solid material storage

TOSS typetransport cask

PADIRACtransport cask

Tokai

Okuma center(NCL CPF)(Building 1)

AnalysisCA-5 cell*

Physicalproperties
evaluation cell

Glove box
hood

Iron cell

Glove box
hood

*Cell name “CA-5 cell”

Fig. 1 Flow leading to the analysis of the cesium adsorbent (being investigated)©International Research Institute for Nuclear Decommissioning343

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Sampling)◼

The analysis samples of solid waste were obtained in coordination with the construction work and the operation of the processing facility. A
plan for collecting around 75 rubble samples from Units 1 to 4 and waste storage locations was created while referring to plans such as
those concerning construction work, etc. 90 samples were collected from buildings in Units 1, 2 and 3, Unit 3 southern yard, common pool,temporary storage area C for rubble, and flange tank dismantling facility. (Table 1)◼ Also, the process water was collected from the contaminated water treatment facility. A total of 18 samples were collected from between theadsorption vessels in Sarry II over the course of 3 sampling rounds (August 27, 2021, November 20, 2021 and February 22, 2022) with 6
samples collected during each round by means of the sampling line for operation management.
 Sampling of the ALPS adsorbent had been planned. However, it was unable to be pursued as the equipment were in operation.

Table 1 Breakdown of the number of samples of rubble, etc. collected
Sampling locations
Number of samples collected
Unit, etc.

Rubble andother debris

Smear

Total

First floor

-

3

3

Third floor

2

1

3

T/B

First floor

1

1

2

Rw/B

First floor

1

1

2

R/B

First floor

-

2

2

Second floor

-

2

2

First floor

3

4

7

Second floor

3

3

6

Rw/B

Basement
Level 1

1

1

2

R/B

First floor

5

11

16

Second floor

-

7

7

Unit 3 southern yard (Fig. 1)

6

-

6

Common pool

8

-

8

Temporary storage area C for rubble (Fig. 2)

20

-

20

Flange tank dismantling facility

4

-

4

Unit 1

Unit 2

Building
R/B

T/B

Unit 3

Floor

Fig. 1 Unit 3 Southern yard
(Numbers indicate the dose rate (mSv/h),
Red dot indicates the sampling location)

Fig. 2 Crushed rubble at temporary storage area C©International Research Institute for Nuclear Decommissioning344

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Sampling)◼ Incinerated ash generated during the operation of the miscellaneous solid waste incineration facility is filled in drums and stored
in the temporary storage facility.
 There are 2 incineration systems (System A and B), and both systems are identical.
◼ A total of 25 samples were collected form the incinerated ash generated by both the systems (Table 1). The samples werecollected after the ash was filled in the drums before closing the lid. (Fig. 1)◼ The dose rate of the samples showed a tendency to have a correlation with the surface dose rate of the drum. It does not form astraight line perhaps due to the background (Fig. 2).

Table 1 Breakdown of collection of incinerated ash samples
SystemSystem A
System B

SampleA
B

Number ofsamples collected13
12

Sampling location (Temporary storage chamber) The incinerated ash drum and the collected samples

Fig. 1 Collection of incinerated ash samples

Drum

Fig. 2 Dose rate of the collected incinerated ash samples (mSv/h)©International Research Institute for Nuclear Decommissioning345

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Storage management and transportation of samples)◼ The collected and obtained samples are stored in the solid waste storage facility (Fig. 1) and continue to be managed as before.
◼ The analysis samples were transported to the analysis facilities, namely JAEA (Nuclear Fuel Cycle Engineering Laboratories),
NDC, and NFD.
 Samples were transported in 2 batches with the first batch sent on September 16. The second batch was sent on February 17.
(Table 1)
 And, the Unit 3 torus room stagnant water reached the Nuclear Fuel Cycle Engineering Laboratories (NCL) in August.
Table 1 Transported analyzed samples
Classification

Number ofsamples

Batch

Destination

Demolition waste

Units 1, 2 exhaust stack
core

3

First

NDC

RubbleUnit 1 Reactor Building,
Unit 2 Turbine Building,
Units 1 - 3 Radwastebuilding

11

First

NDC

Carbonate precipitation
slurry

4

First

NFD

Concentrated liquid waste
slurry

2

First

NFD

Sub-drain adsorbent

4

First

NDC

Titanate 2 (FST)

3

Second

JAEA, NCL

Silver zeolite

1Second

JAEA, NCL

Cerium oxide

2Second

JAEA, NCL

Secondary waste
generated fromcontaminatedwater treatment

Fig. 1 Status of sample storage at the solid waste storage facility(Top: Samples in the back side of the spare room, Bottom: Soilcore from the control room)

Sample©International Research Institute for Nuclear Decommissioning346

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Rubbles)◼
◼

With regards to the properties of the rubble and demolition waste, the contamination of the exhaust stacks in Units 1 and 2 was analyzed.
Approximately 60 m of the top portion of the exhaust stacks in Units 1 & 2 were dismantled. Core samples from the exhaust stack wereobtained, and the inner surface was analyzed (Fig. 1).
◼ 90Sr, 125Sb, 134, 137Cs, 238U were detected. Actinide nuclides other than 3H, 14C, 60Co, 63Ni, 79Se, 99Tc, 106Ru, 129I, 152, 154Eu and 238U were notdetected (Fig. 2).
◼ When the transport ratio (Cs-137 as reference) of the detected nuclides was calculated along with the analytical value of the smear solution
obtained by wiping off the inside of the Standby Gas Treatment System (SGTS) pipes in Units 1 and 2 (analyzed this year), and thecontaminated water*2 from the exhaust stack drain sump pit in Units 1 and 2, regardless of the sampling location, the values for 90Sr and
125Sb were the same.
 Meanwhile, the results from the inner surface of the exhaust stack and the inside of the SGTS pipes differed for actinide nuclides.

Activity
concentration

Transport ratio (Cs-137
as reference)

Exhaust stack drain water
Smear obtained by wiping off
the inside of the SGTS pipes
Exhaust stack M1
Exhaust stack M10
Exhaust stack M15

Units 1 and 2
exhaust stacks
Exhaust stack
core

Units 1 and 2
exhaust stacks
Exhaust stack
core

Units 1 and 2
exhaust stacks
Exhaust stackcoreNote) The value indicated for activity concentration is the value with attenuation corrected on March 11, 2011
The values for Co-60, RU-106, Eu-152, Eu-154, H-3, C-14, Ni-63, Se-79, Tc-99, I-129, U-234, U-235, U-236, Np-237, Pu-238, Pu239+240, Am-241, Cm-244 are less than the lower detectable limit.

Fig. 1 Sampling location of thecore of the Units 1 and 2
exhaust stacks*1

Fig. 2 Results of analyzing the core samples
collected from exhaust stacks in Units 1 and 2

Fig. 3 Transport ratios of the Units 1 and 2 exhaust stackcore, samples of the smear obtained by wiping off SGTS
pipes, and exhaust stack drain water*1 TEPCO HD, “Fukushima Daiichi Nuclear Power Station Units 1 and 2 exhaust stack dismantling work progress status” February 27, 2020.
*2 A. Shimada et al., Sci Rep 12, 2086 (2022). https://doi.org/10.1038/s41598-022-05924-2©International Research Institute for Nuclear Decommissioning347

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Stagnant water sludge)◼

The solid components (sludge) in the stagnant water in the building basement is important for ascertaining the status of contamination ofdemolition waste, and hence data on the sludge is being accumulated. The tank water collected while dismantling the tank in which
stagnant water from the Unit 3 torus room and the main steam isolation valve (MSIV) room as well as the treated water were stored, wasanalyzed.
◼ All the water samples contained dark brown deposits (Fig. 1). When the samples were filtered sequentially using filters with hole size 10,
1, 0.1, and 0.02 µm, almost all the sludge was collected using the filter with hole size 10 µm in the case of all the samples (Fig. 2).
◼ 99.6% or more of the Pu, Am and Cm nuclides were detected from the residue of the filter with hole size 10 µm. Fe was the main elementthat the residue (sludge) of the filter with hole size 10 µm consisted in the case of all the samples.*◼ When the nuclide composition of Pu, Am, and Cm was compared with that of fuel, it was similar in the case of stagnant water, and in thecase of tank water the proportion of Cm was small (Fig. 3)* ICP-AES analysis of the solution obtained by dissolving the residue in nitric acid, etc. was conducted. It was found that 100% weight
proportion was the most common with respect to the detected elements.

Unit 3 Torus room water

Unit 3 MSIV room water

E-D1 tank water

Unit 3 Torus room stagnant
water

Unit 3 MSIV room water

Unit 3 fuel composition

Fig. 1 Appearance of stagnant water samples
Filter hole size

Mass (g)Appearanceof thefiltration
residue

Fig. 2 Filtration results of Unit 3 Torus room water

E-D1 tank water

Units 1 - 3 Average fuelcomposition

Fig. 3 Analysis results of Pu, Am and Cm from the stagnant water sludge (relative proportion)©International Research Institute for Nuclear Decommissioning348

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Stagnant water)◼
◼

◼

Ascertaining the concentration of radionuclides in stagnant water and treated water is helpful in estimating the status of contamination of structural
material of buildings and secondary waste generated from contaminated water, and in investigating the treatment and disposal methods. Hence 79Se wasanalyzed.
If the usual method using hydrochloric acid is used in equipment made of stainless steel, the equipment gets corroded. Hence the method of conducting
the separation operation in a nitric acid solution system was developed and applied (Fig. 1).
79Se was not detected in any of the 17 samples of stagnant water and treated water that were submitted for analysis so far (Table 1). Further, thecollection rate of Se carriers was good at 74 - 96%.

Table 1 Results of analyzing 79Se present in stagnant water and treated water
Sample number

Sampling date

Sampling locations

Concentration (Bq/cm3)

LI-2RB5-1

August 7, 2013

Unit 2 PCV< 5×10- 2

LI-3RB5-1

October 22, 2015

Unit 3 PCV (near the water surface)< 5×10- 2

LI-TS-02

February 13, 2020

Unit 2 Torus room (filtrate of 0.22µm filter)< 5×10-2

Analysis sample
Se carrier
Nitric acid, 10% Sodium nitrite
Anion-exchange resin

0.1 M nitric acid

LI-1RB-1

December 8, 2016

Unit 1 R/B basement (collected from the water sampling port)< 5×10- 2

Cation-exchange resin

0.1 M nitric acid

LI-1TB5-6

September 30, 2015

Unit 1 T/B (top of OP1900)< 4×10- 2

LI-2TB7-1

September 25, 2015

Unit 2 T/B basement< 4×10- 2

LI-3TB7-1

October 15, 2015

Unit 3 T/B basement

LI-RW2-2

March 11, 2014

Centralized RW basement (KURION inlet)< 4×10-2

LI-RW4-2

March 9, 2015

Centralized RW basement< 5×10- 2

LI-HTI2-2

February 11, 2014

HTI/B basement stagnant water (SARRY inlet)< 5×10-2

LI-KU3-1

July 13, 2011

LI-KU3-3

September 3, 2014

LI-KU4-3

March 9, 2015

Cesium absorption apparatus outlet< 5×10- 2

LI-SA2-2

August 13, 2013

Second cesium adsorption apparatus (SARRY) System-B outlet< 5×10-2

LI-SA2-4

February 11, 2014

Second cesium adsorption apparatus (SARRY) System-B outlet< 5×10-2

LI-SA3-2

August 5, 2014

Second cesium adsorption apparatus (SARRY) System-B outlet< 4×10-2

LI-AR3-1

July 13, 2011

AMP
Filtering
10% L-ascorbic acid

Sedimentation (waste)
Filtering
Filtrate (waste) Sedimentation
Dissolution

Concentrated nitric acid,pure water
Sodium hydroxide

Solid phase extraction resin (TRU-resin)

Cesium absorption apparatus outlet
Cesium absorption apparatus outlet< 5×10-2< 7×10-2< 5×10-2

1 M nitric acid

Sodium hydroxide (pH6-8)

Radioactivity measurement and evaluation of collection rate

Decontamination system outlet< 5×10-2

Fig. 1 Method of analyzing 79Se developed for the nitric acid solution system©International Research Institute for Nuclear Decommissioning349

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Adsorbent)◼ During the early days of the accident, sandbags containing zeolite andactivated carbon got introduced into the Process Building (PMB) and theHigh Temperature Incinerator Building (HTI)*. These sandbags weresequentially analyzed and data on activated carbon was obtained (Fig. 1).
◼ 60Co, 90Sr, 125Sb, 134Cs, 137Cs, Pu, 241Am and 244Cm were detected. (Table
1)◼ As compared to zeolite, except 90Sr, 134Cs, 137Cs, the concentration ofactivated carbon was high.
Fig. 1 Samples of activated carbon from the sandbags

Table 1 Results of analyzing the samples of activated carbon from the sandbags (Bq/g)*1
54Mn

60Co

94Nb

134Cs

137Cs

154Eu

125Sb

No.

Sample number

1

PMB-AD-CH-1< 5x101(1.0±0.1)x103< 2x101(3.3±0.1)x104(5.4±0.1)x105< 3x101(7.9±1.0)x102

Refer
ence

Sandbag zeolite*2
PMB-Ze-01

ND

ND

ND

(8.5±0.1)x106(1.4±0.1)x108

ND

ND(Approx. 312 days) (Approx. 5.3 years)

90Sr

No.

Sample number(Approx. 2.0×104(Approx. 2.1 years) (Approx. 30 years) (Approx. 8.6 years) (Approx. 2.8 years)
years)

238Pu(Approx. 29 years)(Approx. 88 years)

239Pu+240Pu(Approx. 2.4×104 year,
Approx. 6.6×103 year)

241Am

244Cm

(Approx. 4.3×102 year)(Approx. 18 years)

1

PMB-AD-CH-1(5.8±0.1)x105(1.6±0.1)x100(5.2±0.3)x10-1(5.6±0.1)x100(1.8±0.1)x100

Reference

Sandbag zeolite*2
PMB-Ze-01(1.1±0.1)x107

ND

ND(2.4±0.6)x10-1

ND*1 Activity concentration was calculated based on the fragmented mass. Activity concentration was corrected on the sampling day (February 27, 2020). The
numerical value after ± in the analytical value is the error in coefficient value.
*2 Activity concentration value is the value corrected on the sampling day (February 12, 2020)©International Research Institute for Nuclear Decommissioning* The Commission on Supervision and Evaluation of the Specified Nuclear Facilities (87 th), Document 3-3

350

(a) Efficient characterization ① Investigation on analytical inventory estimation
methodsi. Implementation of analysis (Analysis results Iron co-precipitation slurry)◼ 4 samples of the iron co-precipitation slurry from the Advanced Liquid
Processing System (ALPS) were analyzed. (Table 1, Figure 1)◼ During the analysis, the samples were dissolved in acid (nitric acid) and inalkali fusion (a mixed flux comprising sodium peroxide and sodium
carbonate), and the results were compared. Even though the status of
dissolution differed, the results of radioactivity analysis by means of
dissolution in acid and those in the case of alkali fusion were the same.
◼ In all samples Sr-90 concentration was dominant, and the analysis results
indicated similar tendencies. Pu-239, Pu-239+240 and Am-241 weredetected in all samples、and Cm-244 was detected in one of the samples.

Fig. 1 Appearance of the analysis samples

Table 1 Iron co-precipitation slurry analysis samples
Water treatment results

Sampling
HIC dose rate
(mSv/h)Sampling date

Sampling locations(HIC number)Collection location

0.785

October 15, 2018

Second facility(PO641180-55)

In the vicinity of the center of
HIC(Depth 1000mm)

Existing ALPS B system
October 29, 2014 November 3, 2014
Iron co-precipitation slurry

2.093

October 16, 2018

Second facility(PO646393-109)

In the vicinity of the center of
HIC(Depth 1000mm)

EAL-S7-1

Existing ALPS B system
Iron co-precipitation slurry

March 14, 2014

April 19, 2014

1.0

October 17, 2018

Second facility(PO641180-11)

In the vicinity of the center of
HIC(Depth 1000mm)

EAL-S10-1

Existing ALPS A system
Iron co-precipitation slurry

November 11,
2013

November 18,
2013

3.612

October 18, 2018

Second facility(PO625899-054)

In the vicinity of the center of
HIC(Depth 1000mm)

Reference
AL-S1

Existing ALPS
Iron co-precipitation slurry–

June 5, 2014

0.633

June 5, 2014

Existing ALPS
(PO641180-96)

In the vicinity of the center of
HIC(Depth 1000mm)

Sample number

Sample name

EAL-S6-1

Existing ALPS C system
Iron co-precipitation slurry

EAL-S8-1

Start date ofpassing water

End date ofpassing water

July 5, 2014

July 10, 2014

©International Research Institute for Nuclear Decommissioning351

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Iron co-precipitation slurry)

Reference(Reference) AL-S1

EAL-S10-1
EAL-S7-1
EAL-S8-1

1.E+06

EAL-S6-1

1.E+04

Reference
AL-S1

1.E+02
1.E+00

0%

50%

100%

Mg(OH)2

Ti(OH)2

Co(OH)2

Al2O3

NuclidesFig. 1 Transport ratios of radionuclides present in the iron co-precipitation slurryCm-244

Zn(OH)2
Am-241

1.E-04

Pu-239+ Pu-240

Ca(OH)2

Pu-238

SiO2

Sr-90

FeO(OH)・H2O
Sb-125

1.E-02

Cs-137

Transport rate (Cs-137 standard)

Transport ratio (Sr -90 standard)◼
◼
◼

Based on the transport ratio of Sb-125 using the Sr-90 standard, it was found that there was a strong correlation between the migration behavior of both
nuclides. (Fig. 1)
The transport ratio of Pu-238 and Pu-239+Pu-240 and Cm-244 using the Sr-90 standard suggested that the migration behavior differed.
The variation in the transport ratios of Sb-125 and α emitting nuclides was smaller in the case of the Sr-90 standard as compared to the Cs-137 standard.
Fe was the main chemical component of the slurry along with Si, Ca, Zn, etc. (Fig. 2)

Co-60

◼

150%

Fig. 2 Chemical composition of the iron co-precipitation slurry (chemical compounds are assum©International Research Institute for Nuclear Decommissioning352

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Analysis results Concentrated liquid waste slurry,carbonate slurry)◼ The concentrations of Pu, Am and Cm nuclides in the concentrated liquid waste slurry and carbonate slurry from ALPS havebeen obtained in the past, and in addition 6 samples were analyzed for U and Np nuclides (Fig. 1).
◼ 237Np concentration was in the range of 0.105 – 0.68 Bq/g. 236U was detected in all the samples, and thus the samples
contained fuel components. With 236U as reference, when the transport ratio to slurry was calculated, it was found that thetransport ratio for 237Np was about 10 to 300, which was equivalent to or greater than the transport ratio of contaminated water(stagnant water and treated water) (Fig. 2). It became clear that 237Np which migrated along with the contaminated water
migrated to these slurries.
1.0E+03
237Np

238Pu

239Pu

244Cm

Transport ratio
236U基準）236輸送比（U as reference

1.0E+02

1.0E+01

Data on Np obtained this year

1.0E+00

Before heating

1.0E-01

1.0E-02

1.0E-03採取日

13/08/07

1

13/08/07

2

15/10/22

5

15/09/25

6

15/10/15

7

13/07/09

19

15/03/09

20

14/06/05（ 増

炭酸

18/03/11（ 増

炭酸

18/03/04（ 濃

炭酸

20/01/12（ 濃

濃廃20/01/15（ 増

濃廃

18/08/23（ 既

炭酸

18/08/23

K

鉄

15/05/13

S

18

16/07/25

S

17

16/07/25

S

16

14/08/05

H

15

13/08/13

H

14

16/07/25

H

13

15/09/08

H

12

14/11/25

集

11

14/08/05

集10

13/08/13

集

9

14/11/25

2

8

14/09/03

水 3

4

16/12/08

2

（ 既

（ 既

号

号

面 号

レ 号

号

号

号

中

中

中

T

T

T

T

A

A

A

U

U

U

鉄 設

炭 設

濃 縮

H 縮

炭 設

炭 設

炭 設

炭 設

機

機

近 機

ー 機

R

機

機

R

R

R

I

I

I

I

R

R

R

R

R

R

共

酸

縮 廃

2 廃

酸

酸

酸

酸

P

P

傍 P

チ P

/

T

T

W

W

W

地

地

地

地

R

R

R

I

I

I

沈 多

塩 多

水 液

ブ 液

塩 多

塩 多

塩 多

塩 多

C

C

C

ン C

B

/

/

地

地

地

下

下

Y

Y

O

O

O

ス 核

沈 核

タ ス

ル ス

沈 核

沈 核

沈 核

沈 核

V

V

V

グ V

地

B

B

下

下

下

N

N

N

ラ 種

殿 種

ン ラ

ー ラ

殿 種

殿 種

殿 種

殿 種

近

下

地

地

リ 除

ス 除

ク リ

タ リ

ス 除

ス 除

ス 除

ス 除

下

下

ー 去

ラ 去

ー

ン ー

ラ 去

ラ 去

ラ 去

） 設

リ 設

A

ク

リ 設

リ 設

リ 設

リ 設

備

ー 備

0

A

ー 備

ー 備

ー 備

ー 備

）

0

）

）

）

）

）

2

採取場所

After heating
Fig. 1 Appearance of the
carbonate slurry sample(EAL-S21-2)

3

15/10/22
グ 3

傍

1

2

滞留水試料番号下Contaminated water

下

Y

K

K

9

）

A

C

系

系

統

統

Slurry炭酸

ラ 去Fig. 2 Transport ratio of actinide nuclides to the contaminated water and slurries (236U as reference)©International Research Institute for Nuclear Decommissioning353

(a) Efficient characterization ① Investigation on analytical inventory estimation
methodsi. Implementation of analysis (Analysis results Adsorbent)◼ In the Advanced Liquid Processing System (ALPS) the adsorbent used in the multi-nuclide removal system (treatment
column) becomes secondary waste. Multiple adsorbents are used, and the nuclides that used adsorbents containdifferent adsorbent depending upon the type. Hence, it is important to analyze the radioactivity for each adsorbent inorder to investigate treatment and disposal methods. Samples from resin based adsorbents 1 and 2 collected fromthe storage container (HIC) and ferrocyanide adsorbent were analyzed.
◼ The method of analyzing each adsorbent was investigated prior to the analysis (Fig. 1, 2, 3).
◼ 60Co, 125Sb, 134Cs, 137Cs were detected in all the adsorbents. 90Sr was detected only in the resin based adsorbent 2.
106Ru was detected in the resin based adsorbent 1 and 2. α nuclides (Pu, Am, Cm) were not detected.
Sampleγ ray analysis
Ge semi-conductor
54Mn, 60Co, 94Nb,

Heated-air drying<80℃

Thermal decomposition

106Ru, 125Sb, 134,

270℃
180 minutes

4M nitric acid

137Cs, 152, 154Eu

Heating

Total 300 minutes

Filtering Membrane filter
Dissolution residue
γ ray analysis
Ge semi-conductor

Chemicaltreatment

After thermal
decomposition
During dissolution in acid

After filtering andachieving constantvolumeChemical treatment

β ray analysis α ray analysis
Gas flow
Si semi-conductor
90Sr

Fig. 1 Appearance of theadsorbent (ferrocyanide) sample

238 ,239+240Pu,
241Am, 244Cm

Fig. 2 Method of analyzing theadsorbent (ferrocyanide) sample

Dissolution residue
Fig. 3 Sample while analyzing theadsorbent (ferrocyanide)©International Research Institute for Nuclear Decommissioning354

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Database of analysis results)◼ The obtained analysis results were recorded in the database of analysis results “FRAnDLi” and thus data was expanded on an
ongoing basis.
 The results of radioactivity analysis of the contaminated water, the quantity of treated water and the amount of waste publishedby TEPCO, as also the data on radioactivity analysis, element analysis, etc. obtained during this project, were recorded. Dataon 109 analysis samples was recorded this year. (Table 1)◼ FRAnDLi will continue to be released on the Internet so that it will be available for use to the concerned parties. (Fig. 1)

Analysis

Data obtainedfrom

Number ofcases

Rooftop rubble

This project4

Building sludge

This project

8

This project

54

Tokyo ElectricPower
Company

24

This project5

Tokyo ElectricPower
Company

9

SlurryThis project4

AdsorbentThis project1

Amount of stored waste, quantity oftreated water, etc.

Tokyo ElectricPower
Company

SampleDebrisStagnant waterContaminatedwater
Treated water

Secondary waste
generated fromcontaminatedwater treatment

AmountNumber of accesses
per month

Number of accesses per month (cases)

Classifi
cationApr Jun Aug Oct Dec Feb Apr Jun Aug Oct Dec Feb Apr Jun Aug Oct Dec
2019

1176

Total number ofaccesses

Total number of accesses (cases)

Table 1 Data recorded (As of end of February 2022)

2020

2021

Fig. 1 Shift in the number of times the analysis
database (FRAnDLi) is accessed* Fukushima Daiichi Radwaste Analytical Data Library (FRAnDLi)
URL:https://frandli-db.jaea.go.jp/FRAnDLi/©International Research Institute for Nuclear Decommissioning355

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Leverage in decommissioning)◼ In order to reduce the risks in storage management, measures are being investigated and implemented, andanalytical data is being used.

Stagnant water sludge
Tokyo Electric Power Company Holdings Inc., “Status of studies aimed at α nuclide removal”
Secretariat of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment (86th session), January 28, 2021

Sludge from the decontamination system
Tokyo Electric Power Company Holdings Inc., “Status of studies aimed at stabilization treatment of ALPS
slurry” The Commission on Regulation of Radioactive Waste in the Specified Nuclear Facilities (7th session),
Document 2, July 23, 2018.

ALPS Slurry
Tokyo Electric Power Company Holdings Inc., “Design aimed at slurry stabilization
treatment” The Commission on Supervision and Evaluation of the Specified Nuclear
Facilities (88th session), Document 1-2-1, February 22, 2021.

Zeolite sandbag
Tokyo Electric Power Company Holdings Inc., “Results of basement environment investigationin the Process Main Building” Secretariat of the Team for Countermeasures for
Decommissioning, Contaminated Water and Treated Water (93rd session), August 26, 2021
©International Research Institute for Nuclear Decommissioning356

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Samples database)◼ Data on the source and properties of the samples is important while using the analytical data. Also, information on stock isessential for managing the analysis samples. Hence, a database for managing analysis samples was considered.
◼ The information recorded in the samples database is closely related to waste management and analysis supply chain in 1F (Fig.
1), and is present in multiple formats (numbers, text, photos, drawings, etc.). The software is expected to be able to manage thehistory concerning updates to the recorded data. Hence an independent database of analysis results was developed, based on
the Wiki software (MediaWiki).

Waste management

Analysis supply chain

Database

Planning

Analytical data

Sample information
Waste
generation

Waste
information

Collection
Sample information

SampleDatabase
Not released at present

Stabilization

Storage

StorageAnalytical data

Treatment

Disposal

Sample information

Transport

AnalysisAnalysis results・ Okuma・ Ibaraki (CPF, NDC, NFD, etc.)

Analysis Results
Database(FRAnDLi)
ReleaseFig. 1 Relation between waste management activities and information related to analysis©International Research Institute for Nuclear Decommissioning357

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
i. Implementation of analysis (Samples database)◼ Trial runs of the Wiki database for recording data on samples are being carried out (Fig. 1). At present, data
entry is going on internally at JAEA. Data will be expanded gradually for the use of concerned parties who
access the database externally (Fig. 2).
Basic information on waste

JAEA

ServerAnalytical
data
(CSV)

PDF

Sampleinformation(CSV)Media wiki
Upload or link

Incorporation of data

Conversion
XML dump(Backup)

List of samples
Downloadable
Detailed information on samples

Displayed in thebrowser

Import to a
separate
Media wiki

External users(Already registered)Fig. 2 Configuration of the database(Stage at which database is sharedwith external parties in the future)

Fig. 1 Hierarchy of data in the database (temporarily called as SampleBase)©International Research Institute for Nuclear Decommissioning358

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods

Results of analyzing actual waste
Nuclides for which significant values wereobtained from the analysis results
Nuclides evaluated using thelower limit of detection

Nuclide concentration [Bq/t]◼ Achievements up to FY2020
 Methods for estimating inventory continued to be
developed and improved thus making it possible toquantitatively evaluate uncertainties along withaccumulation of analytical data, and thus results
conforming with the analytical data were obtained (Fig. 1).
These results will be made available for future
investigations on disposal, etc.
 Promising fundamental measures were investigatedregarding the methods for further reducing uncertainties(Fig. 2).
◼ Goal Reduction of uncertainties in inventory estimation so thatthe estimation matches analytical data.
◼ Details of implementation
 Conducting investigations on contamination mechanismthat contributes to the investigations on contamination
composition of nuclides (grouping), and investigations on
waste classification (population) that use the status of
utilization of newly obtained analytical data and the statusof generation and storage of waste as reference, with the
purpose of enhancing the accuracy of estimation.
 Investigating measures for enhancing the accuracy during
analytical estimation of solid waste inventory based onthese, and evaluating the results.
◼ Indicators for the achievements of goal Classification based on the occurrence and storage form,
etc. of solid waste is investigated and presented.
 The model of the contamination mechanism investigatedlast year is reviewed and shown.
 Applicability to the solid waste inventory analyticalestimation method is evaluated.

Statistical estimation
Maximum value
95%ileMode
Geometric mean
50%ile5%ile
Minimum value*

Fig. 1 The estimated ALPS adsorbent inventory and analytical data (yellow mark)

Results of analyzing actualwaste

SARRY inlet concentration [Bq/ cm3]

Analytical data on the
SARRY inlet and outlet
upstream water

Estimated distributionof the SARRY
decontamination factor

Nuclide concentration [Bq/t]

Estimated distribution
and analytical data on
carbonate slurry

Fig. 2 Investigation of new methods for estimating the inventory of
secondary waste generated from contaminated water treatment(Method of estimating the decontamination factor of the previous process
based on the water analysis data)©International Research Institute for Nuclear Decommissioning359

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (Difference in contamination
depending on the unit)

建屋内気中放出割合のグルーピング検◼ Introduction of the method using Bayesianstatistics is being considered in order toquantitatively determine waste classification.
 Parameters of the t distribution of 2 groups arestochastically obtained and the overlap of
distribution is evaluated.*◼ The uncertainties in distribution parameters of thenuclide migration ratio in each unit were evaluatedfocusing on the group of elements that are
somewhat volatile (semi-volatile), as the nuclide
composition of contamination. (Fig. 1)
 While the data distribution is similar in Units 1 and3, since the distribution range in Unit 2 is wider ascompared to Units 1 and 3, classification into Unit
2 and Units 1/3 is considered to be worth
investigating.
 However, since the data deviates depending on
the Unit from where the analysis samples are
obtained, these results are not definitive.
◼ Waste classification methods that consider
difference in contamination depending on the unit
have promising prospects for quantitativeevaluation.

Semivolatile Elements
Units
1+2+3
1＋2＋3号機

1＋3号機
Units
1+3

• 2号機のデータは、1・3と比較して分布幅が大き• 1, 3号機はデータ分布がぼ同じ(2号機)/(1+3号機)の区化が検討に値する1, 3号機
分布がほ

2号機
Unit 2

1号機
Unit 1

3号機
Unit 31, 3号機は分布がほ
1・3号をま確信の程機に比較

Fig. 1 Breakdown of the migration percentage of semi-volatile
nuclides (H-3) to the air depending on the Unit
*The purple section indicates credible interval (in Bayesianstatistics). Dark color: 25-75%tile, Light color: 2.5-97.5%tile* John Kruschke, “Bayesian Estimation Supersedes the t-Test,” Journal of Experimental Psychology, 142(2), 573–603 (2013).
©International Research Institute for Nuclear Decommissioning360

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (Difference in contamination
depending on the element)◼

The method of waste classification that takes into consideration the difference in contamination depending on the unit as explained in theprevious section, can be considered for application with respect to the difference in contamination depending on the element as well.
◼ Regarding volatile nuclides, since the behavior of C in terms of transport ratio is different as compared to Se and I, the application is being
carried out on a trial basis. (Fig. 1)◼ Among the volatile nuclides, the distribution of Se and I is similar to each other, however, C has a wider range of distribution and shows analtogether different picture. Hence, it is believed that it would be appropriate to separate C into a different group.
建屋内気中放出割合のグルーピング検討: 核種(2)◼ In this manner, trials are being continued.
Volatile Elements

建屋内気中放出割合のグルーピング検討

Volatile Elements

Total forTransport ratio

核種合計
Nuclides

Other
than
Se以外
Se

Se

Transport ratio

Se

Cの輸送比がグループの他の元素に比較して大きい

Volatile Elements: C, Se, I
Other

C

C

I

Other

Ithan
以外
I

Transport ratio

Transport ratio

I

Cthan
以外
C

Fig. 1 Example of investigation related to classification ofvolatile nuclides

17

and I do not need to be• Se,Se
I は分離を要さない
separated• Cの輸送比がグループの他の元素に比較して大きい

Fig. 2 Difference in distribution parameters of transport ratio of volatile nuclides*The purple section indicates credible interval (in Bayesian statistics).
Dark color: 25-75%tile, Light color: 2.5-97.5%tile©International Research Institute for Nuclear Decommissioning361

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (High radiation rubble)◼ In the past, inventory estimation was carried out for the large amount of rubble categorized by dose rate. However, it is desirable
to perform estimation for waste that is actually stored. Whenever waste is generated in the form of rubble or felled tree, it isrecorded in the Management Form and managed. Considering that these forms would be used, investigations were conducted
on development of methods for estimating contamination distribution of high radiation rubble, while referring to data on surface
dose rate of the rubble, etc. (metal, concrete) that is stored.
◼ Rubble that is stored in containers and is categorized under surface dose rate 30 mSv/h or more, is stored in the solid waste
storage facility and managed (Fig. 1). It was assumed that the rubble gets mixed and averaged out in the accumulation zone,
and since it is likely to have been generated indoors, attempts were made to calculate the nuclide migration using thepercentage of emission in air inside the building (Fig. 2).
Calculating the amount of nuclides
based on combustion calculation

Metal
Quantity[m3]

Total quantity of nuclides
in core fuelQuantity[m3]

Adding nuclides from the reactor internal
structures•The quantity (cladding, structural material, etc.) is
calculated based on the Establishment permit
application, etc.
•The amount of radioactivity is the value
calculated by means of combustion calculation

Concrete

Indoor air emission ratio
Calculating the amount of nuclides released
in air indoors

Composition assuming contamination
inside the building (Cs-137 ratio)

Estimated values of the surface dose rate distribution of containers in which high radiation rubble is stored, which refer to the
distribution of the surface dose rate of stored waste mentioned on the “Rubble and Felled Trees Management Form”
Set surface dose rate
Metal (30-43mSv/h)
Metal (43-63mSv/h)
Metal (63-91mSv/h)
Metal (91-131mSv/h)
Metal (131-189mSv/h)
Metal (189-274mSv/h)
Metal (274-395mSv/h)
Metal (395-572mSv/h)
Metal (572-826mSv/h)
Metal (826-1194mSv/h)
Total

Volume [m3]
2,551
1,447
2,388
1,628
1,628
832
651
109
109
109
11,451Category
Concrete (30-43mSv/h)Concrete (43-63mSv/h)Concrete (63-91mSv/h)Concrete (91-131mSv/h)Concrete (131-189mSv/h)Concrete (189-274mSv/h)Concrete (274-295mSv/h)Concrete (395-572mSv/h)Concrete (572-826mSv/h)Concrete (826-1194mSv/h)
Total

Volume [m3]
2,125
1,077
1,420
634
669
550
483
665
242
302
8,166

Calculating the amount of Cs-137 adhered to the rubble basedon the obtained surface dose rate
Corrected on March 11, 2011 (20 years)

Dose rate [mS/h]

Fig. 1 Distribution of the surface dose rate of the stored waste

Amount and composition of nuclides in the rubble outside thebuilding for each category based on the obtained dose rate

Uncertainties
considered

Fig. 2 Approach towards developing the high radiation rubble inventory estimation method©International Research Institute for Nuclear Decommissioning362

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (Reflection of increase in data)◼ The inventory of waste is calculated once again by adding the data obtained last year and obtaining thenuclide migration percentage again.
◼ Quantity of data is increased by adding data obtained last year (Table 1). By doing so, the parameters(average and dispersed) of nuclide migration percentage improve on the whole.
Table 1 The quantity of analytical data used for
calculating the nuclide migration percentage(percentage of release in air inside the building)Semi-volatile

Group

2015

2016

2019
-2017

2020

2021

Semi-volatile

14

27

38

59

60

Volatile

16

37

54

73

78

Alkali earth
metal

15

30

42

65

70
Transition metal

Transition
metal
Lanthanoid
Actinoid

16

40

57

98

106

36

55

69

133

139

Volatile

Lanthanoid, actinoid

Alkali earth metal

Average value μ

Fig. 1 Yearly shift in the nuclide migration parameters (average value µ in the case of normal distribution)©International Research Institute for Nuclear Decommissioning363

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (Contamination mechanism)◼ Investigations continue to be carried out focusing on the difference in medium (air, water) as the contamination mechanism ofradionuclides. Analytical data that serves as reference for the process of settling down and stabilization of nuclides is being
accumulated.* Data on borings cores or samples that have cracks is available for rubble contamination (Fig. 1). In the case of contamination
through water, the composition of nuclides and elements in the deposits present in stagnant water are reported (Analysis
results, stagnant water sludge), and reports related to penetration of α nuclides such as Pu, etc. to concrete have been
made*1. The data suggests stability of the nuclide composition during the process of migration through air, and in the case ofcontamination through water, the possibility of precipitation after briefly dissolving in water.
◼ Topics pertaining to the mediums of air and water that need to be handled while further investigating contamination mechanism
were clarified and identified (Fig. 2).
Correction of data
Addition of dataobtained throughexperiments

Approachbased onnumerical
calculations

3-2
Changes in the status ofelement sorption due to an
event

Boring core
1-1
Simulation of element
diffusion from damaged
fuelconcentration

obtained throughexperiments

Approachbased onanalysis

Rubble90Sr activity

Summary

3-1
Material dependence ofelement sorption

Transition in thegeneral situation ofcontamination

2-1
Distribution of elements in the
structures inside the Reactor
building (R/B) and the building
annex

1-2
Simulation of element
diffusion from spent fuelWaste
inventory
estimation

1-3
Simulation of the element
migration and calculation of theamount of activation productgenerated

137Cs activity concentration (Bq/g)

Fig. 1 Correlation between the activity concentration of
137Cs and 90Sr related to the contamination of rubble that is
likely to be penetrated

4-1
Identification of the key
nuclides for using the scaling
factor method

4-2
Behavioral analysis ofelements

Fig. 2 Correlation diagram of the topics to be investigatedrelated to contamination (Example: contamination through air)

*1 G. Igarashi et al., Journal of Advanced Concrete Technology, 19(9), 950-976 (2021).
©International Research Institute for Nuclear Decommissioning364

(a) Efficient characterization ① Investigation on analytical inventory estimation methods
ii. Investigation on analytical inventory estimation methods (Reflection of improvement
measures)◼

Element group classification of nuclide migration parameters was carried out on a trial basis, the results were reflected and trial calculation of the inventory
was carried out (Fig. 1).
 It is presumed that improved results of the element group are reflected in the estimated value of C, Se and I, and expanded analytical data is reflected in
the estimated value of Co, Ni, Ag and Tc.
◼ The element groups were further broken down based on indoor air emission ratio, outdoor emission ratio and stagnant water migration ratio bydetermining the uneven distribution using Bayesian estimation focusing on nuclides that were identified as important with respect to disposal safety.
 The results of trial calculation are meant to facilitate imagination of the progress when classification is carried out. It should be noted that they have notbeen verified by means of benchmarking, etc. of actual waste.

Increasein C

Increasein Se

Increasein Tc

Estimation before
breakdown

Maximum value
95%ileMode

Geometric mean
50%ile5%ile

Nuclideconcentrat
ion

Minimum value*

Estimation of elementclassification

Maximum value
95%ileMode
Geometric mean
50%ileDecrease in
Co, NiDecrease in
Ag

5%ile
Minimum value** Reference value calculated by eliminating the realization that concentration has become zero. Hence, even if 50%tile and 5%tile are zero, theminimum value has been plotted.

Fig. 1 Results of estimating concrete rubble (obtained rough surface dose rate class < 0.1 mSv/h)©International Research Institute for Nuclear Decommissioning

365

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan
◼ Achievements up to FY2020
 As a requirement of the process of waste management, analysis
has been carried out mainly for the purpose of establishing waste
classification and improving the accuracy of inventory evaluationthat contributes to the investigation of treatment and disposal
technology (Fig. 1).
 Characterization was carried out with implementation of analysis
and estimation of inventory at the core, and in response to theoutcome of the investigations on disposal concepts, the nuclides
to be analyzed were re-examined (Fig. 2).
◼ Goal To propose the policy for characterization from the FY2021
milestone onwards.

•Development of inventory evaluation methods•Establishment of inventory originating from
impurities•Setting of migration rate for each element
group

Setting of
wasteclassification•Setting of waste classification•Setting the evaluation method (SF method *, etc.)classification

Standardization of analysis

Provision of amethod to verify
disposal•Investigation of representativeness of sampling•Verification of application of statisticalmethods•Verification of waste

Investigation of storage and treatment methods

Disposalverification•Analysis for verification of the SF method, etc.
•Analysis by means of the raw waste method•Investigation of storage measures•Investigation of treatment measures•Streamlining the nuclides to be analyzed•Verification of standard analysis methods

Fig. 1 Change in the purpose of analysis associatedwith the progress in decommissioning
Improvement of◼ Details of implementation
 Organization of information related to the analytical capacity(constraints such as evaluation items, volume handled, etc. ofwaste, nuclides, etc. that can be analyzed) of the facilities locatedin Ibaraki prefecture that have conducted analysis in the past as
also Building 1 of the Analysis and Research Center, as analysis
facilities that can be used in the mid-and-long-term.
 Creation of the mid-and-long-term analysis plan considering thestatus of investigations on the requirements mentioned in ii laterin this document, along with waste samples and analytical dataobtained so far and the decommissioning process, and using theresults of the trial analysis plans mentioned in iii later in thisdocument.
◼ Indicators for the achievements of goal Mid-and-long-term analysis plan is created aiming forcharacterization of solid waste.

Improvement ininventory evaluation
accuracy

分析法の改良analysis method

Analysis
分析

Development of
採取法の開発collection
methods

Data
データ

Annual画
年間計planMid-and-long-画
中長期計
term plan

Information
on
物の情報
廃棄
Database
updateデータベース更新

waste

Investigation of
汚染メカニズムの検討
contamination
mechanisms

Investigationof
物の特徴の検討
廃棄characteristics of waste

Model of contamination
汚染機構のモデル
mechanism
Waste
model
物のモデル
廃棄
Development of inventory
インベントリ推定手法の構築
estimation methods
Analysis and
解析評価手法evaluation
methodEstimatedinventory
推定したインベントリ
Investigation of
処分概念の検討
disposalconcepts

Informationthat would be
Prospective
処分概念の候補
今後必要な情報
required in the future
disposal concepts

Fig. 2 Mid-and-long-term analysis plan during the flow of characterization©International Research Institute for Nuclear Decommissioning366

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Method of investigation)◼ Analytical data is being accumulated aiming for final disposal of waste. The analytical data will be used gradually while
investigating inventory evaluation, waste classification and waste verification methods (Fig. 1 on the previous page).
◼ The score of the analysis is roughly the product of the scores required until the type of waste and final waste verification method
is determined, and forms the foundation for the mid-and-long-term analysis plan (Fig. 1).
 The score of the samples needs to be several 1000 points. For performing radiochemical analysis of the difficult-to-measurenuclides, analysis needs to be continued for the long-term with 10 years as the unit for the duration.*◼ Meanwhile, there are uncertainties in the waste classification and the sample score depending on the purpose of analysis.
Hence, investigations were carried out to review the mid-and-long-term analysis plan based on the analysis results, information
on waste and progress of decommissioning, etc. achieved so far.
 With regards to the analysis score, the purpose of the analysis was specifically investigated (Fig. 2). Also, waste classification
was implemented using the analytical inventory estimation method, and quantitative evaluation of the score was implemented
using the analysis planning method respectively.
Waste

Waste
A
B
C
D

A
B
C
D

The analysis score is evaluateddepending on the purpose, and isquantitatively calculated using theanalysis planning method. Differs
depending on the properties and stage.

Waste classification

Waste classification isquantitatively calculatedusing the statistical method

K1
K2
L
M

K
L
M

Improvement ininventory evaluation
accuracy
Setting of waste
classification

Analysis score

Fig. 2 Calculation of the analysis
score that was investigatedFig. 1 Calculation of the analysis score at present

Provision of a method
to verify disposal
Disposal verification*Tokyo Electric Power Company, “Ability to analyze Fukushima Daiichi accident waste”
The Commission on Regulation of Radioactive Waste in the Specified Nuclear Facilities, 2nd session, Document 4, February 12, 2016.
©International Research Institute for Nuclear Decommissioning367

(a) Efficient characterization ② Proposal mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Method of investigation)
Decommissioning process• Road map• NDF Strategic Plan
• TEPCO Execution plan

Preconditions
Setting the goals of mid-and-long-term
research & development related to characterization
Assuming 10 years of term 3–①

Objective of analysis
Investigation of secondary and tertiary classification

Classification and selection of waste to be analyzed
Actual results of analysis

Waste to be analyzed• Analytical data (database)• Analysis methods (knowhow)

Identification of
expectations from analysis,
and their quantification

Substitution with calculation• Probability of combustion calculation

Plan for engineering work, etc.
• Storage management plan
• Implementation plan for engineering work, etc.

Resources for analysis• Facilities (Okuma, Tokai CPF, NDC, NFD)
• Equipment• Human resources• Budget (subsidies)

Nuclides to be analyzed, andtheir concentration

Prospects related to sampling

Application of soft tools,
references• Inventory estimation methods(CRIEPI)
Investigation of the analysis process• Analysis planning method(DQO + Bayes)• Samples, evaluation items, quantitative targets• Annual developments in analysis
Mid-and-long-term analysis plan • Soft tool development plan
• Requirements related to development of analyticaltechnologies©International Research Institute for Nuclear Decommissioning368

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Purpose of analysis)◼ There is a gradual shift in the requirements (purpose of analysis) pertaining to characterization while pursuing research &development.
◼ The purpose of analysis suitable for waste management is set assuming term 3−① and considering the approach of analysis foreach waste.

Term
-

3
①
C)

Next term

E)

Rational grouping–
Grouping (splitting or combining populations) is investigated using Bayesianstatistics and with the nuclide composition, SF applicability and tentative disposal
constraints as the indicators.

Selection of the methods for determining activity concentration–
The applicability of the methods for determining activity concentration isinvestigated using Bayesian statistics, and methods for determining activity
concentration are presented for each main waste.

SF Feasible SF Not feasible

NuclideNuclide Nuclide
II
Ⅲ
I
Activity concentration

Key nuclide concentration
Concentration

D)

Investigation of applicability of the scaling factor method–
The migration process of the main nuclides is investigated based on the
progression of events at the time of accident and the analytical data, physical andchemical properties, etc.
–
Bearing in mind the relation between the migration process of nuclides and thescaling factor method (SF method), the correlation with key nuclides (137Cs and60Co, etc.) is investigated.
–
The inventory is estimated by using the accumulated analytical data to reduce theuncertainties in nuclide migration parameters in the analytical inventory estimation
method, and setting the element groups.

Waste A

Concentration

B)

Estimation of activity concentration distribution and nuclide composition–
The activity concentration distribution and nuclide composition is estimated basedon analytical data, and the nuclide composition is evaluated based on transport ratio.
–
The nuclide migration parameters are set using the analytical inventory estimation
method, and the inventory is estimated.

FrequencyA)

Upper limit of activity concentration

Nuclides to be evaluatedDetermination of waste verification method / Determination of groups–
The method for determining activity concentration is established, and the process
moves on to full-fledged actions towards implementation of disposal.

©International Research Institute for Nuclear Decommissioning369

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Waste priority indicators)◼ The purpose of analysis of each waste and indicators related to waste management were scored for each waste, andthe product of those scores was considered as the indicator of priority of the wastes.

Table Indicators for determining the order of priority for analysis (Proposed)

Purpose of analysis

Details

Priority and quantificationA) Estimation ofactivity
concentration
distribution andnuclide composition

The presence of analytical data (results) to be madeavailable for the analytical method in order toestimate the activity concentration distribution andcomposition, is considered as the indicator.

Waste that has not been analyzed so far isprioritized.
Depends on whether analysis results are available.

1: Analysis results are not available
0：Analysis results are available

B) Investigation of
applicability of thescalingfactormethodThe applicability of the scaling factor method (SFmethod) as a method for determining activity
concentration is investigated on a priority basis.
Applicability of SF method evaluated qualitatively
based on existing information on waste, isconsidered as the indicator.

In order to determine applicability of the SF method,
waste that has a high regularity of contamination isprioritized.
Depends on the stage at which the SF method is
expected to be applicable.

2: Waste to which the SF method is considered tobe applicable.
1: Waste to which the SF method is considered tobe applicable by splitting the population.
0: Waste to which it is difficult to apply the SF
method.

Since the variation in the waste itself and the waste
storage container have a strong impact on theevaluation of the population, these variations are
considered as indicators.

The regularity of contamination seems to becomparatively higher, but waste with large variations
is prioritized.
Depends on the necessity to take measures toreduce variation.

2: Waste with large variation, for which measuresneed to be taken to reduce variation during
the treatment process, etc.
0: Waste that does not require any measures.

The uncertainties in inventory estimation have astrong impact on the uncertainties in disposal
category. Hence this is considered as an indicator.

Waste that has high uncertainties in disposal
category is prioritized.
Depends on the fluctuation in disposal category due
to inventory reference value.

1: Waste in which the disposal category fluctuates
due to inventory reference value (50%tile,
95%tile).
0: Waste in which the category remains constant.

Risk reduction

The necessity to reduce risks during storage in Term
3-① is considered as an indicator.

Waste that is likely to undergo change in properties
due to the stabilization treatment, etc. is prioritized.
Depends on whether or not measures are taken
during Term 3-①.

2: Measures taken
0: Measures not taken

Reduction in storage
load

The necessity to reduce load by taking measures
such as consolidating storage, reduced volumestorage, etc. during Term 3-① is considered as anindicator.

Waste that is likely to undergo change in properties
due to load reduction measures is prioritized.
Depends on whether or not measures are taken
during Term 3-①.

2: Measures taken
0: Measures not taken

AmountSince waste that is large in quantity has an impact
on overall management plan, the amount of waste isconsidered as an indicator.

Waste that is large in quantity is prioritized, and 3
stages are set.

3: Large (exceeding 10,000 m 3)
2: Medium (several 1000 m3)
1: Small (several 100 m3)C) Rational waste
classification(Grouping)

Setting of indicators (numerical value)

Waste management©International Research Institute for Nuclear Decommissioning370

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Priority related to disposal)◼ The estimated values of the inventory are statistically distributed. They are used as reference values (50%tile,
95%tile, geometric mean, etc.) for disposal safety evaluation (Table 1). As a result, in the case of waste for which thedisposal category fluctuates (yellow portion in Table 1), as those distributions are close to the disposal constraints,
reducing uncertainties by means of analysis is strongly required. Hence it is believed that analysis should be
conducted on a priority basis.

Table 1: Impact of inventory reference value (50%tile / 95%ile) on disposal category*
Type of waste

Amount(㎥)

Disposal category ([50%ile (Barrier present/absent)]/[95%ile
Disposal category (Barrier present / absent]) based on the reduction factor of the(50%tile / 95%tile) impact of salts as against the distribution coefficient of instantbased on nearing release in a groundwater scenario
scenario (trench 50
years, pit 300 years)
Facility and natural
Facility
Natural barrier
barrier

Amount(㎥)

Type of waste

Disposal category ([50%ile (Barrier present/absent)]/[95%ile
Disposal category (Barrier present / absent]) based on the reduction factor of the(50%tile / 95%tile) impact of salts as against the distribution coefficient of instantbased on nearing
release in a groundwater scenario
scenario (trench 50
years, pit 300 years)
Facility and natural
Facility
Natural barrier
barrier

KURION

2847

L1/L1

[L2/L3]/[L3/L3][L3/L3]/[L2/L3][L2/L2]/[L2/L2]

Metal inside the building (equivalent to L3 before the accident)

23,677

L1/L1[L1/L2]/[L1/L2][L1/L2]/[L1/L2][L1/L2]/[L1/L2]

AREVA

597

L1/L1

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L2/L3]

Metal inside the building (non-radioactive waste before the accident)

6,454

L1/L1

[L3/L3]/[L3/L3][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

SARRY

1196

L1/L1

[L2/L3]/[L3/L3][L3/L3]/[L2/L3][L2/L2]/[L2/L2]

Concrete inside the building (equivalent to L3 before the accident)

8,766

L1/L1

[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

Slurry 1 Iron co-precipitation

1,833

L1/L1

[L3/L3]/[L1/L2][L3/L3]/[L1/L2][L2/L3]/[L1/L2]

Concrete inside the building (non-radioactive waste before theaccident)

370,586

L1/L1[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L1/L2]

Slurry 2 Carbonate precipitation

7,887

L1/L1

[L2/L2]/[L1/L2][L2/L2]/[L3/L3][L2/L2]/[L1/L2]

Waste inside RPV (L1-debris before the accident)

163

Strata/Strata[L1/L1]/[L1/L1][L1/L1]/[L1/L1][L1/L1]/[L1/L1]

ALPS-4 Ag impregnated activated carbon

222

L1/strata[L3/L3]/[L2/L2][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

Waste inside RPV (L1-stagnant water before the accident)

67

Strata/Strata[L1/L1]/[L1/L1][L1/L1]/[L1/L1][L1/L1]/[L1/L1]

ALPS-2 Titanate

332

L2/L1

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Waste inside RPV (L2-debris before the accident)

131

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L1]/[L1/L1]

ALPS-3 Ferrocyanide

222

L2/L1

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Waste inside RPV (L2-stagnant water the accident)

55

L1/strata[L3/L3]/[L2/L2][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

ALPS-6 Chelating resin

443

L2/L1

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Metal inside PCV (L2-debris before the accident)

176

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

ALPS-5 Titanium oxide

222

L3/L2

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Metal inside PCV (L2-stagnant water before the accident)

988

L1/strata[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

ALPS-7.1

46

L2/strata[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Metal inside PCV (L3-debris before the accident)

181

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Rubble 1 (<0.0005mSv/h)

190,000

L2/L2[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

Metal inside PCV (L3-stagnant water before the accident)

1,030

L1/strata[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

Rubble 2 (0.005 to 0.1mSv/h)

48,000

L2/L2[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

Concrete inside PCV (L2-debris before the accident)

66

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Rubble 3 (0.1 to 1mSv/h)

26,700

L1/L1[L2/L2]/[L1/L2][L2/L2]/[L1/L2][L2/L2]/[L1/L2]

Concrete inside PCV (L2-stagnant water before the accident)

365

L1/strata[L3/L3]/[L2/L2][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

Rubble 4 (1 to 30mSv/h)

52,000

L1/L1[L1/L2]/[L1/L2][L1/L2]/[L1/L2][L1/L2]/[L1/L2]

Concrete inside PCV (L3-debris before the accident)

68

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Rubble 5 (>30mSv/h)

36,000

L1/L1[L1/L2]/[L1/L2][L1/L2]/[L1/L2][L1/L2]/[L1/L2]

Concrete inside PCV (L3-stagnant water before the accident)

381

L1/strata[L3/L3]/[L2/L2][L3/L3]/[L2/L2][L3/L3]/[L2/L2]Felled trees

18,000

L1/L1[L2/L2]/[L1/L2][L2/L2]/[L1/L2][L2/L2]/[L1/L2]

Metal inside the building (L3 air pollution before the accident)

20,143

L2/L2[L2/L2]/[L1/L2][L2/L2]/[L1/L2][L2/L2]/[L1/L2]

Waste inside RPV (L1 before the accident)

230

Strata/Strata[L1/L1]/[L1/L1][L1/L1]/[L1/L1][L1/L1]/[L1/L1]

Metal inside the building (L3 stagnant water before the accident)

3,534

L1/L1

[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L2/L2]

Waste inside RPV (L2 before the accident)

185

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L1]/[L1/L1]

5,491

L2/L2

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Metal inside PCV (L2 before the accident)

1,164

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Metal inside the building (non-radioactive air pollution before theaccident)
Metal inside the building (non-radioactive stagnant water before theaccident)

963

L1/L1

[L3/L3]/[L3/L3][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

Metal inside PCV (L3 before the accident)

1,211

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Concrete inside the building (L3 air pollution before the accident)

7,444

L2/L2

[L3/L3]/[L3/L3][L3/L3]/[L3/L3][L3/L3]/[L3/L3]

Concrete inside PCV (L2 before the accident)

431

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

Concrete inside the building (L3 stagnant water before the accident)

1,322

L1/L1

[L3/L3]/[L3/L3][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

Concrete inside PCV (L3 before the accident)

449

Strata/Strata[L1/L2]/[L2/L2][L2/L2]/[L1/L2][L1/L2]/[L1/L2]

315,069

L2/L2[L3/L3]/[L2/L2][L3/L3]/[L2/L2][L3/L3]/[L2/L2]

55,517

L1/L1[L2/L2]/[L2/L2][L2/L2]/[L2/L2][L2/L2]/[L1/L2]* FY2020 “Provision of disposal concepts and development of safety assessment methods”

Concrete inside the building (non-radioactive air pollution before theaccident)
Concrete inside the building (non-radioactive stagnant water before
the accident)©International Research Institute for Nuclear Decommissioning371

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Priority pertaining to waste management)◼ High priority waste was set based on the prospects of the status of management in Term 3-① in light of theindicators (proposed) for determining the order of priority of analysis.
Outdoorstorage*

Coveringwith sheet

Storage incontainers

Soil-covered-type
temporary storage
facility

Solid waste
storage facility

Table 1 Status of waste management that must be
considered as a pre-condition (Example)

At present

More than 30 mSv/h

Storage iscontinued

Waste
Flammablematerials

Used protective clothing

Branches, leaves and roots
Temporary storage tank

Dissolution
equipment

Miscellaneous solid
waste incineration
facility

Additional
incinerator pretreatment facility

Additional miscellaneous
solid waste incineration
facility* Metal, concrete, pieces of dismantled flange tanks, etc. that have extremely
low dose rate and are to be recycled or reused, will continue to be
temporarily stored in the solid radioactive waste storage facility for the time
being.

Solid waste storage
facility
Building 1 - 9
Additional solid waste
storage facility

Waste sludge (Sludge from thedecontamination system)

Concentrated liquid waste
slurryStored in the temporary storage
facility for spent adsorption towers

Stored in the waste sludge storage facility

Stored in concentrated liquid waste
storage tanks (ready to use)
(Tank)

Those that require
additional shield

Those that require
additional shieldBox culvert

HIC-supporting type(water-tight) box culvert(Example) SARRY
adsorption vessel(Example) KURION
adsorption vessel

ALPS high integrity
container (HIC)

Drainedmedia

Storage after
treatment

Photo showing area
being prepared byproviding dikes,
shield, roof

ALPS
SlurrySolid waste storage facility

Fig. 1 Storage management plan*1*: HIC containing drained media are
stored in the box culvert as well

Continuation ofstorage
Incineration
disposalIndoor storage

Contaminated soil

Indoor storage

Reduction in storage load
(volume reduction)

YesーReduction in storage load
(volume reduction)

No

Yes

ー

No

Stabilization
treatment
Volume reduction
treatment

Risk reduction
Reduction in storage load
(volume reduction)

Yes

Waste sludge

Stabilization
treatment
Volume reduction
treatment

Risk reduction
Reduction in storage load
(volume reduction)

Yes

Concentrated liquid waste
slurryStabilization
treatment
Volume reduction
treatment

Risk reduction
Reduction in storage load
(volume reduction)Yes

Spent adsorbent slurry, etc.

Volume reduction treatment or stabilization treatment(Treatment strategy is being discussed. Whether or not the treatment is necessary will be determined separately.)

Large-sized waste storage facility

Volume reduction
treatment
Indoor storage

Flammable waste

Those that mainlycontain moisture*

Rack

Prototype storage

Waste stored in solid waste
storage facility

Setting up solid waste storage facilities will
be considered based on the amount ofwaste that will be generated in about 10
years in the future.

Adsorption tower(Process main building basement)

Indicators of waste
management

Likelihood ofchange innuclidecomposition

Temporary storageof contaminated soil

Roots and trunks
Outdoor storage

Waste stored outdoors
Volumereduction
facility

Waste
management inTerm 3ー①(*1 Tokyo Electric Power Company Holdings Inc., “Tokyo Electric Power Company Holdings Inc. Fukushima Daiichi Nuclear Power Station solid waste
storage management plan July 2021 edition” July 29, 2021)©International Research Institute for Nuclear Decommissioning372

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Example of trial run of the priority investigation)◼ When evaluation was carried out on a trial basis, the priority level of KURION, SARRY, ALPS slurry / adsorbent, felled trees, etc.
turned out to be high.
◼ In order to steadily continue with the 1F waste stream, the stakeholders need to examine all aspects including the indicators andconfirm the priority.

Table 1 Investigation of the order of priority for analysis (Example)
Demolition Waste
waste
incidental
inside thetocontainme
debrisnt vessel retrieval

Purpose of analysis

Building
concrete
waste

RubbleBuilding
Rubble (accumulat
Protective
metal(stored in ed / low
clothing
waste containers) contaminati
on)

SARRY

AREVA
sludgeALPS
slurryALPS
adsorbent

Concentrat
Activatededcarbonliquidzeolitewaste
sandbagslurryNotavailable(1)

Available(0)

Available(0)Notavailable(1)

Not
available(1)

Not
available(1)

Not
available(1)

Available(0)

Available(0)

Available(0)

Available(0)

Difficult(0)

Splitting(1)

Splitting(1)

Applicable Applicable Applicable
(2)
(2)
(2)

Splitting(1)

Splitting(1)

Applicable(2)

Splitting(1)

Splitting(1)

Applicable Applicable(2)(2)

Required(2)

Required(2)

Required(2)

Not
required(0)

Required(2)

Required(2)

Required(2)

Not
required(0)

Required(2)

Required(2)

Not
required(0)

Not
required(0)Present(1)

Absent(0)Absent(0)Absent(0)

Present(1)

Present(1)

Present(1)

Present(1)

Present(1)

Present(1)

Present(1)

Present
(1)

4

4

2

5

2

6

5

5

3

4

4

3

3

Risk reduction

Absent(0)Absent(0)

Present(2)

Absent(0)Absent(0)Absent(0)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Absent(0)Absent(0)Absent(0)

Present(2)

Absent(0)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Present(2)

Absent(0)Medium
(2)

Small
(1)

Large
(3)

Large
(3)

Large
(3)

Large
(3)

Large
(3)

Medium(2)

Medium(2)

Medium
(2)

Small
(1)

Medium(2)

Medium
(2)

Small
(1)

Small
(1)

Sub-total

2

1

3

3

7

3

5

4

6

6

5

6

6

5

3

Total

8

4

12

12

14

15

10

24

30

30

15

24

24

15

9

Felledtrees

KURIONNot
available(1)

Splitting(1)

Required(2)

Not
required(0)Present(1)

Present(1)

4

4Absent(0)Absent(0)

Reduction in storage load

Absent(0)

AmountA) Estimation ofactivity
Notconcentration
Actual results ofavailable
distribution andanalysis(1)
nuclidecomposition
Applicability ofB) Investigation on
Difficult
SF method
applicability of SF(0)methodMeasures for
Required
reducing(2)C) Rational waste
variation
classification
Change in(Grouping)
Present
disposal(1)
category
Sub-total

Available(0)

Waste management
indicators©International Research Institute for Nuclear Decommissioning

Available(0)373

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Plan for construction works, etc.)◼ In this project, sampling of rubble and secondary waste generated from contaminated water treatment was planned and carriedout in coordination with construction works (List 1) and equipment operation (① Investigation on analytical inventory estimation
methods, i. Implementation of analysis).
 Coordination with construction works and equipment operation is essential for acquiring samples in an efficient manner. Hencea short term plan will be put in place.
◼ The plan for the mid- and long-term will be investigated using the plans for construction works, research & development, etc. as
reference.
 Specific construction methods, schedules related to generation of waste will be considered using the mid-and-long-term plans
as reference (Table 1).
◼ A sampling plan will be investigated in conjunction with TEPCO.
List 1 Works during which samples, etc.
were acquired.
Year

Commencement of fuel debris retrieval in the first unit (In FY2021)

RM Milestones*Due to the spread of the Covid-19 pandemic, there is expected to be a delay of about 1 year.
Improvement of environment inside the buildings<Legend>
Manufacturing and installation of investigation and retrieval equipment

Trial
retrieval(Unit 2)

Duration of work

Safety system operational change

Analysis of fuel debris properties

Period during which change isanticipated

Removing interferences blocking the openings

Correlation between processes

Trial retrieval and internal investigations

Added processes
Improvement of environment inside the buildings

Modified processes

Fuel debris retrieval equipment
Designing & manufacturing

Fuel debris retrieval•Units 1 to 3
– Environment improvement around thenitrogen gas loading facility– RCW system dose reduction
measures– Debris removal from the southern sideof Unit 3– Flange tank decontamination / storage– Collection of flammable materials /hazardous materials•Temporary storage area C– New construction of water discharge
channel– Construction of access roads within
the premises– New installation of communicationlines within the premises– Repairs of outer weir drain valve
control panel

Table 1 Plan for the mid- to long-term when various types of waste will be generated*

Gradually
increasing thescale ofretrieval(Unit 2)

Installation

Safety system Designing, manufacturing and installation
Fuel debris temporary storage facilities / maintenance equipment

Designing, manufacturing and installation
Fuel debris retrieval (gradually increasing the scale of retrieval)Analysis of fuel debris properties* Further increasing the scale of retrieval in Unit 2 will be considered in the future based on trial retrieval, gradual increase in the scale of retrieval, etc.

Dismantling completed in April 2020, a series of tasks completed in May 2020
Removal of top portion of the exhaust stacks in Units 1 & 2
Site investigation and preparation for repair work

Removal of SGTS piping in Units 1, 2

Improvement of the environment within andoutside the buildings in Unit 1
Removal of lower portion of theexhaust stacks in Units 1 & 2

Inside the buildings: Dose reduction / Interference removal
Outside the building: Transformer removal, etc.

Unit 1 PCV internal investigation

Furtherincreasing
the scale ofretrieval(Units 1/3)

Removal of rubble present on the
ground on the southern side of Unit 3

Improvement of the environment within and outside the buildings in Unit 3
PCV water level reduction
Removal of exhauststacks in Units 3 & 4

Inside the buildings: Dose reduction / Interference removal, etc.
Transformer removal
Unit 3 PCV internal investigation

Conceptual investigation

Fuel debris retrieval equipment / Safety systems / Fuel debris storage facilities / Maintenance equipment / Training facility*
Verification of on-site applicability / development
Designing(Remote installation, controlling dust scattering, etc.)

Manufacturing, installation and retrieval* Assuming that investigations will be carried out giving precedence to Unit 3, and Unit 1 will be investigated thereafter.
Note: The contents may change depending on future investigations.

* Tokyo Electric Power Company Holdings, Inc., :FY2021 Mid-and-long-term Decommissioning Plan”, March 25, 2021
©International Research Institute for Nuclear Decommissioning374

(a) Efficient characterization ② Proposal for mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Actual results of analysis and resources)◼
◼

The rubble samples for analysis are obtained from the reactor building mainly from tasks such as construction works, etc.. Planned sampling is limited to
low dose rate areas (Unit 4). Coordination with the decommissioning plan is important for full-fledged sampling from the high-dose rate areas.
In Term 3-① from FY2022, in addition to the analyses conducted at facilities in the Ibaraki area (Table 1), analysis will start being conducted at theRadioactive Substances Analysis and Research Facility Building 1. This is considered to be the precondition for the analytical capabilities for the time
being.

Table 1 Actual results of analysis
Containment vessel

Unit 1 Reactor Building

Containment
vesselTorus room

Torus room

Rooftop

Rooftop

Fifth floor

Fifth floor

Fourth floor

Fourth floor

Third floor

Third floor

Second floor

Secondfloor

First floor

First floor

Unit 2 Reactor Building

Year
2012
2013
2014

Number of analyses

Number of analyses
Unit 3 Reactor
Building
Fifth floor

Fifth
floor

Unit 4 Reactor
Building(Rubble in the fuel pool)

Fourth
floor
Third
floor

Fourth floor

Third floor
Second floor

Secondfloor

First floor

First
floor
Number of analyses

2015
2016

Concrete

Curing tape

Metal

Cloth

Sand

Paper

Pebbles

Smear

Heat insulating
material

Deposits

Solids in stagnant
water

Foreign materials
in the TIP pipe

2017
2018

2019
2020

Paint
Collected samples

2021
Number of analyses

Number of analysis cases
Total
Breakdown
Solid
18
27
Liquid
9
Solid
39
51
Liquid
12
Solid
98
98
LiquidーSolid59
Liquid75
Slurry11
158
Sludge8Incinerate
5
d ash
Solid
49
Liquid105
177
Slurry3
Sludge14
Adsorbent
6
Solid
73
Liquid24
138
Slurry9
Sludge21
Adsorbent
11
Solid
15
Liquid17
62
Slurry16
Sludge4
Adsorbent
10

Agency conducting the analysis
JAEA
NDC
NFD
18
ー
ー
9
ー
ー
39
ー
ー
12
ー
ー
98
ー
ー
ー
ー
ー
42
17
ー
40
20
15
7
4
ー
6
2
ー

ー

5

ー

20
75
3
1
6
60
18
7
2
11
ー
17
10
3
6

29
12
ー
10
ー
11
6
ー
18
ー
14
ー
ー
ー
4

18
ー
3
ー
2
ー
2
1
ー
1
ー
6
1
ー

Fig. 1 Actual results of analysis of samples collected from inside the reactor building©International Research Institute for Nuclear Decommissioning375

(a) Efficient characterization ② Proposal mid-and-long-term analysis plani. Development of mid-and-long-term analysis plan (Analysis plan)◼ The waste anticipated to be generated based on the mid-and-long-term plan will be investigated, and a specific analysis plan will be developed.
 Since fuel debris retrieval will gain momentum, specific sampling methods, number of samples, etc. will be investigated pertaining to sampling
during the process of construction (Improvement of environment within and outside the buildings) which will be carried out assuming thesamples will be acquired from high-dose rate regions.
 The ballpark analysis scores will be kept under control as before (waste classification x standard scores), and they will be concretized andquantified by means of the DQO process.
◼ It is desirable for concerned parties to carry out investigations by bringing together specific information that would contribute to planning.

Table 1 Example of development of mid- and long-term analysis plan
FY2021

Waste from contaminated water
countermeasures“Floor surface sludge”
Waste from contaminated water
countermeasures“Zeolite sandbag”

FY2022

FY2026FY2023

FY2032
Manufacturing, installation and retrieval of collection device for floor surface sludge, etc.

Investigation of methods for collecting floor surface sludge

DQO
Designing measures such as for zeolite sandbag

Manufactu
ring &
installation

Start of analysis
Retrieval

DQO
Start of analysis
Additional miscellaneous solids waste incineration facility

Rubble, etc. processing waste“Incineration processing waste”

Incineration of temporarily stored waste

DQO
Start of analysis
Volume reduction treatment facility

Rubble, etc. processing waste“Volume reduction processing waste”

Installation work

Volume reduction treatment facility

Volume reduction of temporarily stored waste

DQO
Start of analysis
Slurry stabilization treatment

Installation of slurry stabilization treatment facility

Secondary waste from contaminatedwater treatment“Slurry stabilization processing waste”

DQO

Start of analysis
Unit 3 High radiation equipment retrieval
Debris retrieval

Pool fuel removal waste“High radiation equipment, etc.”

Unit 2 High radiation equipment retrieval
Unit 4 High radiation equipment
retrieval

DQO
Improvement of environment within and outside the buildings

Debris retrieval waste“Waste from improvement of theenvironment within and outside thebuildings”

Unit 1 High radiation equipment retrieval

Start of analysis
Inside the buildings: Dose reduction / Interference removal, etc.

Continuation of analysis

DQO
Start of analysis“Interference removal waste”©International Research Institute for Nuclear Decommissioning376

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planii. Investigation of analytical goals based on requirements
Table 1 Examples of exposure dose affecting nuclides
identified while investigating disposal concepts◼ Achievements up to FY2020
 The important nuclides (Table 1) identified while
investigating the disposal concepts were used as
reference and reflected in the analysis plan.
 The estimated inventory and analytical data wascompared, and along with investigating the certainty of
estimation, information contributing to setting specificgoals of analysis was obtained (Fig. 1).

(Cited from the FY2020 Report)
Dose rate*◼ Goal To set specific goals of analysis◼ Indicators for the achievements of goal
 Analytical goals based on the analytical requirements
associated with decommissioning are investigated.

Dismantling waste

KURION

1.4E+02

C-14

I-129

AREVA

9.1E-01

C-14

I-129

Slurry 2 Carbonate precipitation

3.6E+01

C-14

I-129

Rubble 1 (0-0.005mSv/h)

5.1E+00

C-14

I-129

Rubble 2 (0.005 - 1mSv/h）+01.7E+1

C-14

I-129

RPV waste(L1 before the accident)

7.3E+03

C-14

Metal inside PCV(L2 before the accident)

6.5E+00

C-14

Concrete inside the building(Non-radioactive waste before theaccident)

7.5E+00

C-14

Pu-239

Tc-99, Po-210,
Np-237, U-234,
I-129

Ratio ofexposure doseto dominant
nuclides1/1000<

U-238, Pb-210,
Ra-226, Pu-242,
U-236

I-129

＊Highlighted yellow if the dose rate exceeds 10 μSv/h.

Results of analyzing actual waste
Nuclides for which significant values wereobtained from the analysis results
Nuclides evaluated using the lower
limit of detection

Nuclide concentration◼ Details of implementation
 Consolidating the analytical data accumulated so far,the level of difficulty of the analysis methods,
likelihood of alternative means of calculation, etc.
along with the contamination mechanisms investigatedin the previous section ①ii and the requirements ofmultifaceted analyses that is considered in the next
section iii.
 Identifying nuclides for which there is a stronger needfor analysis in the mid-and-long-term based on thoseresults, and deriving targeted concentration, accuracy,
etc. as quantitatively as possible.

RubbleRatio ofexposure doseto dominant
nuclides1/100<

μSv/yType of waste

Secondary waste
generated fromcontaminated water
treatment

Ratio ofexposure doseto dominant
nuclides1/10<

Dominant
nuclidesStatisticalestimation
Maximum value
95%ileMode
Geometric mean
50%ile5%ile
Minimum value*

Fig. 1 Comparison between the results of inventory estimation of
ALPS carbonate slurry and analytical data
©International Research Institute for Nuclear Decommissioning*1 The 56th Session of Secretariat of the Team for Countermeasures for Decommissioning
and Contaminated Water Treatment, Jul 26, 2018.

377

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planii. Investigation of analytical goals based on requirements (Method of investigation)◼ The nuclides to be analyzed for the representative /high priority waste and the target concentration
during analysis were set as the goals of analysis.
◼ In the past, analysis was conducted by selecting 38
nuclides using the domestic track record asreference.
 This is assumed in the design and operation of theRadioactive Substances Analysis and Research
Facility Building 1 as well. For inventory estimation,
nuclide selection was more extensive. 134 nuclides
were selected and evaluated.
 In FY2016, nuclides were selected on a trial basis.
The number of target nuclides was reduced from38 to 30 and it was proposed that 23 nuclides be
selected based on calculations.
◼ The outcome of (i) was used as reference for the
purpose of analysis, nuclides were selected, andtheir respective target concentration wasquantitatively investigated. (Fig. 1)

Purpose of analysis
Group of nuclides

Selection of nuclides
Nuclides to be analyzed
Constraints of analysis
Investigation of target concentration

Goal of analysis

List of nuclides to be analyzed and their target concentration
Waste
A

14C

90Sr

Nuclides
129I
137Cs

125Sb

236U

237N

238Pu

p
B
C
D
E

xxx
xxx
xxx
xxx

xxx
xxx
xxx
xxx

xxx
xxx

xxx
xxx
xxx

–
–
–
–

xxx

xxx
xxx

xxx
xxx
xxx
xxx

Fig. 1 Flow of the Investigation of analysis goals©International Research Institute for Nuclear Decommissioning378

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planii. Investigation of analytical goals based on requirements (Selection of nuclides)◼ The nuclides to be analyzed were selected once again using the methods investigated previously and the results of thoseinvestigations as reference (Fig. 1).
◼ As a result 47 nuclides were selected for analysis (for inventory estimation), and upon considering the results of detection andfuture possibilities it was thought that it would be appropriate to select 17 out of those nuclides using calculations (Table 1).
To be subjected to radiation protection• Using ICRP Pub. 107 as reference

Table 1 Nuclides subject to analysis and calculation

1252 nuclides

Narrowing down
target nuclides

Excluded based on half-life

The range which was 1.5 y in thepast, was increased (includes
106Ru that is detected (half-life1.0y))• T1/2 ≥1.0 y

Classific
ation

NuclidesAnalysis

H-3, C-14, Cl-36, Ca-41,
Co-60, Ni-63, Se-79, Sr-90,
Zr-93, Nb-94, Mo-93, Tc-99,
Ru-106, Pd-107, Ag-108m,
Sn-126, Sb-125, I-129, Cs137, Eu-154, U-234, U-235,
U-236, U-238, Np-237, Pu238, Pu-239, Pu-240, Am241, Cm-244(30 nuclides)

Calculation

Ni-59, Cs-135, Sm-151,
Eu-152, Pb-210, Po-210,
Ra-226, Th-229, Th-230,
Pa-231, U-233, Pu-241,
Pu-242, Am-242m, Am243, Cm-245, Cm-246(17 nuclides)

138 nuclides

Narrowing down

Identification based on disposal safety
45 nuclides

Expansion

Addition contributing to the developmentof inventory estimation methods

Along with referring to existing domestic cases(38 nuclides from the past), the results ofinvestigations related to disposal carried out in
FY2020 were used as references as well (7nuclides were added to the above).
From the perspective of reflecting thecontamination mechanisms, Ru-106 andSb-125 which are next to Cs in terms ofvolatility and which are γ ray nuclides
were added.

47 nuclides

Narrowing down

Substitution with calculation considering
the possibility of not being detected

Nuclides for which isotopes or radiation
equilibrium can be calculated wereconsidered focusing on nuclides that havenot been detected in the past.

Nuclides to be analyzed and subject to calculations in the future (proposed)
17 nuclides to be subjected to calculations from among the 47 nuclides

Fig. 1 Flow of selection of nuclides to be analyzed©International Research Institute for Nuclear Decommissioning379

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planii. Investigation of analytical goals based on requirements (Investigation of quantitative of goals)◼

The results of analysis (quantitative value, limit value) of the selected nuclides carried out so far were consolidated. These values werecompared with the indicators.
 Here, during the analysis of many nuclides, as an indicator, the concentration limit in water was assumed to be 1/100, considering thatsolid samples were subjected to chemical separation in the form of an aqueous solution.
◼ As a result of organization, the nuclides for which the indicator was not achieved were identified as nuclides for which it was desirable toinvestigate analysis methods. (Table 1)
 It is desirable to continue investigating the indicators based on disposal safety.

Table 1 Flow of the investigation of quantitative goals
Comparison
between the
results andconcentration in
the left column*

NuclidesLower detection
limit

Quantitative value

Underwater
concentration limit/ 100 (Bq/cm3)

0.1

Cs-1377.0E-03

1.0E-03

9E-04

Comparison
between the
results andconcentration in
the left column*
1.1

2E-02

1.0

Eu-152

3.0E-02

6E-03

5.0

9E-03

0.4

Eu-1542.0E-02

4E-03

5.0

2E-04

2.0

Actual results of analysis
NuclidesH-3

Lower detection
limit

Quantitative value

Underwater
concentration limit/ 100 (Bq/cm3)

3.0E-02

5.2E-02

6E-01

3.3E-02

Actual results of analysis

C-14

2.0E-02

Cl-36

0.004

Ca-41

3.0E-02

5.2E-02

4E-02

0.8

U-233

4.0E-04Co-58

3.0E-02

5.2E-02

1E-02

3.0

U-234

1.6E-06

8.7E-06

2E-04

0.0

7.5E-10

3.7E-09

2E-04

0.0

3.2E-01Co-60

9.0E-03

7.8E-03

2E-03

3.9

U-235

Ni-63

5.0E-02

5.5E-02

6E-02

0.8

U-236

7.2E-09

7.8E-09

2E-04

0.0

Se-79

4.0E-02

5.2E-02

2E-03

20

U-238

4.0E-09

6.3E-10

2E-04

0.0

Sr-89

1.3E-01

6.9E+00

3E-03

43

Np-237

5.0E-05

2.0E-05

9E-05

0.2

Sr-90

2.0E-03

4.2E-03

3E-04

6.7

Pu-2389.0E-05

1.5E-04

4E-05

2.3

Nb-94

8.0E-03

3.4E-01

4E-03

2.0

Pu-239

1.0E-02

1.4E-02

4E-05

250

4.3E+02

1E-02

43000

Pu-2401.0E-02

5.0E-03

4E-05

125

5.0E-02

1.8E+022E-03

25

2.0E-06

2.0E-03

4E-05

0.1

2.7E-04

5E-05

1.8

5E-05

28000

Nb-95
Tc-99

6.0E-03

Ru-103
Ru-106

4.0E+00

Ag-110m

6.4E-03

1E-02

0.6

Pu-241

5.0E-01

1E-02

50

Pu-242

6.8E-02

1E-03

68

Am-241

9.0E-05

1.1E+02

3E-03

36667

Am-242m

1.4E+00

Sn-126

5.0E-02

5.2E-01

2E-03

25

Am-2438.0E-04

5E-05

16

Sb-125

7.0E-02

3.9E-03

8E-03

0.5

Cm-2422.0E-03-03.9E+2

6E-04

3.3

I-129

2.0E-022.0E-02

9E-04

22

Cm-244

9.0E-05

1.1E-04

7E-05

1.3

6.8E-03

5E-05

136

6.7E-03

5E-05

134

Cs-134
Cs-136

6.0E-01

1.3E-03

6E-04

2.2

Cm-245

2.5E+01

3E-03

8333

Cm-246* Compared with the lower detection limit or quantitative value, whichever is lower. Font color is changed to indicate that the value has exceeded 1.
©International Research Institute for Nuclear Decommissioning380

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods◼ Achievements up to FY2020
 The usefulness of the analysis planning method in which
the DQO process considered to be the standard methodin the environment restoration domain is used as thebase (Fig. 1), and the method of calculating the required
analysis score using the Bayesian statistical method isused in combination, was verified (Fig. 2).
◼ Goal To establish a method for quantitatively calculating therequired analysis score.
◼ Details of implementation
 Investigating an analysis planning method in which the
DQO process developed by the US and Bayesianstatistical method are used in combination, consideringthat the population of waste to be investigated is
tentative, considering the multifaceted requirements ofdisposal, etc. and that the statistically required analysis
score is calculated.
 Selecting several representative solid wastes, andconducting trial runs with existing analytical data and themost recent analytical plan as reference. And
consolidating the appropriateness for future use, andissues with respect to practical application based on the
results.

Fig. 1 DQO process used for planning waste analysis◼ Indicators for the achievements of goal For waste with a tentative population, analysis planning
methods using statistical techniques are investigated
and their appropriateness is evaluated.
Fig. 2 Calculation of the analysis score required for verifying
whether the expected criteria is met (Example)
Research Institute for Nuclear Decommissioning*1 The 56th Session of Secretariat of the Team for Countermeasures for Decommissioning©Internationaland Contaminated
Water Treatment, Jul 26, 2018.

381

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods (Calculation techniques to be considered)◼ The method of combining the Bayesian statistical method with the DQO process continues to be considered for the
analysis planning method.
◼ As a condition for evaluating the analysis score, preparations are underway for investigating 3 purposes (Fig. 1) inaccordance with the purpose of analysis investigated as part of the mid-and-long-term analysis plan.

A) Estimation of activity concentration distribution and nuclide composition

Waste A

Frequency− The activity concentration distribution and nuclide composition are estimatedbased on analytical data, and the nuclide composition is evaluated based ontransport ratio.
− The nuclide migration parameters are set using the analytical inventory
estimation method, and the inventory is estimated.

Nuclides Nuclides
NuclidesActivity concentration− Grouping (splitting or combining populations) is investigated using Bayesianstatistics and with the nuclide composition, SF applicability and tentative
disposal constraints*1 as the indicators.

D) Selection of the methods for determining activity concentration

Next term

Concentration

SF Notfeasible

Key nuclide concentration
Upper limit of activity concentrationC) Rational grouping−SF
Feasible

The applicability of the methods for determining activity concentration is investigatedusing Bayesian statistics, and methods for determining activity concentration arepresented for each main waste.

E) Determination of waste verification method / Determination of groups− The method for determining activity concentration is established, and theprocess moves on to full-fledged actions towards implementation of disposal.

Concentration

Term 3-①

B) Investigation of applicability of the scaling factor method− The migration process of the main nuclides is investigated based on the
progression of events at the time of accident and the analytical data, physical andchemical properties, etc.
− Bearing in mind the relation between the migration process of nuclides and thescaling factor method (SF method), the correlation with key nuclides (137Cs and60Co) is investigated.
− The inventory is estimated by using the accumulated analytical data, reducing theuncertainties in nuclide migration parameters in the analytical inventory estimation
method, and setting the element groups.

Calculation technique 1)(Investigated this year)
Ascertaining the 95%tile value ofconcentration distribution

Nuclides to be evaluatedCalculation technique 2)(Investigated this year)
Applicability of the SF method

Calculation technique 3)(Investigated this year)
Rationalization of grouping for usingSF

Calculation technique 4)(Investigated last year)
Grouping based on concentration
Calculation technique 5)(Investigated last year)
Comparison with disposal constraints

Fig. 1 Calculation technique for calculating the analysis score©International Research Institute for Nuclear Decommissioning382

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods (Method of analysis)◼ 3 methods were investigated related to Step 6 of evaluation of analysis score during the DQO process. The incinerated ash,carbonate slurry, and rubble were respectively investigated with respect to the concentration distribution, scaling factor method,and grouping (waste classification).
◼ The DQO process was implemented in a workshop format with participation by the concerned parties (TEPCO, NDF, IRID) (Fig.
1). The third session is planned to be held shortly. The progress of events is reported in this report.
◼ Applying Bayesian statistics for evaluating the analysis score in Step 6 was considered with the cooperation of UK's NNL. As aresult of preliminary research, it was believed that it would be appropriate to carry out grouping upon comprehensive
examination. With the alternative method of Kruschke's t test as the basis, the Bayesian statistical method investigates theuncertainties in the parameters of t distribution (Fig. 2) *1 Calculated with the statistical computing environment R using theBEST package *2
Step 1
Step 2

Specification of items to be evaluated and their goals

Step 3Specification of the type of data to beentered and the source of that information
Specification of the scope ofdevelopment of the analysis plan

Step 4
Step 5No
Are theYes
investigation
results
valid?

Clarification of issues

Step 6

Step 7

Evaluation of the relevance of theentered data as against the issue

Yes

No
Calculation of the detection probability
using the Bayesian estimation method

Development of the analysis plan

Implementation of analysis

Will the acquired data resolve
the issue?

Implementation ofwaste management

Fig. 1 DQO process for the waste

Fig. 2 Prior distribution and parameters handled by means ofKruschke's method*2(Average µ of t distribution, standard deviation σ, shape ν)*1 J. Kruschke, “Bayesian estimation supersedes the t test,” J.
Experimental Psychology, 142, 573–603 (2013).
*2 J. Kruschke, et al. “Package ‘BEST, (May 18, 2020.
©International Research Institute for Nuclear Decommissioning383

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods (Trial run of the DQO process)“Calculating the analysis score for estimating the activity concentration distribution of incinerated ash discharged from the miscellaneous
solid waste incineration facility” was considered as the purpose of Steps 1 and 2.
◼ In Steps 3 and 4, information on the incinerated ash and the incineration facility that was the source of its generation was consolidated, andthe target was specified.
 The miscellaneous solid waste incineration facility is the target. The additional facility where other objects are incinerated is not be
considered as the target. The ash (approx. 2000 units) generated so far is considered for the population, and the 25 samples collectedduring this project will be analyzed.
 There is 10% (target value) volume reduction by means of incineration. The activity concentration and rate of generation (4, 20 kg/h) of thebottom ash and the fly ash are different. As ash gets stored in drums in the order of generation, there is variation in the ash between
drums and inside the drums.
◼ In Step 5, it was found that further analysis is required.
◼

Exhaust stack
Bag filter

Incinerator
Waste supply
equipment
Exhaust gas blower
Incinerated ash is stored in
drumsType of reactor

Rotary Kiln

Processing
capacity

300 kg/h × 2 system (24 h/d operation)

Objects to beincinerated

Miscellaneous solid waste
・Accessories (Tyvek, underclothing, rubber gloves, etc.)
・Construction work waste material (Waste cloths, trees,
packing material, paper, etc.) and other items

Sample (mSv/h/g)
SampleSecondary
Loading combustion
wastechamber

Exhaust gas filter(HEPA filters 2 levels x 5 rows)
Exhaust gas cooler

Drums
Drums (mSv/h/g)
Fig. 2 Variation in dose rate of collected samples (25 samples)Fig. 1 Overview of miscellaneous solid waste incineration facility* Tokyo Electric Power Company, “Results of the hot run during the Fukushima Daiichi Nuclear Power Station Miscellaneous Solid Waste Incineration Facility
installation work” The Commission on Regulation of Radioactive Waste in the Specified Nuclear Facilities (3 rd session), March 17, 2016.
©International Research Institute for Nuclear Decommissioning384

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods (Calculation using Bayesianstatistics)◼ In Step 6, the analysis score was probabilistically investigated. Ascertaining the 95%tile value was considered
as a parameter of concentration distribution, but it was unable to be calculated using BEST package, andneeded to be programmed. Hence the Highest Density Interval (HDI) at 95% was obtained instead (Fig. 1).
 Investigation of the 95%tile (Fig. 1 top) will be considered next year.
◼ The prior distribution in the case of Bayesian statistics was calculated using the BEST package by referring to
the dose rate of the drums and the samples collected, converting it to 137Cs concentration and thereafter using
it as the input condition (Table 1) for the calculation (Fig. 2).
95% HDI of the 95%ile

Initially envisioned
95% HDI of distribution

Modified conditions

Table 1 Conditions for setting prior distribution(Variation in 137Cs concentration)

Variation

Collection ofdrums *1

Inside thedrums(Sample) *2

Synthesis

137Cs
concentration[Bq/g]

2.6×102-4.2×104

2.4×103(Mean)–

Geometric
mean

3.51

3.38

3.51

Standard
deviation

±1.10

±1.01
(RSD=30%)±1.50*1 Set so that a wider scope is covered based on storage conditions. It is
assumed that the scope can be reduced by using data on stored drums.
*2 Values are being examined closely.

Fig. 1 Modification of calculation conditions

Fig. 2 Prior distribution created bymeans of the makeData function of
the BEST package (n = 2,000)*1 J. Kruschke, et al. “Package ‘BEST, (May 18, 2020.
©International Research Institute for Nuclear Decommissioning385

(a) Efficient characterization ② Proposal for mid-and-long-term analysis planiii. Investigation of analysis planning methods (Creating the analysis plan)◼

In Step 6, further, the posterior distribution was calculated as the distribution of 25 samples most likely to be collected (Fig. 1). The
statistical power (analysis score) was calculated while letting the expected range (95%HDI) to be 102 to 104 (Fig. 2). The distribution of thesamples assumed to be in the drums was wider than the collected samples. Hence many samples are required in order to obtain 95%HDI
with certainty.
 Realistically speaking, calculation that assumes the current number of stored samples to be the population seems to be effective. The
examination will be continued.
◼ In Step 7, the stakeholders carried out discussions based on this information and determined the analysis score.
 Since incineration continues, continuing sampling while the process of analysis begins, increasing the number of analysis samples andconducting evaluation once again based on the DQO process is presumable.
◼ The slurry and rubble will be investigated and the results will be consolidated.

Fig. 1 Posterior distribution of 137Cs calculated bymeans of the BESTmcmc of the BEST package

Fig. 2 The probability of 102 to 104 Bq/g calculated by meansof BESTpower of BEST package, entering 95%HDI©International Research Institute for Nuclear Decommissioning386

(a) Efficient characterization
Summary
Current results
With regards to implementing the analyses, the analytical data on samples (activated carbon
sandbag, iron co-precipitation slurry, 79Se in contaminated water) for which analysis has been
completed, was consolidated, it was sequentially stored in a database and released.
✓ With regards to inventory estimation, the likelihood of a method of classifying (grouping)elements and waste based on the concentration distribution of nuclides using Bayesianstatistics was found out.
✓ With regards to the mid-and-long-term analysis plan, the purpose of the analysis wasconcretized assuming Term 3–① starting from next year.

Issues, etc.
✓Continuing consolidation of analytical data
✓Finalizing the mid-and-long-term analysis plan✓Completing the 3 case studies on analysis planning methods©International Research Institute for Nuclear Decommissioning

387

(b) Development of sampling technology◼ Achievements up to FY2020
➢

➢
➢
➢

Separate preliminary tests were conducted on elemental technologies (making an opening, sampling, closing an opening)that are necessary for sampling from the adsorption tower, and the design specifications of the sampling equipment were
finalized.
Simplified Integrated Sampling Module (hereinafter, S-ISM), Simplified Working Dock (hereinafter, S-WD), simulated
adsorption tower, Simplified Working Stand (hereinafter, S-WS), etc. were fabricated as elemental test apparatus (Fig. 1).
Elemental test plan was drafted, and the elemental tests were conducted using the simulated adsorption tower simulating
the top portion of the adsorption tower.
It was confirmed by means of the elemental test using S-ISM, that the series of tasks including making an opening,
sampling and closing an opening can be competed in the case of the KURION adsorption tower and the SARRY
adsorption tower.
S-ISM

Drill bit

Making an
opening in theadsorption tower

S-WD
Closing the
opening

Simulatedadsorption
tower

Closing plug
Collecting
samples
10-20mL
Sampling head

S-WS

Fig. 1 FY2020 elemental test equipment and overview©International Research Institute for Nuclear Decommissioning

388

(b) Development of sampling technology◼ FY2021 Goals
To design and manufacture sampling system capable of being used for sampling from the 1F usedadsorption towers and the required incidental facilities based on the results of the element tests, andto verify their performance. To draft a sampling plan (including work area, and items to be collected)
for collecting actual samples from the 1F used adsorption towers.
◼ FY2020 Details of implementation• Fabrication of element test apparatus was completed, and element tests on making an opening in
the adsorption tower, collecting samples, and closing the site of the opening were implemented.
• The results of the element tests showed prospects for sampling methods and equipment that canbe used on the site.
◼ FY2021 Details of implementation• The design of the sampling system was investigated based on the results of the element tests andthe investigations on conditions during sampling carried out so far.
• Parts of the sampling system were fabricated.
• Investigations on the sampling work flow and work area and selection of items to be collected wasstarted.
• Fabrication of the sampling system and the incidental facilities was continued and performanceverification tests were implemented.
• A plan (including selection of work area, selection of items to be collected) for collection from
actual adsorption towers, etc. in order to verify the sampling technology was drafted.

©International Research Institute for Nuclear Decommissioning389

(b) Development of sampling technology◼ Indicators for the achievements of goal

FY2021

・The sampling system and the required incidental facilities are investigated,designed and manufactured, and performance tests, etc. are conducted.
・Sampling plan is drafted and presented.
[Target at completion: TRL: Level 4 → 5]
TRL level 4

Stage at which functional tests are implemented at the test manufacturing level as adevelopment and engineering process.

TRL level 5

Stage at which a prototype is manufactured based on the actual equipment and verified ina simulated environment at the plant, etc.
[Excerpts from the general information manual on subsidized projects]◼ Goals and results

FY2021・Investigating the design of the sampling system ⇒ Completed
・Manufacturing the sampling system ⇒ Completed・Performance verification tests of the sampling system ⇒ Manufacturer tests
completed. Verification tests are planned to be conducted after delivery
・Designing the incidental facilities ⇒ Planned to be completed
・Investigating and drafting the sampling plan ⇒ Investigation of work area andselection of items to be collected planned to be completed©International Research Institute for Nuclear Decommissioning390

(b) Development of sampling technology- Development process so far ➢ The development process aimed at on-site introduction is indicated in Fig. 1.

•

Conceptual design[Completed]

Element test [Completed in
FY2020]Investigation of sampling methods,
specifications of devices for on-site
introduction, etc.

Elemental technologies required for sampling(making an opening, sampling, closing an
opening) were tested and their viability wasverified.

No. 2 GB(forsampling)

Working
stand

Installing theadsorption
tower

Sampling system[Manufactured in FY2021]
•
•

S-ISM

No. 1 GB (formaking an
opening)

Outcome of element tests and reflection ofissues
Implementation of the functions and safetymeasures required for hot testing(Radioactive materials leakage prevention
and exposure countermeasures)

S-WD

TemporaryshieldSimulatedadsorption tower

Working dock

(KURION/SARRY)〇 Examination of methods•Investigation of the methods of
implementing “Making an opening in theadsorption tower”, “Sampling” and “Closing
the opening in the adsorption tower”, andintroduction of those methods.

〇 Investigation of specifications for on-site
introduction•Investigation of the methods for preventing
leakage of contamination from inside theadsorption tower•Investigation of sampling methods
considering prevention of excessive
exposure of workers

Testing ofelemental
technologies〇 Outcome・Viability verification of the technology formaking an opening, sampling, and closing thehole・Understanding of the operation parameters
・Feasibility of manual work, etc.

Outcomeof element
tests
Reflectionof issues△ Issues
・Ease of handling (Alignment of S-ISM / S-WD,
rod installation, etc.)
・Variation in pressure resistance while closing

Specifications for on site introduction arereflected while conducting hot testingFig. 1 Sampling system development process©International Research Institute for Nuclear Decommissioning

391

(b) Development of sampling technology
Basic conditions for sampling➢ The basic conditions for sampling are indicated in Table 1.
Table 1 Basic conditions for sampling
Items

Basic conditions
KURIONAll adsorption towers

SARRY

Cesium adsorption towers with a laminated flange type adsorbent
filling part (simultaneous adsorption tower is outside the scope)

KURION

Adsorption tower top board surface

SARRY

Flange with opening for filling adsorbent

Amountcollected

3 or 10 to 20 mL / sample*The amount to be collected is selected depending on the
anticipated activity concentration of the sample

Collection
depth

Up to about 100mm from the surface

Number ofsamplescollected

1 sample in the circumferential direction x 1 sample in the depth
direction(Total 1 sample / Adsorption tower)

Sampling
targetMaking an
opening
Sampling

Sampling

Closing the opening

Installation /removalPressure
resistance

Adsorbent

Fig. 1 KURION adsorption tower top part

LaminatedflangeIt should be ensured that the opening is properly sealed and thatthe seal can be removed even after the opening has been closed.

Pressure resistance during operation should be ensured assuming
removal of adsorbent, etc.

Adsorbent

Fig. 2 SARRY adsorption tower top part©International Research Institute for Nuclear Decommissioning

392

(b) Development of sampling technology
Components and functions of the sampling system (1/3)◼ Overall composition⚫ Work area is set up in an existing building that has a crane that meets the required
specifications.
⚫ Fig. 1 illustrates an image of the overall composition of the sampling system.
Crane
Outside the building

Inside the building

Control panelon the systemside

Sampling
systemWorking dockCompressor

Generator

MonitoringPC
Remotecontrol panel

Adsorption tower

Ventilation unit
Working standFloor surface

Fig. 1 Image of the overall composition of the sampling system©International Research Institute for Nuclear Decommissioning

393

(b) Development of sampling technology
Components and functions of the sampling system (2/3)
Table 1 Components of the sampling system and their overview
Overview

Sampling system[Integrated Sampling
Module (ISM)]

Has the functionality for a series of tasks including “Making
an opening”, “Sampling” and “Closing the opening”. Usedin common for KURION and SARRY

Working dock
[Working Dock (WD)]

Connection device for installing the ISM on the adsorption
tower.
Has the functionality to shield the radiation source at the
top part of the adsorption tower.

Working stand[Working Stand (WS)]

Ensures that the adsorption tower is stably fixed, andsecures a work floor in the area around the lateralshielding and sampling system.

Tools

Composition

Rod

Head

ISM

Dedicated tool heads
WD

Drill bit
S
A
R
R
Y

Laser range finder
WS

Fig. 1 Overall composition

Sampling head

Closing plug

K
U
R
I
O
N

Endoscope

Cleaning buff

Fig. 2 Dedicated tools for ISM©International Research Institute for Nuclear Decommissioning394

(b) Development of sampling technology
Components and functions of the sampling system (3/3)◼

Sampling system (ISM)➢ The shielding lid and canister are activated by means of the motor operated sliding unit, and the shielding at the access port at the bottom of
the ISM and the opening / closing are controlled automatically.
➢ Packing is added to the opening/closing panel for improving air-tightness.
➢ Leakage of contamination outside the system is prevented by means of the cut pieces suction recovery unit and the ventilation unit inside
the system.
➢ The main body frame is made up of stainless steel. Hence it is difficult to get contaminated and even if it does, decontamination is easy.
[Vertical motor][Dedicated tool connector](Including the rotating motor)[Suction filter][Cut pieces suction
recovery unit][Dedicated tools]

Monitoring the operation status of thesystem from a low radiation area

Remote control panel / laptop
Opening/closing
panel[Glove port](Gloves are not shown)

Switchboard
OR
Generator
200V
50Hz
50A

20m

20m

[Motor operated sliding unit]
Opening/closing
panel[Guide mechanism]

Fig. 1 Overall view of ISM (Front)[Cutting lubricant
spraying unit][Control panel on thesystem side]

Ventilation unit inside thesystem(HEPA filter)

Fig. 2 Overall view of ISM (Back)©International Research Institute for Nuclear Decommissioning

395

(b) Development of sampling technology
Items from the outcome of the element tests to be reflected in the sampling system (1/4)◼ Technology for making an opening⚫

⚫
Functions and methods established by means of element tests(Fig. 1, Fig. 2, Fig. 3 (c))➢
Making a stepped hole of the size through which sampling can be
performed, and which can be closed➢
Making a high precision opening by means of the guide mechanism➢
Integrated structure for spraying cutting lubricant and recovering
cut pieces by suction➢
Drill bit structure such that the resulting cut pieces can be easily sucked out

Rotating motor

Improvements➢
The entire structure was reviewed and as a result rods were shortenedby 300 mm➢
A center run-out control function was added to the Φ40 hole making rod➢
In order to increase the efficiency of recovering cut pieces, a skirt
covering the outer periphery of the Φ40 drill bit was added (Fig. 3)➢
In order to improve strength, the shape of the tip of the Φ40 drill bit was
improved➢
In order to review the internal structure of the WD and to improve theinstallation precision, the guide mechanism installation location waschanged from the WD to the bottom of ISM

Suction recovery

Stepped hole

Coupling
Cutting lubricant

Drill bit

The guide mechanism
supports the outer
periphery of the drill bit

Guide mechanism

The hole itself
supports the outer
periphery of the drill bit(a) Before measures(b) After measures

Sliding

Hole making rod

Suction

Skirt(c) Schematic diagram of skirt

Fig. 3 Φ40 skirt⚫
(a) At the time of starting the processof making a Φ60 hole(b) At the time of drilling the hole
Φ60 deep

Functional test (Fig. 4)➢
All aspects were appropriately reflected and it was verified that thepredetermined opening can be made.

Fig. 1 Schematic diagram of the work of making an opening

Suction port
Lubrication hole
Fine grooves in
the cut piece(a) Guide mechanism(b) Outer appearance of the Φ60 drill bit

Fig. 2 Outer appearance of the drill bit(a) The opening(b) When going through Φ50(c) Data log

Fig. 4 Status during the functional test
©International Research Institute for Nuclear Decommissioning396

(b) Development of sampling technology
Items from the outcome of the element tests to be reflected in the sampling system (2/4)◼ Sampling technology

⚫

Range
finderThroughhole③ Sampling

Fig. 1 Sampling work procedures(g/mL)

Vane opening/closing type
Small vane opening/closing type⚫
Functions and methods established by means of element tests
The physical properties of the actual adsorbent obtained are measured, and the simulated
adsorbent that has physical properties similar to those of the actual adsorbent isselected.
➢
Establishment of the procedure for collecting adsorbent (Fig. 1)➢
Development of 2 types of sampling heads (SH), a small vane opening/closing
type sampling head and a bottom opening/closing type sampling head, which has
a dual cylindrical structure and in which the sampling port opens and closes bymeans of clockwise and anti-clockwise rotation (Fig. 2).
➢
The predetermined amount of adsorbent can be collected by running the program(Fig. 3)

Endoscope① Surface measurement ② Surface observation

Specific gravity

Types of sampling heads

Dry sample: 0.75
Wet sample: 1.1

Amount collected

Bottom opening/closing type

Predetermined amount:
10-20mL

Dry samples

Wet samples

Fig. 3 Results of adsorbent collection tests

Improvements➢ The program was improved so that the sampling location
gets set automatically when the value measured at thesurface is entered.
➢ The S-ISM sampling program was available only for lateralsampling. The option of bottom sampling was added.
➢ The function of automatic insertion of the SH into thecanister after sampling was added (Fig. 4).

Canister

Shielding lid

Sampling port
CanistershieldOpening of the vane
Penetration / drilling

Sampling port
(Drilling and sampling)① Initial arrangement of thecanister and the shielding lid② Sliding of canister③ Insertion of SH

Fig. 4 Automatic insertion of sampling head into the canister(a) Small vane opening/closing type
(Lateral sampling type)(b) Bottom opening/closing type
(Bottom sampling type)Fig. 2 External appearance and structure of SH⚫

Functional test
➢ Verifying that sampling operation can be executed according to
the settings©International Research Institute for Nuclear Decommissioning

397

(b) Development of sampling technology
Items from the outcome of the element tests to be reflected in the sampling system (3/4)◼ Technology for closing the opening⚫

Rises to the surface if the frictional
force is not strong enoughPressuregauge

Torquewrench(a) Old closing plug

2.06 Mpa(approx. 4000 N)(b) Balance around theold closing plug

Compressive force

Pressure andtemperature logger

Compressive force

Range finder

Expansion

Frictional force

Frictional force

Functions and methods established by means of element tests➢ The developed closing plug can be installed remotely into the
opening by running the program➢ The pressure resistance test is implemented so that therequired pressure resistance performance is attained (Fig. 1)

Expansion

2.06 Mpa (approx. 4000 N)(c) Balance around the
new closing plug

Screw

Seal

Range finder / torquedisplay tablet

Closingplug

Fig. 1 Pressure resistance test apparatus during theelement test on closing⚫

SpacerPressure resistancetest apparatus

Improvements➢ In the FY2019-2020 version of the closing plug, the frictional forceof the seal alone was counteracting against the internal pressure.
Hence the event of the plug rising to the surface occurred during
the pressure resistance test. Therefore, a new closing plug wasdeveloped which is equipped with a wire net that prevents the plugfrom rising to the surface so that seal is not subject to excess
sheer force (Fig. 2).
➢ The installation position for appropriately installing the closing plug
was determined and the program pertaining to the conditions fordetermining proper closing was improved.

Ring spring
Parts at the bottom

Initial condition
Expanded(d) New closing plug(e) Internal structure of
the new closing plug

Fig. 2 The outer appearance of the closing plug and the difference in the forces acting on
the old and new closing plugs⚫

Functional test➢
Easy remote installation and good pressure resistance (Fig. 3 and Fig. 4)

Fig. 3 Expansion of ring spring

Fig. 4 Pressure resistance test of the
new closing plug

©International Research Institute for Nuclear Decommissioning

398

(b) Development of sampling technology
Items from the outcome of the element tests to be reflected in the sampling system (4/4)◼ Other⚫

Functions and methods established by means of element
tests➢ It was evaluated that the design concept of carrying out
the series of operations including making an opening,
sampling and closing the opening by changing the tools
with the help of the vertical sliding mechanism and therotating motor is reasonable.
➢ It was confirmed that the difference in the structure of the
top portion of KURION and SARRY is taken care of by
the WD and that both can be handled with a single ISM.
➢ It was confirmed that the safety function of automatic
shutdown if a person opens the panel while any of the
motors are in operation, and the emergency shutdown
button functioned properly (Fig. 1).

⚫

Improvements➢ Along with placing the guide mechanism in the ISM rather
than the WD, measures were taken to ensure that the guidemechanism does not come in contact with the WD when the
ISM is installed on the WD.
• Since the ISM is suspended and lowered rectilinearly, theguide pin was installed on the WD (Fig. 2).
• A clinometer that displays the direction in which thecrane should move so that the ISM is in a straight line
while the guide pin is being inserted, was developed andinstalled on the ISM (Fig. 3).

Guidance about direction in which crane should be operated

Pin (long)Pin (short)(a) Door open/close
sensor(b) Main body control panel
emergency shutdown

Fig. 1 Safety function(a) Clinometer (testmanufactured)Fig. 2 ISM installation with the help of
the positioning pin(b) Clinometer(incorporated in the main
body control panel)Fig. 3 Crane operation guidance clinometer(test manufactured)©International Research Institute for Nuclear Decommissioning

399

(b) Development of sampling technology
Items to be reflected in the sampling system for hot testing (1/3)◼ Leakage prevention➢ The functions that the sampling system is provided with for preventing leakage of radioactive
substances during hot testing, based on the investigations directed towards on-site introduction
during the conceptual designing, are indicated in Table 1.
Table 1 Radioactive substance leakage prevention functions (1/2)
Risk of leakage

Inflow of dust from inside theadsorption tower into the
ISM

Function
Assumption: Although the adsorption tower is stored with the hydrogen vent, etc. open, dust is not detected andexcessive dust is not afloat.
Hole making program that makes a through hole without suction by stopping suction right before boring
Shielding lid with seal that opens and closes the seal insertion opening at the bottom of ISM, and its motor
operated sliding mechanism

Diffusion of dust from inside
the adsorption tower to theoutsideSealing mechanism at the region of contact between the guide mechanism and the upper surface of theadsorption tower (Fig. 2), and its pushing mechanism (Fig. 3)

Shielding lid

Seal
Guide
mechanism
Inside theadsorption
tower
Fig. 1 Preventing diffusion of dust inside theadsorption tower(a) Extended
Fig. 2 Sealing mechanism under the guidemechanism(b) Compressed

Fig. 3 Pushing mechanism for theguide mechanism©International Research Institute for Nuclear Decommissioning

400

(b) Development of sampling technology
Items to be reflected in the sampling system for hot testing (2/3)◼ Leakage prevention

Table 1 Radioactive substance leakage prevention functions (2/2)

Risk of leakage

Function
Assumption: According to the results of the element tests, there is no visible dispersion of the collected samples.

Contamination of theinside of the ISM due tothe collected samples

The flow of air is reduced by shutting down the ventilation system during the process from collecting samples to storing them in canisters.
The collected samples are stored in the canisters inside the ISM while they are still enclosed in the sampling head.
The ISM door, panel, cable penetration section have seals and are air tight (Fig. 1).
As there are HEPA filters respectively in the air supply and exhaust systems, the air inside the ISM passes through filters even after the ventilation is
shutdown (Fig. 2).

Diffusion of dust frominside the ISM to theoutsideDust is measured before the ISM door is opened (Fig. 3)
If dust is detected inside the ISM, the air is replaced by means of the ventilation unit (Fig. 3).
The ISM is enclosed inside a house (optional).
Air supply
HEPA filter

Air intake side

Suction recovery unit
HEPA
filterExhaust side

Container
lid

Dust sampler ①
Ventilation unit
HEPA
filterStorage Shielding
lidcontainerGuidemechanism(a) Door(b) Cable port

Fig. 1 Improved air-tightness

Fig. 2 Air supply/exhaust HEPA
filterFig. 3 Image of the composition of the ventilation system(With the valve open during dust measurement)©International Research Institute for Nuclear Decommissioning401

(b) Development of sampling technology
Items to be reflected in the sampling system for hot testing (3/3)◼ Reduction of radiation exposure➢ Functions provided for reducing exposure during hot testing are indicated in Table 1.
Table 1 Exposure risk reduction function
Exposure risk

Function

Exposure during preparations andtidying up

Installation is carried out remotely using a crane.
Shielded by means of the WD that is filled with lead particles (Fig. 1).

Exposure from the upper surface of
the adsorption tower

A remote control panel separated from the ISM main body is provided in order to minimize the time for which one stays on the
WS scaffolding.
Shielded by suspending and lowering a lead shielding mat on the WS.

Exposure from the sides of theadsorption tower

Remote control panel is installed at a distance from the adsorption tower.
Device is remotely operated so that people can keep a distance.

Exposure from the beam shaped
radiation from the penetration section

When the tool rises up back to its original position, the shielding lid automatically closes the penetration section by means of the
motor operated slide (Fig. 2).
The process from sampling to inserting the sampling head (hereinafter, SH) into the canister is carried out automatically (Fig. 3).

Exposure around the collected
samples

The area around the SH is shielded by the canister and the canister shield.

The SH and the rod are separated using a dedicated long reach spanner through the glove port and hence the hands can be
kept away.

Shielding lid
Seal
Lead particle shield
Guide mechanism

Fig. 1 Structure of WD (for KURION)

Fig. 2 Automatic shielding of penetration section bymeans of the motor operated slide

Fig. 3 Automatic insertion of SH into the canister©International Research Institute for Nuclear Decommissioning402

(b) Development of sampling technology
Specifications of the sampling system (1/2)◼ The points in which the results of the element tests were reflected and the points in which items
investigated for hot testing were reflected while developing the sampling system are indicated in Table 1.
Table 1 The points reflected from the results of the element tests and the points reflected from the items investigated for hot testing
Procedure

Reflection of element test results

1. Preparation•Enhancement of the ease of remote installation of ISM withthe help of the positioning pin and clinometer

2. Making an
opening•Reliance on the process of making a stepped hole for puttingin the closing plug
Prevention of center run-out by means of the guidemechanism and by shortening the rod
Continuation of spraying of cutting lubricant and of the cut
pieces suction recovery function
Addition of a skirt to the Φ40 hole making rod
Reliance on the design of the end milling drill bit and its
strength enhancement

•
•
•
•

3. Sampling

•

•
•
•

4. Closing
the opening•••

Setting of sampling procedure in which the internalverification is carried out first
Reliance on the SH sampling port opening/closing
mechanism that has a dual cylindrical structure
Setting of separate procedures for lateral sampling andbottom sampling according to the adsorbent
Provision of automatic sampling program corresponding to
the lateral sampling SH and bottom sampling SH respectively

Reflection of investigations conducted aiming for hot
testing•

Exposure reduction through addition of shielding around theadsorption tower•

Exposure reduction by carrying out work remotely•

Prevention of the flow of dust from inside the adsorption tower
into the ISM and shielding of the beam shaped radiation fromthe through hole by means of the automatic sliding of theshielding lid•

Prevention of diffusion of dust from inside the adsorption
tower to the outside and shielding of radiation by means of
the air-tight structure of the ISM itself and the seal under theguide mechanism•

Prevention of contamination of the inside of the ISM throughautomatic insertion of the collected samples into the canister

Reliance on the basic design for remotely closing the opening
by means of the extendable closing plug
Securing of stable pressure resistance by reviewing thestructure
Provision of the function to determine the position andfastening state by improving the hole closing program©International Research Institute for Nuclear Decommissioning403

(b) Development of sampling technology
Specifications of the sampling system (2/2)◼ The main specifications of the sampling system are given in Table 1, and photos taken when the functional
tests were conducted by the manufacturer are shown in Fig. 1.
Table 1 Basic specifications of sampling system
Items

Specifications

External dimensions

L1.420xW1, 206xH3,604 mm

Mass

Approx. 1,600kg

Power supply

3Φ200V, 50Hz, 50A

Drive systemMotor
specifications

Verticaldrive

StrokeRotational
drive

Motor
specifications

Making an opening

Sampling

Ball screw vertical drive
Output / torque: 0.75kW / 20Nm
Vertical speed: up to 25 mm/s
1,900mm
Sampling depth equivalent to maximum 200
mm**SARRY standards
Output / torque: 5kW / 300Nm
Rotational speed: up to 450 rpm

Hole making rod, drill bit (60, 50, 40mm)
Cut pieces suction function
Cutting lubricant spraying function
Endoscope rod, laser range finder, endoscope
sampling rod, sampling head

Closing the opening

Perforated part cleaning rod, cleaning buff
Closing plug rod, extendable closing plug

Data logger

Recording and verification of the status ofmovement
Recording of the motor load

Fig. 1 External appearance of the sampling system
during the functional test
©International Research Institute for Nuclear Decommissioning404

(b) Development of sampling technology
Sampling plan for verification of sampling technology◼ Objective
To plan sampling from actual adsorption towers, etc. for verifying the samplingtechnology that uses the sampling system.

◼ Overview
Sampling tests (cold testing and hot testing) were conducted using actual
adsorption tower through which contaminated water has not passed (cold) and aused adsorption tower (hot) for verifying the sampling technology.
The following items were investigated for conducting cold and hot testing.
⚫ Selection of the target adsorption tower⚫ Selection of the sampling work area
⚫ Details of work procedures⚫ Handling of collected samples⚫ Calculation of time required for sample collection⚫ Investigation of overall schedule⚫ Estimation of total exposure dose

©International Research Institute for Nuclear Decommissioning405

(b) Development of sampling technology
Selection of the target adsorption tower (1/3)◼ Adsorbent to be sampled from KURION⚫ KURION system configuration is shown in Fig. 1. (Number of systems and number ofSkidadsorption towers have been reduced)
SMZ
skid

Cs removal
operation

Cs / Sr removal
operation

Stagnant
water

Stagnant
water

SMZ
or
Silica sand

Silica
sand

H
or
EH

H skid

AGH
skid

H

AGH
or
H

TSG

H
or
EH

Box type container for storing
the adsorption tower
Adsorbent abbreviations
SMZ: Surface Modified Zeolite
H, EH: Chabazite based zeolite
AGH: Silver added zeolite
TSG: Granular titanosilicate

Treatedwater

Treatedwater

Fig. 1 KURION system configuration• Silica sand has been used as the adsorbent in SMZ skid on an ongoing basis except in the initial stage.
• H and EH have similar properties. Note that, after the H and EH adsorption tower is used in the AGH skid, it is
moved to the H skid and used.
• TSG, which is granular titanosilicate, is a mineral and its properties are likely to be different than otheradsorbents.
• It appears that the AGH adsorbent used for I-131 has adsorbed I-129 which is an important nuclide intreatment and disposal. Hence the facts pertaining to its collection need to be ascertained.

⚫ Silica sand, H, TSG and AGH are sampled from KURION during the samplingtechnology verification test.
©International Research Institute for Nuclear Decommissioning406

(b) Development of sampling technology
Selection of the target adsorption tower (2/3)◼ Types of SARRY adsorption towers and sampling target⚫ There are various types of SARRY adsorption towers as indicated in Fig. 1 based on the manufacturer,
date of manufacturing, and application.
⚫ In this research, Cs adsorption towers pertaining to which information has been obtained, areinvestigated.
⚫ The lead particle type adsorption towers are excluded as there are uncertainties such as sticking,deformation, etc. of lead due to long term storage. Classification based on the shielding
structure at the adsorbent filling port
SARRY adsorption
towersCs adsorption
towersSimultaneousadsorption towers

Flange type

Shielding by laminating stainless steel flange on
the closing flange of the filling port

Lead particle
type

Shielding by filling lead particles in the space above
the closing flange of the filling port

TYPE-A

Same structure as Cs adsorption
towers (lead particle type)

TYPE- B

TYPE-B1
TYPE-B2
TYPE- B3

Fig. 1 Types of SARRY adsorption towers⚫

The SARRY adsorption towers from which sampling is carried out during the verification test is
selected from the flange type adsorption towers.
©International Research Institute for Nuclear Decommissioning407

(b) Development of sampling technology
Selection of the target adsorption tower (3/3)◼ Adsorbent to be sampled from SARRY⚫ SARRY system configuration during the period when flange type adsorption towers were used, isshown in Fig. 1.
(Number of systems and number of adsorption towers have been reduced)

FilterStagnant
water

Coarsesand

Adsorption tower

Fine
sand

IE96

IE911

Media
filterAdsorbent abbreviations
IE96: Chabazite based
zeoliteIE911: Silico titanate

Treatedwater

Fig. 1 SARRY system configuration when flange type adsorption towers were used⚫

There are 4 types of adsorbents used in the SARRY flange type adsorption towers, namely, coarse sand, fine sand,
IE96, and IE911.
(The shape of the media filter is different than those used in other SARRY adsorption towers. Media filters did not have
any track record as far as it was checked. Hence these filters have been excluded from the very beginning.)⚫

A maximum of 4 adsorption towers from those containing coarse sand, fine sand, IE96 and IE911 aresampled from SARRY.

©International Research Institute for Nuclear Decommissioning

408

(b) Development of sampling technology
Selection of the sampling work area (1/2)◼ Conditions for selecting the sampling work area
⚫ Availability of the required lifting equipment- Maximum crane load: Should be able to lift the KURION and SARRY adsorption towers.
- Crane lifting height: KURION 7.9 m or more, SARRY 8.0 m or more
⚫ The impact of risks of natural disasters such as typhoon, etc. should be sufficiently small.
◼ Contents of investigations⚫ The following work locations have been selected based on the condition of availability of required lifting
equipment.
Working Machine Building (Indoor work)
High Performance ALPS Building (Indoor work)
Second temporary storage facility (Outdoor work)⚫ Due to the condition that the impact of risks of natural disasters such as typhoon, etc. should be
sufficiently small, the second temporary storage facility where work would be carried out outdoors was
excluded.
⚫ Considering the efficiency of on-site work, the Working Machine Building was excluded.

◼ Investigation results⚫ It was evaluated that the High Performance ALPS Building would be the most suitable as the
sampling work area.
⚫ Dedicated working stands are built inside the building and the sampling technology is verified foradsorption towers of both KURION and SARRY.

©International Research Institute for Nuclear Decommissioning409

(b) Development of sampling technology
Selection of the sampling work area (2/2)◼

Building the work area in the High Performance ALPS Building ◼

➢ WS and maintenance stand (hereinafter, MS) are installed at the back side of
the truck bay entrance.
➢ A generator and compressor are installed outdoors.
➢ After installation from adsorption tower to ISM, the ventilation unit isconnected to the ISM.
➢ Thereafter, a house is installed on the WS.

Remote control panel / PC

Generator /Compressor
Gantry craneHigh performance
ALPS➢ The layout diagram of the main facilities related to sampling work is shown in
Fig. 2.
➢ The adsorption towers are temporarily placed in the second temporarystorage facility, or are temporarily stored in the temporary storage facility.
➢ In the following circumstances, the required measures are taken in advanceat the second temporary storage facility.
• If the adsorption towers need to be decontaminated or covered.
• If the adsorption towers need to be rotated 90° so that they face in the
direction in which WS has been installed.
➢ While sampling, the adsorption towers are transported to the High
Performance ALPS Building one at a time.
➢ Collected samples are transported on-site to the solid radioactive waste
storage facility.
➢ After sampling the adsorption towers are planned to be returned to theoriginal storage facility.
➢ On-site investigation of the second temporary storage facility is carried out.

Solid waste
storage facility

Second temporarystorage facility

Seismic safety building

7.1m

House

Facilities related to sampling work and details of
implementation

Temporary storage facility(Fourth facility)

WS
MS

Ventilation
unit

High Performance ALPS Building(Sampling work area)

PN
Fig. 1 Image illustrating construction of the work
area in the High Performance ALPS Building

Fig. 2 The layout of the main facilities related to sampling work
©International Research Institute for Nuclear Decommissioning410

(b) Development of sampling technology
Details of work procedures (1/2)◼ Main work procedures when the sampling system is used
1. Measuring the distance of the surface where theopening is to be made (adsorption tower top board
location) by means of a laser range finder [Remote
operation]

2. Deep hole drilling using a Φ60 drill bit

Laser range
finderΦ60 drill bit

4. Drilling a through hole using a Φ50 drill bit

5. Measuring the distance up to the welded bottom plate with thehelp of the laser range finder and endoscope, and checking theinside of the adsorption tower. [Remote operation]

3. Measuring the depth of the Φ60 perforated holeusing a laser range finder and checking the innersurface of the perforated hole [Remote operation]

Laser range
finderEndoscope

6. Drilling a through hole using a Φ40 drill bit*Not possible in the case of KURION

Φ50

Φ50

Φ5050 drillbitΦ40 drill bit
Laser range
finderEndoscope©International Research Institute for Nuclear Decommissioning

411

(b) Development of sampling technology
Details of work procedures (2/2)◼ Main work procedures when the sampling system is used

11. Cleaning the perforated hole and checking theinner surface of the perforated hole.

Cleaning buff
Canister

Endoscope*The buff is replaced through the glove port
*Tools are replaced by opening the panel after ventilation

H
E
P
A

Blower

10. Removing the canister in which the samplinghead has been placed.

Isolating the air
insidethe adsorption
towerwith the shielding
lid

Dust monitor

EndoscopeSampling
head

Laser range
finder

9. Measuring the radioactive dust concentration
inside ISM.

Blower

8. Inserting the sampling head, and collecting samples.
Placing the sampling head into the canister after collecting
samples.

Dustmonitor7. Measuring the distance up to the surface of the adsorbent withthe help of the laser range finder and endoscope, and checking theinside of the adsorption tower [Remote operation]

12. Putting in the closing plug.

Closing plug

*Tools are replaced by opening the panel after
ventilation©International Research Institute for Nuclear Decommissioning

412

(b) Development of sampling technology
Handling of collected samples⚫
⚫

An image of the canister in which the sampling head (hereinafter, SH) is placed, is shown in Fig. 1.
The flow from collecting the samples to analysis is shown in Fig. 2.

Dosimeter

On-site transport container

Scale① The dose rate around the
SH after sampling ismeasured remotely.

Solid radioactive
waste storage facility

Characterization

Dosimeter

Dosimeter

Canister

Dosimeter

Development of sampling technology

Fig. 1 Image of the canister

ISM
Rod
SH

② SH is inserted into thecanister remotely. Aftermeasuring the dose rate,
the rod and SH are
separated manually.

Transport cask

③ The canister is removed and its
mass is measured. The massof the canister and SH
measured earlier is subtracted
and the mass of the samples iscalculated.

Analysis facility
iron cell

Additional shielding
Canister

Fragmentation/ Cs removal④ The canister is placed in theon-site transport cask
(shielded), it is transported onsite in a vehicle and handed
over for characterization.

Analysis facility
GB or hoodSampleAnalyzer⑤ The samples are stored in thedesignated location andadditional shielding is installed.

⑥ Each canister is placed in thetransport cask and transported
to the analysis facility⑦ The SH and samples are removed from the canister⑦ After carrying out pre-treatment in
inside the iron cell of the analysis facility, andaccordance with details of analysis byfragmentation / Cs removal treatment, etc. is carried outmeans of GB, etc., nuclide analysis iscarried out.

Fig. 2 The flow from collecting the samples to analysis (Image)©International Research Institute for Nuclear Decommissioning

413

(b) Development of sampling technology
Calculation of the time required for sample collection
Table 1 Work procedures and expected number of days
Expected number of days
No.

Work contents
KURION

SARRY

Sub-total①

Preparation for sampling

1 day

②

Adsorption tower transportation and installation

1 day

③

Series of tasks using the sampling system

Total 244 minutes

Total 677 minutes•

Measurement of the distance of the surface where the opening is
to be made (adsorption tower top board location)•

Φ60 deep hole drilling•

Measurement of the depth of the Φ60 perforated hole andchecking of the inner surface of the perforated hole

19 minutes•

Φ50 through hole

44 minutes•

Measurement of the distance up to the welded bottom andchecking of the inner surface of the perforated hole－

19 minutes•

Φ40 through hole－

44 minutes•

Measurement of the distance up to the surface of the adsorbent
and checking of the inside of the adsorption tower

19 minutes•

Sampling, insertion of the sampling head into the canister

11 minutes•

Measurement of the radioactive dust concentration inside ISM

30 minutes•

Removal of the canister containing the sampling head

3 minutes•

Cleaning of the perforated hole and checking of the inner surfaceof the perforated hole

41 minutes•

Placement of the closing plug

6 minutes•

Checking of the status of placement of the closing plug

16 minutes

1 minutes
54 minutes

424 minutes

1 day *
*The functioning of the system can be arbitrarily
discontinued and resumed, but the series of tasks
should be completed within the same day by working
in 3 shifts so that the adsorption towers are not left
unattended at any time without closing their openings.

④

Sampling system removal / tidying up

1 day

⑤

Adsorption tower transfer

1 day

⑥

Sampling system maintenance / decontamination / transportation ofsamples

1 day⑦

Extra days

1 day

⚫

Collecting samples from 1 adsorption tower requires 1 day only for the series of tasks and 7 days including preparation and tidying up.
©International Research Institute for Nuclear Decommissioning414

(b) Development of sampling technology
Investigation of overall schedule◼

Reflection of work procedures and expected number of days⚫
According to the results of the element tests and the investigations on work procedures, it was calculated that the series of tasks using the sampling system
would require 224 minutes (approx. 5 hours) in the case of KURION, and 677 minutes (12 hours) in the case of SARRY.
⚫
However, the plan is to complete the series of tasks within the same day so that work is not discontinued while the opening in the adsorption tower is yet to
be closed.
⚫
It was calculated that it would take 7 days to collect samples from 1 adsorption tower considering transportation of the adsorption tower and installation of thesampling system.

◼

Implementation items⚫
Along with arranging for the materials and equipment, equipment testing and trainings for the personnel involved in testing are conducted at the in-housefacilities.
⚫
Before collecting hot samples, the functioning of the equipment is verified and training is conducted by carrying out cold testing using adsorption towers
through which contaminated water has not passed.
Preconditions of the overall schedule⚫
It is assumed that the time and work area required for testing is available.
⚫
It is assumed that there would be no additional regulatory requirements besides those that are anticipated at present.
⚫
It is assumed that sampling is carried out from a maximum of 8 adsorption towers. Although it takes 7 days to collect samples from 1 adsorption tower, anevaluation period is set between every 2 collections.
⚫
The proposed overall schedule for collecting samples from 4 adsorption towers each from KURION and SARRY, is shown in Table 1.

◼

Table 1 Proposed overall schedule
Elapsed number of months

Main work steps
1-5

6

7

8

9

10

11

12

Preparation of materials andequipment
・ Procurement andmanufacturing of work materials
and equipment
・Installation of the working stand
SARRY
・Cold mock-up

SARRY 1 adsorption tower / 7 days x 4 rounds

・Sampling (Round 1 to 4)
KURION・Cold mock-up
・Sampling (Round 1 to 4)
・Sample storage and tidyingup

KURION 1 adsorption tower / 7 days x 4 rounds©International Research Institute for Nuclear Decommissioning

415

(b) Development of sampling technology
Estimation of total exposure dose (1/3)
1. Method of calculating the total exposure dose
Table 1 Target dose rate and calculation method
Dose contribution
Impact of work area

Target dose rate

Calculation method

BG value of work building

Set based on survey data

Around the adsorption tower while preparing
for work

Impact ofadsorption tower

Around the adsorption tower after removing
the shielding lid and while installing WD (Only
KURION)Calculated using thecomputing code
“QAD-CGGP2R”

Around the adsorption tower after installingWD
Around the adsorption tower while installingthe sampling system (ISM)
Impact of collected
samples

Surface of canister and the area in its vicinity
Area in the vicinity of sampling headFig. 1 Sample collection work location(High performance ALPS Building)⚫ The number of workers assigned and the duration of work is set based on theinvestigations on schedule and work steps, and the results of the element
tests⚫ The radiation source, etc. that is likely to affect work exposure is set, and theatmosphere dose rate is calculated according to the work steps (Table 1,
Table 1 on next page)⚫ The work classification is set depending on the contents of the target work
(Table 2), and accordingly the work coefficient is set⚫ The exposure dose in each work step is calculated with the following formulaand aggregatedWork exposure dose (person/mSv)= Work coefficientx Number of workers (persons)x Duration of work (h)x Atmosphere dose rate (mSv/h)

Table 2 Work classification
Work
classification

Target Work

Low

Work that requires multitude of trips outside the work area (adsorption towertransportation, preparatory work, etc.)
Work during which it is possible to retreat to a low radiation area (dustmeasurement, etc.)

MediumHigh

Work that requires constant presence in the work area
Short duration work while operating the sampling system (ISM)

Work of manually operating the sampling system (ISM)
Work for which the time set is as short as several minutes and hence isconsidered as a margin©International Research Institute for Nuclear Decommissioning416

(b) Development of sampling technology
Estimation of total exposure dose (2/3)
2. Estimation of atmosphere dose rate⚫

The work steps for which the atmosphere dose rate is calculated are indicated inTable 1, and an example of the dose rate distribution based on the results ofanalysis is shown in Fig. 1.

Inside the High Performance ALPS Building

2

Around the adsorption tower

3

While installing WD

4While installing ISM

5Work on the working stand such as tool change, etc.

6

Inside ISM

7

While inserting sampling head into canister

8

Canister handling inside ISM

9Removing the canister10

Before transporting canister

ISM slinging
operation
location

Adsorption
tower

1

Surrounding work
location

Distance

Case 4: While installing ISM

ISM operator
work location

Adsorption
towerWork step

HeightCase

HeightTable 1 Work steps for which the atmosphere dose rate will be calculated

Distance

Case 5: Work on the working stand such as tool change, etc.
Fig. 1 Example of the image illustrating dose rate distribution
based on the SARRY dose analysis results©International Research Institute for Nuclear Decommissioning

417

(b) Development of sampling technology
Estimation of total exposure dose (3/3)◼ Estimation of total exposure dose and evaluation of feasibility of work➢•

Setting the allowable total exposure dose
The number of work teams and the number of members in each team are set based on the work procedures and thecontent of work•
Considering that the work is assigned in rotation, it is assumed that all workers are uniformly exposed.
•
Assuming that the workers are involved in other tasks as well during the year, the upper exposure limit per person for thistask is set.
⇒ The allowable total exposure dose for this task is set based on the upper exposure limit per person and the total number ofworkers.
➢ Calculating the total exposure dose evaluation value•
The radiation for evaluating the exposure is set considering the radiation from the adsorption tower mentioned in theimplementation plan and the attenuation due to time elapsed up to the day on which exposure is evaluated.
•
It is assumed that the 4 adsorption towers from SARRY and KURION have the same radiation.
⇒ The total exposure dose for this task which includes preparatory work, sampling from each adsorption tower and thework of tidying up, is evaluated.

➢ Evaluation results
The following results were obtained.
Allowable exposure dose > Total exposure dose evaluation value
It was verified that the task is feasible on site.
➢ Issues• Re-evaluation will be carried out based on the latest dose rate of the adsorption tower fromwhich samples are collected, in order to improve accuracy.
• Since working on the KURION adsorption towers involves high level of exposure,enhancement of shielding performance will continue to be investigated.
©International Research Institute for Nuclear Decommissioning

418

(b) Development of sampling technology
Summary◼ Current results➢ The sampling system was designed and manufactured based on the results of investigations directed
towards hot testing and results of the element tests during the FY2019-2020 research, and functional tests
were completed.

➢ The incidental facilities required for the series of tests using the sampling system were designed andmanufactured.
➢ Site investigation of the facilities was carried out in relation to verifying sampling technology, and therequired peripheral equipment such as ventilation unit, etc. were investigated and designed.
➢ Considering the designed peripheral equipment, the sampling plan for verifying the sampling technology
was drafted.
➢ The dose around the adsorption towers under certain radiation conditions was evaluated, and the total
exposure dose was estimated.

◼ Issues, etc.
➢ Manufacturing and functional verification of peripheral equipment such as ventilation unit, etc. required forverifying the sampling technology➢ Verification of the layout of cameras, etc. in accordance with the on-site environment for ensuring ease of
remote operation➢ Investigation of the arrangement of the dosimeter for monitoring changes in air dose rate and of themethods of evaluating the readings➢ Dose evaluation considering the conditions of the actual adsorbent to be sampled as the model conditions➢ Consideration of additional shielding, etc. based on the results of dose evaluation for further reducing
exposure©International Research Institute for Nuclear Decommissioning419

2. Project details (No. 4)
d. Integration of R&D results*Members of IRID in charge of r
©International Research Institute for Nuclear Decommissioning420

(d) Integration of R&D results◼ Achievements up to FY2020
 By FY2020, a system was developed for comprehensively managing
and evaluating the direction of research and development related totreatment and disposal of solid waste, its progress, results andpending issues (Fig. 1).
◼ Goal To put forth view points for selecting waste streams when multiple
options are left to choose from, and to provide the indicators formaking a decision.
 To comprehensively summarize the research results including resultsfrom previous years.

Input

Contents of investigations
Consolidation of the required
information and the issues

Results of research &development (Inventory, etc.)(Treatment flow withmany options)

Input consolidation
chart by waste(Waste list)

Information required forinvestigating the stream ateach stage, and theissues◼

Details of implementation (Details are provided from next pageonwards)
 Investigation of view points for narrowing down waste streams when
multiple streams are left to choose from, based on the waste stream
investigated up to FY2020, and of the indicators for making adecision.
 Conducting of case studies using specific wastes as examples, withrespect to the viewpoints for narrowing down the investigated waste
stream options and the indicators for making a decision, andconsolidation of issues such as too much or too little information thatis required while selecting the option.

◼ Indicators for the achievements of goal The viewpoints for narrowing down the options and the indicators formaking a decision are provided.
 Case studies related to narrowing down options with respect torepresentative wastes are conducted, and issues in narrowing down
the options are consolidated (Example: Information required forselection, etc.).
(Target at the completion: TRL： 4→4）

(Waste management
sheet)

Results of research &development (treatment, etc.)

Setting of the time period forinvestigations at each step bywaste
Temporal axis
organization chart

Output1. Proposed waste streams in which
treatment flows that have been presented
at the current point in time have been
integrated
2. Issues in drafting waste streams

Matters and contents to beinvestigated, status ofinitiatives, and required
actions

Time period required forinvestigations at each stage
by waste(Estimated)
The current proposed treatment
flow and the proposed integrated
waste streams

Mid-and-Long-termRoad-map

Verification of the overview ofwaste management

TEPCO HD
Storage management
plan, etc.

(Narrowed down
treatment flow)

Waste stream(Dashboard)

Fig. 1 Details of investigations conducted up to FY2020
and their output(Flow of integration of research & development results)©International Research Institute for Nuclear Decommissioning421

(d) Integration of R&D results
Role of “Integration of R&D results” and its correspondence to the contents of investigations⚫

Purpose of investigating “Integration of R&D results”
To consolidate the flow (= waste stream) from generation of waste to storage, treatment such as
segregation, volume reduction, stabilization, etc. and then long term storage or disposal based on theuncertain character of information (limited information) related to the properties of solid waste generated
in 1F.
And in doing so, to carry out research & development rationally and efficiently by getting an overall
bird’s-eye view, and in addition by mutually providing information during every stage of the waste stream
as feedback.

Matters to be investigated aiming at “Integration of R&D results” and the output goals
No.

Items to be investigatedOutput goals (Example)

1

Presentation of the direction of the research & development
directed towards treatment and disposal of waste and thegoals

Provision of a mechanism (example: waste stream) for sharing with theconcerned parties, who finally utilize the research & development
results.

2

Provision of a mechanism for investigating the sufficiency of
the R&D topics concerning treatment and disposal of waste

Provision of a mechanism (example: input consolidation chart) for
sharing and making visible the domains of research & development at
each stage to the concerned parties.

3

Provision of a mechanism for sharing and making holisticallyvisible the status of progress of the research & development
topicsProvision of a mechanism (example: temporal axis organization chart)
for sharing and making visible the progress of research & development
and the pending issues to the concerned parties.

4

Provision of items that need to be considered while selecting
the treatment and disposal options

Consolidation of information for making decisions (perspective fornarrowing down the options and the indicators for making decisions)when multiple waste stream options are left to choose from
©International Research Institute for Nuclear Decommissioning422

(d) Integration of R&D results
Overview of the investigations
No. 1
Reflection of progress of
各研究進捗の反映
each research

Update更新

Gathering of information /情報収集・リスト化
Listing up[Legend]【凡例】
Investigation items：検討項目：インプット情報等
Input information, etc.
XXX ：検討のアウトプット
Output of investigations

Waste list
廃棄物リストNo. 1
Waste
management
廃棄物管理シート
sheet

Waste
廃棄物の分類classification

Reflection of progress of each
各研究進捗・
researchand storage
management plan, etc.
保管管理計画等の反映

The section in which
representative waste is selected,
and investigations continue to be
carried out at each phase of thewaste stream

Update更新

No. 1
Investigation oftreatment flow

処理フロー検討

Waste stream
廃棄物ストリーム
The section in whichinvestigations will be
conducted this year

No. 2
Interface between the research &発生～処分までの各研究開発のdevelopment from generation to disposal
インターフェース／検討課題の統合/ integration of investigation results

インプット整理表
Input
consolidation chart

No. 3

No. 4

Integration of the status of progress of
廃棄物ストリームと照らした
research& development in accordancewith the waste stream
研究開発の進捗状況の統合

Investigation of information for
複数の選択肢が残る際のmakingdecisions when multiple
判断材料の検討
options
are left to choose from

Temporal axis
時間軸整理表organization
chart

Annualresults
Annualresults（未）

No. X: Items corresponding to the items to be investigated
and the output goals©International Research Institute for Nuclear Decommissioning423

(d) Integration of R&D results
Contents of investigations and procedures
Investigations were conducted in accordance with the following procedures directed towards “Research Goal 1 of the research: To
put forth view points for selecting waste streams when multiple options are left to choose from, and to provide the indicators formaking a decision” and “Research Goal 2 of the research: To comprehensively summarize the research results”

Investigations
pertaining to
Research
Goal 1

Investigations
pertaining to
Research
Goal 2①

Investigation of the indicators for narrowing down options[Overview]
The IAEA safety standards are investigated, and items that must be considered while
selecting and narrowing down waste stream options are investigated.

②

Investigation of information required for the task of narrowing down[Overview]
The information required while carrying out the task of selection and narrowing down isinvestigated along with its source and the conditions under which it is acquired (technicaldevelopment phase, etc.) with respect to the indicators that have been set during theinvestigations in ① above.

③

Narrowing down on a trial basis using the representative waste (Case study)[Overview]
Based on ①/② mentioned above, the idea behind narrowing down the representative
wastes (ALPS) and processes (prospective technologies to be used for waste conditioning),
using information that can be gathered at the current point in time, is illustrated with
examples.

④

Investigation of the procedures for integrating research results[Overview]
In addition to reorganizing the correlation between the research results to be integrated,
including the results from previous years, updated procedures that use those results andwhich are in accordance with the procedures for setting research & development goals
pertaining to treatment and disposal of solid waste, and the progress of the research &development, are investigated.
©International Research Institute for Nuclear Decommissioning424

(d) Integration of R&D results
Contents of investigations: ① Investigation of the indicators for narrowing down
options (1/3)[Details of examination]➢The indicators for selection and narrowing down when multiple waste stream options are left to choose from were
investigated from the viewpoint of “safe”, “reliable”, “rational”, “quick”, “field oriented”, while referring to the Five
Guiding Principles for reducing risks indicated in the “Technical Strategic Plan[1]”.
➢From the viewpoint of “safe”, the IAEA safety standards “Management of radioactive waste before disposal” (GSRPart 5) and related separate safety requirements were investigated, and were reflected while setting the indicators.
➢With regards to “reliable”, “rational”, “quick” and “field oriented”, the indicators of each viewpoint were investigatedwhile referring to the “Technical Strategic Plan” and the results of brainstorming implemented while investigating
options for treatment and disposal of waste [2].
廃棄物ストリーム上のWaste stream
選択肢
options

Compliance「安全」へのwith“safe”
適合性[1]: FY2021 Technical Strategic Plan for Decommissioning of the
Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power
Company Holdings, Inc. (October 29, 2021, Nuclear Damage
Compensation and Decommissioning Facilitation Corporation)[2]: Subsidy Project of Decommissioning and Contaminated Water
Management in the FY2014-15 Supplementary Budget - R&D for
Treatment and Disposal of Solid Wastes

Optionthat does not
適合しない選択肢
comply※※※[Reliable]「確実」[Rational]
「合理的」

※

[Quick]「迅速」[Field「現場
oriented]
指向」

Option that does not
適合しない選択肢
comply<Concept of selecting and narrowing
down options*>

The optionthat is found to be applicable
適用性が認められる選択肢（複数の選択肢が残る可能性も有り）(Multiple options may be left)*Even though compliance with “safe” is a precondition, since the level of priority of the other 4
viewpoints is likely to change depending upon theobject of investigation, they were ranked the same.
©International Research Institute for Nuclear Decommissioning425

(d) Integration of R&D results
Contents of investigations: ① Investigation of the indicators for narrowing down
options (2/3)[Investigation results: Output]➢The “indicators for narrowing down”, “determining factors”, “information required while narrowing down”, and “proposed
methods for narrowing down” were consolidated with respect to 5 viewpoints (Safe, reliable, rational, quick, field-oriented).
“Indicators for narrowing down”:
絞り込み指標Results of investigations選択肢の絞り込みの観点
based on
IAEA safety standards, etc. were
適用実績
consolidated.
B1.選択肢の技術成熟度（技術の適用性）
選択肢の絞り込みの観点

判断要素

B2.明確な判断基準の有無（≒要求事項への適合性判断）―
・適用実績等を踏まえた、選択肢毎のTRL設定結果―
判断時に考慮必要と考える事項（（）内：補足事項）

絞り込み時に必要とする情報―（各廃棄物等からの寄与の合計値として）（処理対象物の性状変動への対応程 適用する技術の受入れ条件（適用可能範囲）

B

選択肢間の技術成熟度（実績、残る検討課題等）について比較評価

絞り込み

処理対象物のインベントリデータ

―

パターン

絞り込みの方法案(凡例は本保管或いは埋設施設の受入れ基準（線量率、強度、許容され（保管・埋設施設の受入れ基準、輸送基準） る核種・化学的性質等）

1mSv/年未満

敷地境界線量への影響

絞り込みの方法案

(凡例は本
文参照）

処理結果の適切性を判断する定量的な基準

B3.柔軟性パターン

絞り込み時に必要とする情報（（）内：補足事項）

研究開発の進捗

絞り込み指標確実

絞り込み
判断時に考慮必要と考える事項

判断要素

文参照）

―

―
（選択肢に依らず確認すべき事項として抽出し、絞り込み方法の検討対象より除外）
敷地境界へ与える線量影響評価の上、サイト内で発生する廃棄物からの総影響が基準値未満であること
廃棄物の性状変動リスクを踏まえ、選択肢毎の適用
B・選択肢毎の適用条件（適用可能範囲を示す情報）を原則として、敷地境界に与える影響がより低い選B
条件により、処理対象物の性状への対応可能性の観・選択肢を適用する廃棄物の性状変動リスク
択肢を選定
点で比較評価度）“Determining factors”: Items
that must be
処理作業に係る作業員被ばく影響評価の上、年間被A1.放射線防護・処理対象物の線量データ“Information required
while narrowing down”:―
considered while
determining basedon the 適用する技術の開発コスト 年間被ばく線量限度
処理等の作業員被ばく影響B
ばく線量限度を下回ることを原則として、より被ば・作業計画線量
選択肢の適用に掛かるコスト評価値
選択肢の開発コスト・設備コスト・ランニングコス
く影響が低い選択肢を選定
The information believed
to be necessary while
適用する技術に用いる設備コスト―
B
indicators for narrowing down were broken down
・処理作業の内容（作業中における飛散リスクの程度）
処理作業の内容、或いは生じる生成物が基準値（作（開発コスト／設備コスト／ランニングコスト）
トの評価結果による比較評価
エリア区分毎の表面汚染密度および空気中のevaluating
the
task
of
narrowing
down
was
C1.選択肢の適用に掛かるコスト
汚染拡大防止・処理後の状態（後段プロセスへ引渡す際の拡散リスクの程
A、B
業エリア毎の管理値等）を逸脱するリスクを比較
and set.
技術適用時のランニングコスト―
放射性物質濃度の基準値
合理的度） set for each narrowing downし、汚染拡大リスクが低い選択肢を選定
investigated and
後段プロセスへのコスト影響評価結果を用いた比較(Example: Radiation protection
適用した技術を適用した際の後段プロセスへのコスト影響―
保管或いは埋設施設へのコスト影響
B
選択した処理を適用した際の物量（保管・埋設容量）
ー
処理後の廃棄物のレベル区分別物量の評価を基に、indicator
anddetermining
factor.
評価⇒ Impact on site boundary dose (impact on the
・処理対象物の物量
以下の2点の評価実施
選択肢の適用範囲（廃棄物性状、インベントリ等）
C2.選択肢の汎用性―
選択肢の技術的な汎用性に関する情報（適用対象廃棄物）
B
選択した処理を適用した際の二次廃棄物物量・処理により発生する物量（評価値）
A、B
・後段プロセス（保管/埋設）の受入れ容量への適合
general public),A2.発生量の抑制
impact on worker
exposure,選択した技術の適用範囲（適用可能な廃棄物） ーの検討結果を基にした比較評価
・後段プロセスの受入れ容量（例：埋設施設の容量）性ー・処理技術の技術的成熟度（TRL.適用実績）
preventing spread of contamination)
技術の適用実績―
後段プロセス（保管・埋設等）の受入れ容量への適合性
・選択肢間での発生物量の相対比較
D1.選択肢の適用に対する理解取得に
目的達成可能であることを前提に、候補となる選択（保管・埋設計画がある場合は施設容量）

供する時間（社会的受容性）

保管管理計画上で示される廃棄物に関して安全性（処理・保管・埋設の各段階）に関する説明性

保管施設の受入れ基準への適合性（例：表面線量率など）技術開発に要する時間

安全・技術適用時のリスク評価結果（放射線防護を中心とした安保管施設の受入れ基準―

全の観点でのリスク）

B

肢の成熟度やリスクを基にした比較評価

事前ラボ試験データ等により入手した生成される廃A
棄物／廃棄体の特性情報と、後段プロセスの受入れ“Proposed
methods
for narrowing
down”: The選択肢間の開発期間を含めた適用までに掛かる期間
methods for evaluating the options―
迅速
選択肢の適用までに掛かる工程情報
る特性情報の取得―
埋設施設の受入れ基準
埋設施設の受入れ基準（WAC)への適合性
D2.選択肢の適用に要する時間
B（現状無し）（後段プロセスが「処分」の場合）（開発項目、設計検討項目等）の評価値による比較評価
wereclassified
into
the following基準の照合により適用可否を評価
2 patterns, andthenproposed methods for
設備設計・製作を含む現地適用までに要する時間―
選択した処理を適用する際の、処理作業中の廃棄物の安定性―
quantificationwhile
carrying A、B
out 事前ラボ試験データ等により技術適用時に影響する
comparative evaluation (Pattern B) were presented (*)
D3.適用から目的達成（処理完了等）
選択肢の処理速度に関する情報処理に伴う状態変化リスク（事前ラボ試験データ等）
事象（例：著しい膨潤、発熱等）の有無を確認する
適用する技術の処理速度―
B
適用する技術の処理速度の想定を用いた比較評価（著しい状態変化の有無）（現状無し）
までに要する時間
と共に、リスクの程度により比較評価<Classification> （処理プロセス、廃棄体1体当りの製作期間等）
エリア制約を満足した上で選択肢の適用が可能であ―A4.閉じ込め性を有する形態
選択した処理を適用した後の状態での廃棄物の性状・Pattern
A: The criteria,
etc. to be followed areclear, and the applicability of the（定量評価時の活用指標
処理後の形態に対する物理的性状の情報
作業エリア・敷地の確保―
選択肢を適用するエリア（設備設置箇所）の制約情報
A、B ることを前提に、選択肢適用に当たっての課題の程（≒安定性／拡散のし難さ）（物理的性状）処理後に生成される形態に対し、SED指標（Form：SED指標におけるForm Factor） option can be determined by checking against thecriteria.
度による比較評価
B
Factor,Control Factor）を適用し、安定性・拡散性に
現場指向
E1.1F現場環境での成立性―
エリア制約を満足した上で選択肢の適用が可能であ選択した処理を適用した後の状態での廃棄物の性状ついて選択肢間で比較評価
・Pattern
B: The criteria,
etc. to befollowed are notclear, and the superiority or・選択肢を適用する形態（装置・システム等）（定量評価時の活用指標
処理後の形態に対する化学的性状の情報（化学的性状）作業性（環境、アクセス性、操作性）―
B
ることを前提に、選択肢を適用する際の形態・作業：SED指標におけるControl Factor） inferiority of the prospective
・選択肢を適用する際の作業内容
options is determinedand
set through comparative
内容を基に運用性の観点で比較評価
・処理後の取扱い形態
・後段プロセス（保管/埋設）での受入れ容器が確定―
evaluation.
後段プロセスへの受渡し形態（バウンダリ構成物）
・後段プロセスからの取扱い形態（バウンダリ構成物）へのしている場合は、当該容器を用いた処理作業の成立

A3.後段プロセスの受入れ判断に資すは、対象の保管施設の受入れ条件・設計条件 （後段プロセスが「保管」の場合）（現状無し）

要求（例：保管期間）

A5.閉じ込め性を維持する措置

A、B

性により適用性を評価

・処理後の取扱い形態（バウンダリ構成物）の設計仕様（耐・後段プロセス（保管/埋設）への引渡し形態が複数*Example)
Technology Readiness
Level ⇒ Options are quantified by setting the TRL
用年数等）※
存在する場合は、SED指標（Facility Descriptor）を（定量評価時の活用指標※Facility
Descriptorの判定要素に係る情報
thencomparative evaluation適用し、閉じ込め性について選択肢間で比較評価
is carried out.
：SED指標におけるFacility Descriptor） and―バウンダリ構成物の性能

A6.後段プロセスの作業に則した形態 後段プロセスへの引渡し形態（ハンドリング性）（検査・取出し等の後段作業要求を考慮した形態）―（後段プロセスでの取扱い計画／要求）・処理後の取扱い形態

・後段プロセスでの取扱い計画（検査・モニタリング等）―
List of
viewpoints and
後段プロセスでの要求（例：検査、モニタリング等）の考慮（後段プロセスでの取扱い計画／要求）
narrowing down indicatorsfor narrowing down options

A、B

・後段プロセス（保管/埋設）での取扱い計画（検査・モニタリング等）への適用性を評価©International Research Institute for Nuclear Decommissioning426

(d) Integration of R&D results
Contents of investigations: ① Investigation of the indicators for narrowing down
options (3/3)[Investigation of the methods for leveraging the Investigation results]
Step 1: Consolidation of
options

選択肢A
Option A

選択肢B
OptionC

選択肢C
OptionB

本事業での検討結果
絞り込み
選択肢の絞り込みの観点

絞り込み指標判断要素

判断基準の有無

絞り込み時に必要とする情報

パターン

絞り込みの方法案

(凡例は本
文参照）

敷地境界へ与える線量影響評価の上、サイト内で発
1mSv/年未満

敷地境界線量への影響

処理対象物のインベントリデータ（各廃棄物等からの寄与の合計値として）B

生する廃棄物からの総影響が基準値未満であることを原則として、敷地境界に与える影響がより低い選択肢を選定

A1.放射線防護処理等の作業員被ばく影響

年間被ばく線量限度

Step 2: Setting of indicatorsfor narrowing down options(including setting the priority)

Narrowing down
評価指標
indicators

放射性物質濃度の基準値

選択した処理を適用した際の物量（保管・埋設容量）

優先度
Priority

参照
Reference

A2.発生量の抑制

絞り込み指標安全る特性情報の取得

保管施設の受入れ基準への適合性（例：表面線量率など）

適用実績
B1.選択肢の技術成熟度（技術の適用
埋設施設の受入れ基準（WAC)への適合性性）

Narrowing down

安全Safety確実
Reliable

評価指標A
indicator A

１

Narrowing down

Narrowing down
評価指標C
indicator C

A、B

処理後の廃棄物のレベル区分別物量の評価を基に、以下の2点の評価実施
A、B・後段プロセスの受入れ容量（例：埋設施設の容量）処理に伴う状態変化リスク（事前ラボ試験データ等）

A、B

選択した処理を適用した後の状態での廃棄物の性状（化学的性状）考慮した形態）

Setting
設定

４

選択肢の適用に掛かるコスト評価値※Facility Descriptorの判定要素に係る情報供する時間（社会的受容性）
迅速までに要する時間

条件により、処理対象物の性状への対応可能性の観性により適用性を評価

・後段プロセス（保管/埋設）への引渡し形態が複数―

・後段プロセスでの取扱い計画（検査・モニタリング等）―
保管或いは埋設施設へのコスト影響

A、B

・後段プロセス（保管/埋設）での取扱い計画（検後段プロセスへのコスト影響評価結果を用いた比較

査・モニタリング等）への適用性を評価
B

評価

―

選択肢の技術的な汎用性に関する情報（適用対象廃棄物）―

・技術の適用実績（成熟度）

・技術適用時のリスク評価結果（特に放射線防護上のリス―

ク）

技術開発に要する時間―

選択肢の適用までに掛かる工程情報

設備設計・製作を含む現地適用までに要する時間
D3.適用から目的達成（処理完了等）

廃棄物の性状変動リスクを踏まえ、選択肢毎の適用点で比較評価している場合は、当該容器を用いた処理作業の成立

選択肢の開発コスト・設備コスト・ランニングコス
存在する場合は、SED指標（Facility
Descriptor）をB（開発コスト／設備コスト／ランニングコスト）
トの評価結果による比較評価
適用し、閉じ込め性について選択肢間で比較評価―

・処理後の取扱い形態

安全性（処理・保管・埋設の各段階）に関する説明性

D2.選択肢の適用に要する時間

B・後段プロセス（保管/埋設）での受入れ容器が確定（後段プロセスでの取扱い計画／要求）

選択した技術の適用範囲（適用可能な廃棄物）

D1.選択肢の適用に対する理解取得に 技術の適用実績

A、B

・バウンダリ構成物の仕様（耐用年数等）※―（後段プロセスでの取扱い計画／要求）―
適用した技術を適用した際の後段プロセスへのコスト影響

後段プロセスでの要求（例：検査、モニタリング等）の考慮

C2.選択肢の汎用性

ついて選択肢間で比較評価

・選択肢を適用する廃棄物の性状変動リスク

・後段プロセスからのバウンダリへの要求（例：保管期間）―

技術適用時のランニングコスト

後段プロセスへの引渡し形態（ハンドリング性）（検査・取出し等の後段作業要求を（選択肢に依らず確認すべき事項として抽出し、絞り込み方法の検討対象より除外）
Factor,Control Factor）を適用し、安定性・拡散性に

・選択肢毎の適用条件（適用可能範囲を示す情報）―

・処理後の取扱い形態

SED指標におけるFacility Descriptor
適用する技術に用いる設備コスト

C1.選択肢の適用に掛かるコスト
合理的A6.後段プロセスの作業に則した形態―

処理後に生成される形態に対し、SED指標（FormB

処理後の化学的性状に関する情報―（現状無し）

適用する技術の開発コスト
バウンダリ構成物の性能

２―と共に、リスクの程度により比較評価

List of viewpoints andnarrowing down indicatorsfor narrowing down options
SED指標におけるControl Factor

適用する技術の受入れ条件（適用可能範囲）

後段プロセスへの受渡し形態（バウンダリ構成物）

事象（例：著しい膨潤、発熱等）の有無を確認する

保管或いは埋設施設の受入れ基準（線量率、強度、許容され選択した処理を適用した後の状態での廃棄物の性状処理結果の適切性を判断する定量的な基準（≒安定性／拡散のし難さ） （≒要求事項への適合性判断）
SED指標におけるForm Factor（保管・埋設施設の受入れ基準、輸送基準）
処理後の物理的性状に関する情報
る核種・化学的性質等）（物理的性状）度）

選択肢間の技術成熟度（実績、残る検討課題等）について比較評価――

事前ラボ試験データ等により技術適用時に影響する―（処理対象物の性状変動への対応程

棄物／廃棄体の特性情報と、後段プロセスの受入れ基準の照合により適用可否を評価
・適用実績等を踏まえた、選択肢毎のTRL設定結果
B（後段プロセスが「処分」の場合）（現状無し）

B3.柔軟性

絞り込みの方法案

文参照）
事前ラボ試験データ等により入手した生成される廃A

埋設施設の受入れ基準（定量的な判断基準の設定対象外）（現状無し）

研究開発の進捗

パターン

(凡例は本
・選択肢間での発生物量の相対比較―（後段プロセスが「保管」の場合）（著しい状態変化の有無）

絞り込み・後段プロセス（保管/埋設）の受入れ容量への適合

絞り込み時に必要とする情報 性

保管施設の受入れ基準―

業エリア毎の管理値等）を逸脱するリスクを比較し、汚染拡大リスクが低い選択肢を選定

・処理により発生する物量（評価値）

は、対象の保管施設の受入れ条件・設計条件

B2.明確な判断基準の有無

A5.閉じ込め性を維持する措置

評価指標B
indicator B

・処理後の状態（後段プロセスへ引渡す際の拡散リスクの程判断基準の有無ー（保管・埋設計画がある場合は施設容量）
保管管理計画上で示される廃棄物に関して選択した処理を適用する際の、処理作業中の廃棄物の安定性

確実
A4.閉じ込め性を有する形態処理作業の内容、或いは生じる生成物が基準値（作

度）

ー

判断要素

後段プロセス（保管・埋設等）の受入れ容量への適合性

ばく線量限度を下回ることを原則として、より被ばく影響が低い選択肢を選定・処理作業の内容（作業中における飛散リスクの程度）ー

選択した処理を適用した際の二次廃棄物物量

選択肢の絞り込みの観点

A3.後段プロセスの受入れ判断に資す

B

・作業計画線量

エリア区分毎の表面汚染密度および空気中の汚染拡大防止観点
Viewpoint

処理作業に係る作業員被ばく影響評価の上、年間被・処理対象物の表面線量率データ

適用する技術の処理速度―

作業エリア・敷地の確保―

作業性（環境、アクセス性、操作性）

―

B

B

（開発項目、設計検討項目等）―

B

選択肢の処理速度に関する情報（処理プロセス、廃棄体1体当りの製作期間等）

B

選択肢の適用範囲（廃棄物性状、インベントリ等）の検討結果を基にした比較評価
目的達成可能であることを前提に、候補となる選択
肢の成熟度やリスクを基にした比較評価

選択肢間の開発期間を含めた適用までに掛かる期間の評価値による比較評価
適用する技術の処理速度の想定を用いた比較評価
エリア制約を満足した上で選択肢の適用が可能であ

…

…

３

現場指向

選択肢を適用するエリア（設備導入先など）の制約情報

A、B

ることを前提に、選択肢適用に当たっての課題の程度による比較評価

E1.1F現場環境での成立性
・選択肢を適用する形態（装置・システム等）
・選択肢を適用する際の作業内容

エリア制約を満足した上で選択肢の適用が可能であB

ることを前提に、選択肢を適用する際の形態・作業内容を基に運用性の観点で比較評価

本事業での検討結果
Reference
参照
Gathering of
必要情報の収集
required information

Step 3: Consolidation of
information required forcomparative evaluation

絞り込み
選択肢の絞り込みの観点

絞り込み指標判断要素

判断基準の有無

絞り込み時に必要とする情報

パターン

敷地境界へ与える線量影響評価の上、サイト内で発処理対象物のインベントリデータ（各廃棄物等からの寄与の合計値として）B

Setting information that

収集すべき情報の設定
must be gatheredを原則として、敷地境界に与える影響がより低い選放射性物質濃度の基準値

A3.後段プロセスの受入れ判断に資する特性情報の取得

保管施設の受入れ基準への適合性（例：表面線量率など）

適用実績
B1.選択肢の技術成熟度（技術の適用
埋設施設の受入れ基準（WAC)への適合性性）

以下の2点の評価実施（現状無し）―（著しい状態変化の有無）（現状無し）

B2.明確な判断基準の有無（化学的性状）度）

適用する技術の開発コスト
バウンダリ構成物の性能

技術適用時のランニングコスト

後段プロセスへの引渡し形態（ハンドリング性）（検査・取出し等の後段作業要求を考慮した形態）

Information
選択肢C
on
Option Cに係る情報

供する時間（社会的受容性）
迅速D3.適用から目的達成（処理完了等）
までに要する時間B―（選択肢に依らず確認すべき事項として抽出し、絞り込み方法の検討対象より除外）
Factor,Control Factor）を適用し、安定性・拡散性について選択肢間で比較評価

・選択肢毎の適用条件（適用可能範囲を示す情報）
・選択肢を適用する廃棄物の性状変動リスク

・処理後の取扱い形態――

処理後に生成される形態に対し、SED指標（Form

処理後の化学的性状に関する情報―事象（例：著しい膨潤、発熱等）の有無を確認すると共に、リスクの程度により比較評価

A、B

B

廃棄物の性状変動リスクを踏まえ、選択肢毎の適用条件により、処理対象物の性状への対応可能性の観・後段プロセス（保管/埋設）での受入れ容器が確定点で比較評価している場合は、当該容器を用いた処理作業の成立性により適用性を評価

List of viewpoints andnarrowing down indicatorsfor narrowing down options―（後段プロセスでの取扱い計画／要求）

選択肢の適用に掛かるコスト評価値※Facility Descriptorの判定要素に係る情報―・後段プロセス（保管/埋設）への引渡し形態が複数

選択肢の開発コスト・設備コスト・ランニングコス
存在する場合は、SED指標（Facility
Descriptor）をB（開発コスト／設備コスト／ランニングコスト）
トの評価結果による比較評価
適用し、閉じ込め性について選択肢間で比較評価―

・処理後の取扱い形態

・後段プロセスでの取扱い計画（検査・モニタリング等）―
保管或いは埋設施設へのコスト影響

A、B

・後段プロセス（保管/埋設）での取扱い計画（検後段プロセスへのコスト影響評価結果を用いた比較

査・モニタリング等）への適用性を評価
B評価（後段プロセスでの取扱い計画／要求）―

選択肢の技術的な汎用性に関する情報（適用対象廃棄物）―

・技術の適用実績（成熟度）

・技術適用時のリスク評価結果（特に放射線防護上のリス

安全性（処理・保管・埋設の各段階）に関する説明性―

ク）

技術開発に要する時間―

選択肢の適用までに掛かる工程情報

D2.選択肢の適用に要する時間

選択肢間の技術成熟度（実績、残る検討課題等）について比較評価

A、B

保管或いは埋設施設の受入れ基準（線量率、強度、許容され・バウンダリ構成物の仕様（耐用年数等）※

選択した技術の適用範囲（適用可能な廃棄物）

D1.選択肢の適用に対する理解取得に 技術の適用実績

棄物／廃棄体の特性情報と、後段プロセスの受入れ

事前ラボ試験データ等により技術適用時に影響する

・後段プロセスからのバウンダリへの要求（例：保管期間）―適用した技術を適用した際の後段プロセスへのコスト影響

後段プロセスでの要求（例：検査、モニタリング等）の考慮

C2.選択肢の汎用性

絞り込みの方法案

文参照）―
処理に伴う状態変化リスク（事前ラボ試験データ等）―（現状無し）

SED指標におけるFacility Descriptor
適用する技術に用いる設備コスト

C1.選択肢の適用に掛かるコスト
合理的A6.後段プロセスの作業に則した形態

SED指標におけるControl Factor

適用する技術の受入れ条件（適用可能範囲）

後段プロセスへの受渡し形態（バウンダリ構成物）

パターン

事前ラボ試験データ等により入手した生成される廃A基準の照合により適用可否を評価
・適用実績等を踏まえた、選択肢毎のTRL設定結果
B（後段プロセスが「処分」の場合）―選択した処理を適用した後の状態での廃棄物の性状（処理対象物の性状変動への対応程

絞り込み

・後段プロセス（保管/埋設）の受入れ容量への適合(凡例は本・選択肢間での発生物量の相対比較―（後段プロセスが「保管」の場合）

埋設施設の受入れ基準（定量的な判断基準の設定対象外）

選択した処理を適用した後の状態での廃棄物の性状処理結果の適切性を判断する定量的な基準（≒安定性／拡散のし難さ） （≒要求事項への適合性判断）
SED指標におけるForm Factor（保管・埋設施設の受入れ基準、輸送基準）
処理後の物理的性状に関する情報
る核種・化学的性質等）（物理的性状）

B3.柔軟性

A、B

絞り込み時に必要とする情報 性

は、対象の保管施設の受入れ条件・設計条件―

業エリア毎の管理値等）を逸脱するリスクを比較

処理後の廃棄物のレベル区分別物量の評価を基に、・後段プロセスの受入れ容量（例：埋設施設の容量）

保管施設の受入れ基準

研究開発の進捗A5.閉じ込め性を維持する措置

処理作業の内容、或いは生じる生成物が基準値（作A、Bし、汚染拡大リスクが低い選択肢を選定

・処理により発生する物量（評価値）

判断基準の有無ー（保管・埋設計画がある場合は施設容量）
保管管理計画上で示される廃棄物に関して選択した処理を適用する際の、処理作業中の廃棄物の安定性

確実
A4.閉じ込め性を有する形態

・処理後の状態（後段プロセスへ引渡す際の拡散リスクの程
度）

ー

判断要素

ばく線量限度を下回ることを原則として、より被ばく影響が低い選択肢を選定・処理作業の内容（作業中における飛散リスクの程度）ー

選択した処理を適用した際の二次廃棄物物量

絞り込み指標B

・作業計画線量

エリア区分毎の表面汚染密度および空気中の選択した処理を適用した際の物量（保管・埋設容量）

選択肢の絞り込みの観点処理作業に係る作業員被ばく影響評価の上、年間被・処理対象物の表面線量率データ

年間被ばく線量限度

汚染拡大防止A2.発生量の抑制

安全Information
選択肢A
on
Option Aに係る情報

生する廃棄物からの総影響が基準値未満であること
択肢を選定

処理等の作業員被ばく影響

後段プロセス（保管・埋設等）の受入れ容量への適合性

Information
選択肢B
on
Option Bに係る情報

絞り込みの方法案

(凡例は本
文参照）

1mSv/年未満

敷地境界線量への影響

A1.放射線防護

設備設計・製作を含む現地適用までに要する時間―

適用する技術の処理速度―（開発項目、設計検討項目等）

選択肢の処理速度に関する情報（処理プロセス、廃棄体1体当りの製作期間等）

B

B

B

B

選択肢の適用範囲（廃棄物性状、インベントリ等）の検討結果を基にした比較評価
目的達成可能であることを前提に、候補となる選択
肢の成熟度やリスクを基にした比較評価

選択肢間の開発期間を含めた適用までに掛かる期間の評価値による比較評価

適用する技術の処理速度の想定を用いた比較評価
エリア制約を満足した上で選択肢の適用が可能であ

現場指向

作業エリア・敷地の確保―

作業性（環境、アクセス性、操作性）―

選択肢を適用するエリア（設備導入先など）の制約情報

A、B

ることを前提に、選択肢適用に当たっての課題の程度による比較評価

E1.1F現場環境での成立性
・選択肢を適用する形態（装置・システム等）
・選択肢を適用する際の作業内容

エリア制約を満足した上で選択肢の適用が可能であB

ることを前提に、選択肢を適用する際の形態・作業内容を基に運用性の観点で比較評価(Scope of investigations
conducted in this project)（本事業検討範囲）

Step 4: Selection andnarrowing down by means ofcomparative evaluation

Comparative
比較評価
evaluation

Results of selection
and narrowing down
選定・絞り込み結果

Assumed procedure for selecting and narrowing
down options

Expected to provide information for making
decisions (viewpoints and narrowing down
indicators) for comparative evaluation of options,and to be leveraged for understanding information
required for selection and narrowing down.
©International Research Institute for Nuclear Decommissioning427

(d) Integration of R&D results
Contents of investigations: ② Investigation of information required for the task ofnarrowing down (1/2)[Contents of investigation]➢The source from where information required for narrowing down is acquired is classified into the following 3 categories basedon the results of Contents of investigations ①.
⚫ From the object to be treated (waste): Inventory, quantity, properties⚫ From the treatment technology and method: Treatment speed, TRL, etc.
⚫ From the storage and disposal facilities / plans: Storage capacity, acceptance criteria, etc.
➢Assuming that efforts will be made to further enrich information in accordance with the stage such as research & development
progress, etc., the processes concerning “objects to be treated (waste)”, “treatment technology and method”, and “storage anddisposal facilities/plans” are set.
➢Based on the set processes, the stages at which information required for narrowing down options can be obtained areinvestigated and consolidated.
絞り込み
選択肢の絞り込みの観点

絞り込み指標性）

適用実績―
・適用実績等を踏まえた、選択肢毎のTRL設定結果（≒要求事項への適合性判断）

1mSv/年未満

処理対象物のインベントリデータ―

適用する技術の開発コスト―
・処理対象物の線量データ処理等の作業員被ばく影響

年間被ばく線量限度

・作業計画線量

適用する技術に用いる設備コスト
C1.選択肢の適用に掛かるコスト
汚染拡大防止―技術適用時のランニングコスト エリア区分毎の表面汚染密度および空気中の放射性物質濃度の基準値

適用した技術を適用した際の後段プロセスへのコスト影響
選択した処理を適用した際の物量（保管・埋設容量）C2.選択肢の汎用性

後段プロセス（保管・埋設等）の受入れ容量への適合性

供する時間（社会的受容性）

る特性情報の取得

技術開発に要する時間―

保管或いは埋設施設へのコスト影響

B

以下の2点の評価実施
選択肢の技術的な汎用性に関する情報（適用対象廃棄物）―

保管施設の受入れ基準―
埋設施設の受入れ基準（WAC)への適合性
設備設計・製作を含む現地適用までに要する時間（現状無し）―（現状無し）

埋設施設の受入れ基準

評価

選択肢の適用範囲（廃棄物性状、インベントリ等）

目的達成可能であることを前提に、候補となる選択
・選択肢間での発生物量の相対比較
・技術適用時のリスク評価結果（放射線防護を中心とした安B
肢の成熟度やリスクを基にした比較評価
全の観点でのリスク）(Course followed by waste)*

事前ラボ試験データ等により入手した生成される廃選択肢の適用までに掛かる工程情報A

選択肢間の開発期間を含めた適用までに掛かる期間
棄物／廃棄体の特性情報と、後段プロセスの受入れ（開発項目、設計検討項目等）

基準の照合により適用可否を評価 の評価値による比較評価

選択肢の処理速度に関する情報

事前ラボ試験データ等により技術適用時に影響する―（後段プロセスが「処分」の場合）―

Bの検討結果を基にした比較評価
・後段プロセス（保管/埋設）の受入れ容量への適合ー

A、B

は、対象の保管施設の受入れ条件・設計条件 （後段プロセスが「保管」の場合）―選択した処理を適用する際の、処理作業中の廃棄物の安定性
適用する技術の処理速度までに要する時間（著しい状態変化の有無）

後段プロセスへのコスト影響評価結果を用いた比較
し、汚染拡大リスクが低い選択肢を選定
処理後の廃棄物のレベル区分別物量の評価を基に、―（保管・埋設計画がある場合は施設容量）

トの評価結果による比較評価

業エリア毎の管理値等）を逸脱するリスクを比較・処理技術の技術的成熟度（TRL.適用実績）
・後段プロセスの受入れ容量（例：埋設施設の容量）性―

保管管理計画上で示される廃棄物に関して

D3.適用から目的達成（処理完了等）

く影響が低い選択肢を選定 B

処理作業の内容、或いは生じる生成物が基準値（作A、B・処理により発生する物量（評価値）

安全性（処理・保管・埋設の各段階）に関する説明性

保管施設の受入れ基準への適合性（例：表面線量率など）

・処理後の状態（後段プロセスへ引渡す際の拡散リスクの程度）

ばく線量限度を下回ることを原則として、より被ば
選択肢の開発コスト・設備コスト・ランニングコス（開発コスト／設備コスト／ランニングコスト）―ーD2.選択肢の適用に要する時間

安全

絞り込み方法の検討対象より除外）

The processes for investigating and developing sources of
information (waste / treatment technology / storage anddisposal facilities) are set based on the investigations andconsolidation of the stages at which the required information
can be obtained

処理作業に係る作業員被ばく影響評価の上、年間被選択肢の適用に掛かるコスト評価値B・処理作業の内容（作業中における飛散リスクの程度）

・処理対象物の物量

D1.選択肢の適用に対する理解取得に 技術の適用実績

迅速

文参照）

ー

選択した技術の適用範囲（適用可能な廃棄物）

選択した処理を適用した際の二次廃棄物物量A3.後段プロセスの受入れ判断に資す

ついて比較評価

廃棄物の性状変動リスクを踏まえ、選択肢毎の適用生する廃棄物からの総影響が基準値未満であること
・選択肢毎の適用条件（適用可能範囲を示す情報）
B
B
条件により、処理対象物の性状への対応可能性の観を原則として、敷地境界に与える影響がより低い選・選択肢を適用する廃棄物の性状変動リスク点で比較評価択肢を選定

適用する技術の受入れ条件（適用可能範囲）（各廃棄物等からの寄与の合計値として）度）

A2.発生量の抑制

選択肢間の技術成熟度（実績、残る検討課題等）に絞り込み

敷地境界へ与える線量影響評価の上、サイト内で発敷地境界線量への影響（処理対象物の性状変動への対応程

合理的

絞り込みの方法案

パターン―
絞り込み時に必要とする情報保管或いは埋設施設の受入れ基準（線量率、強度、許容され絞り込みの方法案(凡例は本―（選択肢に依らず確認すべき事項として抽出し、―処理結果の適切性を判断する定量的な基準 （（）内：補足事項）（保管・埋設施設の受入れ基準、輸送基準） る核種・化学的性質等）

B3.柔軟性A1.放射線防護

B―

判断時に考慮必要と考える事項

判断要素

パターン
(凡例は本
文参照）

研究開発の進捗

確実絞り込み指標B2.明確な判断基準の有無

絞り込み時に必要とする情報（（）内：補足事項）

List of viewpoints andnarrowing down indicatorsfor narrowing down options
B1.選択肢の技術成熟度（技術の適用

選択肢の絞り込みの観点

判断時に考慮必要と考える事項

判断要素

B

B

Generation
発生

Characterization
性状把握

適用する技術の処理速度の想定を用いた比較評価

Treatment/waste

処理/廃棄体化
conditioning

Storage/disposal
保管/埋設処理に伴う状態変化リスク（事前ラボ試験データ等）
A、B
事象（例：著しい膨潤、発熱等）の有無を確認する（処理プロセス、廃棄体1体当りの製作期間等）
と共に、リスクの程度により比較評価

エリア制約を満足した上で選択肢の適用が可能であA4.閉じ込め性を有する形態

選択した処理を適用した後の状態での廃棄物の性状作業エリア・敷地の確保（≒安定性／拡散のし難さ）（物理的性状）

現場指向―（定量評価時の活用指標

選択した処理を適用した後の状態での廃棄物の性状（化学的性状）選択肢を適用するエリア（設備設置箇所）の制約情報（定量評価時の活用指標（現状無し）

・選択肢を適用する形態（装置・システム等）
ついて選択肢間で比較評価
・選択肢を適用する際の作業内容―
バウンダリ構成物の性能（定量評価時の活用指標：SED指標におけるFacility Descriptor）

A6.後段プロセスの作業に則した形態 後段プロセスへの引渡し形態（ハンドリング性）（検査・取出し等の後段作業要求を
後段プロセスでの要求（例：検査、モニタリング等）の考慮

エリア制約を満足した上で選択肢の適用が可能であB

ることを前提に、選択肢を適用する際の形態・作業内容を基に運用性の観点で比較評価

・処理後の取扱い形態

・後段プロセス（保管/埋設）での受入れ容器が確定

・後段プロセスからの取扱い形態（バウンダリ構成物）へのしている場合は、当該容器を用いた処理作業の成立要求（例：保管期間）

A5.閉じ込め性を維持する措置

ることを前提に、選択肢適用に当たっての課題の程

Factor,Control Factor）を適用し、安定性・拡散性に―
処理後の形態に対する化学的性状の情報：SED指標におけるControl Factor）―

A、B

処理後に生成される形態に対し、SED指標（Form度による比較評価
B―

作業性（環境、アクセス性、操作性）

後段プロセスへの受渡し形態（バウンダリ構成物）

考慮した形態）―

処理後の形態に対する物理的性状の情報：SED指標におけるForm Factor）

E1.1F現場環境での成立性

・処理後の取扱い形態（バウンダリ構成物）の設計仕様（耐A、B用年数等）※

性により適用性を評価
・後段プロセス（保管/埋設）への引渡し形態が複数(Flow of development and application of treatment technology)*
要素技術開発Development of elemental
technologies

設備仕様検討Investigation on equipment
specifications

存在する場合は、SED指標（Facility Descriptor）を※Facility Descriptorの判定要素に係る情報

Operation

運用計画設定planning

設備設計・製作

Designing/manufacturing ofequipment

Start of運用開始
operation

適用し、閉じ込め性について選択肢間で比較評価―（後段プロセスでの取扱い計画／要求）―・処理後の取扱い形態
・後段プロセスでの取扱い計画（検査・モニタリング等）

A、B

・後段プロセス（保管/埋設）での取扱い計画（検査・モニタリング等）への適用性を評価（後段プロセスでの取扱い計画／要求）(Flow up to designing and operation of storage/disposal facilities)*
Investigation of items to
受入れ対象物検討
be accepted

The sources from where the required information is obtained
are classified into “objects to be treated (waste)”, “treatment
technology and methods”, and “storage and disposal
facilities/plans”

Safety安全評価
evaluation

Setting of design
設計条件設定conditions

Designing/execution
施設設計・施工of the facilities

Starting of

運用開始
operation*In reality, processes are likely to be implemented in parallel or are likely to be iterated, but the
purpose here is to indicate the stage at which information required for narrowing down options
can be acquired. Hence the processes indicated here have been simplified.
©International Research Institute for Nuclear Decommissioning428

(d) Integration of R&D results
Contents of investigations: ② Investigation of information required for the task ofnarrowing down (2/2)[Investigation results: Output]➢The required Information, the source of information, and the stage at which the information can be obtained areinvestigated and consolidated in accordance with the 5 viewpoints and narrowing down indicators consolidated inContents of investigations ①.
No.

絞り込み時に必要とする情報

関連する評価指標

情報の入手元1

処理対象物のインベントリデータ

A1

処理対象物

発生

性状把握

処理/廃棄体化

2

処理対象物の線量データ

A1

処理対象物

発生

性状把握

処理/廃棄体化

保管/埋設

A1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

A1、E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

処理/廃棄体化

保管/埋設

運用計画設定設備設計・製作

The required information is listed basedon the “List of viewpoints and narrowing
down indicators” consolidated inContents of investigations ①.
3

作業計画線量4

処理作業の内容（作業中における飛散リスク）

5

処理後の状態（後段プロセスへ引渡す際の拡散リスク）

A1

6

処理対象物の物量

A2

7

処理により発生する物量（減容率等に基づく評価値）

A2

In association with therelated narrowing down
indicators

8

後段プロセスの受入れ容量（保管施設容量、埋設施設容量）

9

後段プロセスの受入れ基準（保管施設、埋設施設）

10

情報が取得可能となる段階（赤枠内）

Sources of
information were set
処理対象物

処理技術・方法・作業計画

発生

要素技術開発性状把握
設備仕様検討

保管/埋設

運用開始

The stages at which the required information can be obtained areinvestigated and consolidated.
<Example>
Required information: Information from various sources of waste(Inventory, etc.)

A2

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

A3、B2

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

処理に伴う廃棄物の状態変化リスク（ラボ試験データ等）

A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

11

処理後の形態に対する物理的性状の情報

A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

12

処理後の形態に対する化学的性状の情報

A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

13

処理後の取扱い形態（後段プロセスへ引渡す際の形態）

A5、A6

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

14

後段プロセスからの取扱い形態（バウンダリ構成物）への要求事項（保管期間等）

A5

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

15

処理後の取扱い形態（バウンダリ構成物）の設計仕様（耐用年数等）

A5

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

16

後段プロセスでの取扱い計画（検査・モニタリング等の要求）

A6

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

17

処理技術の技術成熟度（TRL、適用実績）

B1、D1

処理技術・方法・作業計画

要素技術開発18

処理技術・方法の適用条件（適用可能範囲）

B3、C2

処理技術・方法・作業計画

要素技術開発19

処理対象物の性状変動リスク

B3

処理対象物

20

処理技術適用に掛かるコスト評価値（開発コスト/設備コスト/ランニングコスト）

C1

処理技術・方法・作業計画

21

保管・埋設施設へのコスト影響

C1

保管・埋設施設・保管/埋設計画

22

処理技術適用時のリスク評価（安全の観点）

D1

処理技術・方法・作業計画

要素技術開発設備仕様検討23

処理技術の適用までに掛かる工程情報（開発期間、設備設計・製作期間）

D2

処理技術・方法・作業計画

要素技術開発設備仕様検討24

適用技術の処理速度に関する情報25

処理技術を適用するエリア（設備設置箇所）の制約情報

26

処理技術を適用する形態（装置・システム）の情報発生
要素技術開発受入れ対象物検討

設備仕様検討設備仕様検討

性状把握
設備仕様検討
安全評価

Generation
発生

運用計画設定運用計画設定性状把握
Characterization

The stages運用開始
at which information can be obtained during the processes
from waste運用開始
generation to storage/disposal are investigated andestablished.
設備設計・製作
運用開始
保管/埋設

運用計画設定設備設計・製作

運用開始

施設設計・施工

運用開始

運用計画設定設備設計・製作

運用開始

運用計画設定設備設計・製作

運用開始

D3

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

List of information required
for narrowing down options

Storage/disposal
保管/埋設

設備設計・製作

処理/廃棄体化

設計条件設定Treatment/waste

処理/廃棄体化
conditioning【凡例】[Legend]【凡例】：処理対象物（廃棄物）より収集が必要な情報：処理技術・方法及び処理作業計画より収集が必要な情報

Information that needs to be collected from objects to be treated (waste)：処理対象物（廃棄物）より収集が必要な情報
Information that needs to be collected form the treatment technology/methods and the treatment plan
：処理技術・方法及び処理作業計画より収集が必要な情報：保管・埋設施設及び関連する作業計画より収集が必要な情報

Information that needs to be collected form the storage/disposal facilities and related work plans：保管・埋設施設及び関連する作業計画より収集が必要な情報©International Research Institute for Nuclear Decommissioning429

(d) Integration of R&D results
Contents of investigations: ③ Narrowing down on a trial basis using therepresentative waste (Case study) (1/3)[Contents of investigations]➢Specific wastes and options (prospective applicable technologies) are narrowed down on a trial basis based on Contents ofinvestigations ① and Contents of investigations ②,
Note: As limited information can be provided at the current point in time, illustrating the task of narrowing down options with
examples and identifying issues in the viewpoints for selecting and narrowing down options are considered as the purpose of thecase study rather than trying to determine the superiority/inferiority of the options.
[Object of the investigations]⚫Target waste: ALPS slurry*(Reason for selection)•Waste that has been selected as a representative waste for investigations conducted until last year as well, and the results ofinvestigations conducted during the subsidized projects can be leveraged.
•Waste for which treatment and disposal work is planned to be implemented on a priority basis in the solid waste storage
management plan[1].
•Waste for which multiple waste stream options are available in the waste conditioning treatment process*For carbonate slurry and iron co-precipitation slurry to which application of low temperature solidification treatment is being
considered.
Generation ofwaste⚫Target process and options: Waste conditioning treatment
process / low temperature solidification treatment (cement
solidification/AAM solidification)(Reason for selection)• Even though options have been narrowed down till
the storage management process based on thedehydration treatment, multiple treatment technologies
remain as options during the waste conditioning process.

Temporarystorage

Preprocessing,processing, reprocessing(Stabilization treatment)

HIC storage

Continuation ofcurrent storage

HIC storage

Prevention ofleakage and spread

Storage with leakage
and spread prevention

Conditioning

Storage

Storage prior
to disposal

Disposal & reuseContamination
category:

Multi-nuclide
HIC(slurry)

Segregation

Storage inseparate
containers

Storage incontainers

Disposal

Storage incontainers

Segregation

Filling of filler
material

Storage incontainers

Cementkneading andsolidification

Reduction ofmoisturecontent

Basic flow

Disposalverification

Filling of filler
material

Dissolution andsolidification

Solidification

Vitrification
HIC surplus
material

Storage

Metal(SUS reinforcement
body, etc.)

Stream S6

Flammablematerials(PE container, etc.)

Stream S8

Decontami
nationFilling sandDismantling andsegregation[Legend]

Existing options/options in the equipment plan of the operator
Options narrowed down based on assumed conditions
Other options[Processes to be investigated]<Status of investigating the ALPS streams>©International Research Institute for Nuclear Decommissioning430

(d) Integration of R&D results
Contents of investigations: ③ Narrowing down on a trial basis using the representative waste (Casestudy) (2/3)[Investigation results]➢Information that can be gathered at the current point in time on representative waste was provided leveraging Contents ofinvestigations ②.
➢Using this information, the approach towards evaluating the narrowing down indicators for which the method for comparativeevaluation of the options was not clear, from among those set in Contents of investigations ①, was illustrated with examples.
No.

絞り込み時に必要とする情報

関連する評価指標

情報の入手元1

処理対象物のインベントリデータ

A1

処理対象物

発生

性状把握

処理/廃棄体化

2

処理対象物の線量データ

A1

処理対象物

発生

性状把握

処理/廃棄体化

保管/埋設

3

作業計画線量A1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

4

処理作業の内容（作業中における飛散リスク）

A1、E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

5

処理後の状態（後段プロセスへ引渡す際の拡散リスク）

A1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

6

処理対象物の物量

A2

処理対象物

処理/廃棄体化

保管/埋設

7

処理により発生する物量（減容率等に基づく評価値）

A2

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

8

後段プロセスの受入れ容量（保管施設容量、埋設施設容量）

9

後段プロセスの受入れ基準（保管施設、埋設施設）

10

情報が取得可能となる段階（赤枠内）

発生

性状把握

保管/埋設

A2

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

A3、B2

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

処理に伴う廃棄物の状態変化リスク（ラボ試験データ等）

A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

11

処理後の形態に対する物理的性状の情報

A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

12

処理後の形態に対する化学的性状の情報

13

処理後の取扱い形態（後段プロセスへ引渡す際の形態）

14

Results of sorting and investigating the information that can begathered for the purpose of evaluating the ALPS slurry low
temperature solidification treatment
A4

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

A5、A6

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

後段プロセスからの取扱い形態（バウンダリ構成物）への要求事項（保管期間等）

A5

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

15

処理後の取扱い形態（バウンダリ構成物）の設計仕様（耐用年数等）

A5

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

16

後段プロセスでの取扱い計画（検査・モニタリング等の要求）

17(Legend)Black background: Information can be gathered up to a certain
extentGray hatching: Items pertaining to which it is difficult to provide
information at this point in time

運用開始

A6

保管・埋設施設・保管/埋設計画

受入れ対象物検討

安全評価

設計条件設定施設設計・施工

運用開始

処理技術の技術成熟度（TRL、適用実績）

B1、D1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

18

処理技術・方法の適用条件（適用可能範囲）

B3、C2

処理技術・方法・作業計画

要素技術開発運用開始

19

処理対象物の性状変動リスク

B3

処理対象物

20

処理技術適用に掛かるコスト評価値（開発コスト/設備コスト/ランニングコスト）

C1

処理技術・方法・作業計画

要素技術開発21

保管・埋設施設へのコスト影響

C1

保管・埋設施設・保管/埋設計画

22

処理技術適用時のリスク評価（安全の観点）

D1

処理技術・方法・作業計画

23

処理技術の適用までに掛かる工程情報（開発期間、設備設計・製作期間）

D2

24

適用技術の処理速度に関する情報

D3

設備仕様検討運用計画設定設備設計・製作

処理/廃棄体化

保管/埋設設備仕様検討運用計画設定設備設計・製作受入れ対象物検討

安全評価

設計条件設定要素技術開発設備仕様検討運用計画設定処理技術・方法・作業計画

要素技術開発設備仕様検討処理技術・方法・作業計画

要素技術開発設備仕様検討発生

性状把握

Based on the information provided, comparative evaluation at the current point
in time was carried out, or indicators that can be considered were set.
Indicator A2: Controlling amount generated
Indicator A4: Form that has good confinement performance (stable/difficult toget diffused)Indicator B4: Flexibility (Scope of application of technology)
Bearing in mind the use of the evaluation indicators that have been applied for 1F, theapproach of quantifying the qualitative indicators was investigated and illustrated with
examplesExample: Quantification of “physical stability”➢Form Factors (table below) were used for ranking the ease of diffusion from theperspective of physical form, for the SED* indicators.
➢The applicable Form Factor was set and compared based on the properties of matter(Example: uniform solidified substances / brittle solidified substances) resulting from theapplication of the selected option (treatment technology)

Form Factor indicator*2, Note
#

Form

FF

1

Gas, liquid, sludge containing lots of moisture* and condensed particles*

運用開始

2

Other sludge

施設設計・施工

運用開始

3

Powder and loose contaminants (surface contamination, etc.)*

1/10=0.1

設備設計・製作

運用開始

4

Adherent* or penetrating contaminants (surface penetration contamination)*

1/100=0.01

運用計画設定設備設計・製作

運用開始

運用計画設定設備設計・製作

運用開始

5

Brittle, easy to break solids (very porous MCCI)*

1/10,000=1E-4

6

Discontinuous solids (size and weight such that manual transportation is possible, suchas pellets, etc.)

1/100,000=1E-5

7Continuous solids

1/1,000,000=1E-6

25

処理技術を適用するエリア（設備設置箇所）の制約情報

E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

26

処理技術を適用する形態（装置・システム）の情報

E1

処理技術・方法・作業計画

要素技術開発設備仕様検討運用計画設定設備設計・製作

運用開始

1
1/10=0.1*Form that was added to the NDA definition in order to increase its relevance to the Fukushima Daiichi Nuclear Power Station

Information related to the subsequent stages (example:
acceptance criteria), and information related to evaluation basedon the equipment designing and operation plan (example:
treatment cost) is difficult to gather.

*1: Safety and Environmental Detriment*2: Cited from the FY2021 Technical Strategic Plan for Decommissioning of the Fukushima Daiichi
Nuclear Power Station of Tokyo Electric Power Company Holdings, Inc. (October 29, 2021, Nuclear
Damage Compensation and Decommissioning Facilitation Corporation)Note) The SED evaluation indicators (Form Factor, etc.) are not necessarily set in stone, but can be
modified so as to make the definition more specific, etc. in accordance with the items to be evaluated,just like the UK NDA definitions were modified to increase their relevance to 1F.
©International Research Institute for Nuclear Decommissioning

431

(d) Integration of R&D results
Contents of investigations: ③ Narrowing down on a trial basis using the representative waste (Case study)[Investigation results](3/3)➢ALPS slurry / Low temperature solidification treatment technology (Cement solidification / AAM solidification) were subject to comparativeevaluation based on the information that can be currently gathered and on the results of investigations on the narrowing down indicators andproposed methods.
Note: Since the results are based on information that can currently be gathered, the following results do not indicate the superiority or inferiorityof the options, but indicate the information that is insufficient for selection and narrowing down.
Viewpoints for narrowing
down options

Narrowing down indicators

Example case study(ALPS slurry)

Determining factor

Cement solidification

AAM solidification

Impact on the site boundary dose
A1. Radiation protection(Same)

Worker exposure impact during treatment, etc.
Preventing spread of contamination
Amount when the selected treatment is used (storage and disposal capacity)

A2: Controlling amount generated

A3. Acquisition of information on
characteristics that contributes todetermining acceptance in subsequent
processes

A4: Form that has good confinement
performance（≒ stable/difficult to get
diffused)

Safe

Amount of secondary waste when the selected treatment is used

(Difficult to evaluate at this point in time)

Applicability to the acceptance capacity in subsequent processes (storage and disposal, etc.)(Difficult to evaluate at this point in time)

Conformance with the acceptance criteria of the storage facility (Example: Surface dose rate, etc.)(Difficult to evaluate at this point in time)

Conformance with the acceptance criteria (WAC) of the disposal facility(Difficult to evaluate at this point in time)

Stability of the waste during treatment when the selected treatment is used (Presence of significant
change in state)
Properties of waste in the state after the selected treatment is used (Physical properties)
Properties of waste in the state after the selected treatment is used (Chemical properties)

A5. Measures for maintaining good
confinement performance

Reliable

Rational(Difficult to evaluate at this point in time)

Form in which waste is transferred to the subsequent processes (Ease of handling)(Difficult to evaluate at this point in time)

Application track record

Quantitative criteria for determining the suitability of treatment results

Conditions for accepting the technology to be used (Scope of applicability)

D3. Time required from application until
goal achievement (treatment completion,
etc.)

Field oriented

E1.1 Feasibility in the site environment(Difficult to evaluate at this point in time)

Narrowing down from the viewpoint of “Safe”◆There is a wide variety of sources of information (from waste /from treatment technology / from storage and disposal facilities).
◆Information needs to be provided based on the following
investigations in particular for enhancing the evaluation.
• Ascertaining the characteristics of treatment based on theconditions set for the actual treatment (container
specifications, mixing or filling & solidification, etc.)• Design conditions of the storage and disposal facilities(acceptance criteria, capacity, etc.)
Narrowing down from the viewpoint of “Reliable”◆Mainly, information related to the performance of treatment
technologies is required.
◆The acceptance criteria for the subsequent processes (storage
and disposal) needs to be set for evaluating the applicability of
the final technological capabilities.

〇

Cost for developing the technology to be applied(Difficult to evaluate at this point in time)

Cost of equipment used for the technology to be applied(Difficult to evaluate at this point in time)

Running costs when the technology is applied(Difficult to evaluate at this point in time)

Cost impact on the subsequent processes when the applicable technology is applied(Difficult to evaluate at this point in time)

Scope of application of the selected technology (Waste to which the technology can be applied)(Difficult to evaluate at this point in time)

Track record of application of the technology(Difficult to evaluate at this point in time)

Time required for developing the technology
D2. Time required for applying thetechnology(Difficult to evaluate at this point in time)-(Same)Progress of research & development

D1. Time used for obtaining understanding
with respect to application of the option
Explainability related to safety (at each stage of treatment, storage and disposal)(Social acceptability)

Quick(CF:100)

Performance of the boundary structure

B3. Flexibility(Extent to which changes in the properties
resulting from target treatment can behandled)C2. Versatility of the option(FF:1E-4)(Difficult to evaluate at this point in time)B2. Presence of clear judgment criteria(≒ Determining suitability to therequirements)

C1. Costs involved in using the option(Same)〇(FF:1E-6)〇(CF:1,000)

Form in which waste is transferred to the subsequent processes (Boundary structure)

A6. Form suitable for the work in
subsequent processes(Considering the requirements of the work
carried out during subsequent processes
Consideration of the requirements of subsequent processes (Example: Inspection, monitoring, etc.)
such as inspection, removal, etc.)

B1. Technology Readiness Level of theoption(Applicability of technology)〇<Consideration from the perspective of providing
information used for comparative evaluation>(Difficult to evaluate at this point in time)

Results of case studies
related to comparativeevaluation of options(Difficult to evaluate at this point in time)

Time required until on-site application including equipment designing and manufacturing(Difficult to evaluate at this point in time)

Treatment speed of the technology to be applied(Difficult to evaluate at this point in time)

Securing of work area and site

(Difficult to evaluate at this point in time)Work efficiency (Environment, accessibility, ease of operation)(Difficult to evaluate at this point in time): Items that were rated to be highly relevant to the indicators during the evaluation using information thatcan be currently gathered.
: Items that were excluded from the evaluation as only limited information can be currently gathered or itis difficult to gather any information.

Narrowing down from the viewpoint of “Rational”, “Quick”,“Field oriented”◆Information from the treatment technology/storage anddisposal facilities is required. And, for acquiring the required
information, actual designing (example: designing the treatment
equipment, designing the storage and disposal facilities) basedon research & development needs to be implemented.

The level of priority of providing information required for thework of narrowing down from the viewpoint of “Safe” and“Reliable” is evaluated to be high considering the conditions
under which the necessary information is acquired (stage ofinvestigation).
©International Research Institute for Nuclear Decommissioning432

(d) Integration of R&D results
Contents of investigations: ④ Investigation of procedures for integration of R&D results (1/3)[Contents of investigations]➢The following points are investigated along with re-consolidating the roles and applications of the main outputs
related to the integration of research results.
(1) Correlation of the outputs during the entire process while integrating the research results(2) The flow of utilization of the outputs while integrating research results
OutputThe sectionin which

investigatio
ns wereconducted
in previous
years

The sectionin which
investigatio
ns will be
conducted
in FY2021

Overview/Applications

Waste management sheet

Provides information on each waste from various sources, status of investigations at various stages from
temporary storage to disposal/reuse, and research results.

Stream (Treatment flow)

Shows the flow from generation to disposal/reuse of each waste including the options available at each stage.
Uses for indicating the direction and goals of research & development

Input consolidation chart

Consolidates input from the previous processes and input (requirements) of the subsequent processes forthe main contents of investigations at each stage (generation to treatment/reuse) of the waste stream.

Temporal axis organization
chart

Makes the progress of research & development related to solid waste visible by consolidating the status ofinvestigations (whether or not investigations need to be conducted and their status) at each stage fromgeneration to treatment/reuse of each waste.

Dashboard

Summarizes information so that the status of progress of research & development related to solid waste
generated at 1F (① Properties of solid waste, ② Direction of management based on the properties (goals),③ Issues in achieving the goals) can be shared with concerned persons.

List of viewpoints andIndicates examples of viewpoints, indicators and narrowing down methods to be considered while selecting
narrowing down indicators forand narrowing down options available at each stage of the waste streams.
narrowing down options
Indicates the information that is believed to be necessary for the task of selecting and narrowing down
options including the source of information and the phase at which the information can be obtained.
*Difference with the input consolidation chart
List of information required forAlong with indicating information focused on while selecting and narrowing down options, the following points
narrowing down options
are added which are not expressed in the input consolidation chart.
・Linking the viewpoints for narrowing down options/narrowing down indicators and the required information
・Reinforcing the information from the temporal axis related to the treatment technology development process
and the process of investigating storage and disposal facilities during which information can be obtained.
©International Research Institute for Nuclear Decommissioning433

(d) Integration of R&D results
Contents of investigations: ④ Investigation of procedures for integration of R&D results (2/3)[Investigation results: (1) Correlation of the outputs during the entire process while integrating the research results]
Flow of integration of research & development results
Output(Specific point in time)

Contents of investigations

Input

Consolidation of flow from generation to disposal/reuse and options
Various sources of solid
waste(Inventory, etc.)

Selection andnarrowing down
Treatment flow of each waste(Reflection of results ofnarrowing down)

Treatment flow of each
waste
(Including multiple options)
Gathering of
required
information

Waste list
TEPCO HD
Storage management
plan, etc.

Insufficient
information andinvestigations

Selection andnarrowing down
Consolidation of the required
information and flow of information

Ascertaining of the source of information required forselection and narrowing down

Sharing of the progress of research & development at a
specific point in time

List of information
required for narrowing
down options

Dashboard

Mid-and-Long-termRoad-map

Input consolidation
chart by waste

Waste
management
sheet
R&D results

Making the progress of research & development visible, and reflecting it
while setting the investigation targets

Temporal axis
organization chart

Setting the viewpoints / indicators for selection and narrowing down andthe comparative evaluation methods

List of viewpoints andindicators for narrowing
down options

Setting the issues to be addressed
and their level of priority based on thestatus at the time of evaluation⇒ Reflection into subsequent
research & development plans

The section in whichinvestigations will be conducted in
FY2021

Setting the entire process involved in integration of R&D results till the stage when it helps in establishing the issues to beaddressed, by indicating the waste streams and research & development status at a specific point in time, andconsolidating the correlation with the phase in which the outputs are utilized©International Research Institute for Nuclear Decommissioning434

(d) Integration of R&D results
Contents of investigations: ④ Investigation of procedures for integration of R&D results (3/3)[Investigation results: (2) The flow of utilization of the outputs while integrating research results]②: Setting the indicators forcomparative evaluation of options①: Setting the target waste/process and verifying the options
Tools used: Streams for each waste (Treatment flow)

Tools used: List of viewpoints and indicators fornarrowing down options

Evaluation indicator
Evaluation indicator A

Option A

Option C

Safe

Option B

Reliable→ Identification of prospective options with respect tospecific wastes/processes

Tools used: List of information required fornarrowing down options
List of information
required for narrowing
down options

List of viewpoints andindicators for narrowing
down options

Viewpoint③: Verifying information required forcomparative evaluation

Evaluation indicator B✓ Verification of information required for selection andnarrowing down (for each viewpoint and indicator)✓ Verification of source of information

Evaluation indicator C→ Setting of comparative evaluation indicators to befocused on (as required)

Updating in accordance with progressin research & development
If the required
information is
insufficient④: Gathering of
required information
Tools used: Waste management sheet*⑤: Consolidation of the status of provision ofrequired information/ issues
Tools used: List of information required for narrowing down options, input
consolidation chart, temporal axis organization chart

Drafting/implementing research& development plan

*All research results should be used as reference for detailed information

List of information
required for narrowing
down options

Waste
management
sheet

Information
on Option A

Information
on Option B

Information
on Option C

Input consolidation
chart for eachwaste

◆ Ascertaining the sources of information (waste/treatment technology/storage anddisposal facilities)◆ Ascertaining the conditions when information can be acquired (Example: Phase oftechnology development)

Temporal axis
organization chart◆ Understanding the information on
progress of research & development✓ Consolidation of issues to be addressed in the subsequent phase→ Consolidation of information required for the evaluation while selecting and narrowing down options

The procedures related to the method of updating associated with the manner in which output will be utilized or
the progress of research & development, during the cycle of investigating ⇒ Identifying issues ⇒ Planning andexecution of research & development, are provided.
©International Research Institute for Nuclear Decommissioning435

(d) Integration of R&D results
Summary

Current results✓ The functions and mechanisms for rationally and efficiently carrying out research & development related to treatment and disposalof solid waste were investigated, and until previous years the waste management sheet was provided from the perspective ofunderstanding the current situation, the treatment flows (streams) were provided from the perspective of sharing the direction of
the research & development, and the input consolidation chart and temporal axis organization chart were provided from theperspective of ascertaining the correlation between the various types of information and of understanding the progress made.
✓ This year, the idea behind the criteria for selecting and narrowing down waste streams from the multiple options that remain andthe indicators for making that decision were investigated.
✓ During the investigations, the viewpoints and indicators that are believed necessary to be considered while selecting andnarrowing down based on the investigations of international standards, etc. were consolidated by classifying the viewpoints into 5
categories, namely, “Safe”, “Reliable”, “Rational”, “Quick” and “Field oriented”. And, along with that, while actually carrying out thetask of selection and narrowing down, the information that needed to be gathered for each indicator was investigated andconsolidated including the source of information and the conditions under which the information was obtained (stage ofinvestigation).
✓ Further, the roles and applications of each year’s output were re-organized while integrating the research results, and thecorrelation with the method of utilizing the outputs in the entire process of integrating the research results was clarified.

Issues, etc.
✓✓✓

It is believed that during research & development related to treatment and disposal of solid waste, it is necessary to repeatedly
implement the cycle of establishing issues to be addressed based on the status of investigations at a specific stage andplanning and implementing further research & development.
While implementing this cycle, during “Integration of research results”, the functions and mechanisms believed to be necessary
at each stage of the cycle were investigated and the entire process of integrating the research results was clarified based on the
results of investigations conducted this year.
Since the output of integration of research results that was investigated will be utilized and the research progress will bereflected accordingly, and since a variety of information will be required for selecting and narrowing down options, it is believedthat the results of clarifying the required information will need to be utilized and inter-disciplinary information will need to beprovided.

©International Research Institute for Nuclear Decommissioning436

3. R & D management*Members of IRID in charge of reporting©International Research Institute for Nuclear Decommissioning437

- Knowledge gathering, decommissioning industrial cluster,
human resource development ◼ Expertise from within Japan as well as overseas is being called upon
 Since FY2014, along with continued cooperation with research institutions within Japan and overseas(CRIEPI, Corporate research institutions, UK’s NNL, etc.), useful information was gathered from experts andexperienced individuals.
◼ Realization of the concept of Fukushima Innovation Coast and forming decommissioning industrial clusters inFukushima In order to facilitate participation of local companies in the decommissioning related industry, collaboration
with local companies such as those in the Fukushima Prefecture Hamadori region, etc. was considered andsome of the trial manufacturing of test equipment related to the development of the sampling system wascarried out in collaboration with the local companies.
◼ Mid-and-long-term human resource development
 The IRID symposium was held and explanatory videos on the results of research & development were playedon the IRID website to promote understanding and awareness by presenting the research results and futureplans.
 Along with research result presentations including 18 items during the “FY2021 Fall Meeting” of the AESJ,
and 2 items (planned) during the “FY2022 Spring Meeting” of the AESJ, reports on treatment technology were
presented during the seminar of the back-end task force to promote understanding and awareness.
 Some of the results of characterization were presented during the intensive courses held at the University ofTsukuba (held in July) and the National Institute of Technology, Fukushima College (held in December) as
educational activities for the students.
 Some of the results of characterization were presented during the decommissioning human resource
development training (held in January 2022) in an effort to promote participation by young researchers.

©International Research Institute for Nuclear Decommissioning438

- Clarification of conditions and specifications, setting of indicators, coordination
with other research & development ◼ Clarification of test conditions or development specifications
 Efforts were made to clarify the test conditions and specifications by bringing about exchange of opinions between
concerned parties on the details of the outputs of the project.
◼ Setting of indicators for the achievements of goals

 In the beginning, the indicators that are supposed to serve as the criteria for determining achievement of project goals
were investigated, and the Technology Readiness Level (TRL) that was considered to be the goal was set.
◼ Coordination with decommissioning work and other research and development
 Information on the status of progress of the projects, the issues, etc. was shared by holding regular meetings between
related projects. Also, the common system for managing information on project plans and results, which has beendeveloped so far, continued to be utilized.
 Efforts were made for opinion exchange and information sharing with the Canister Project and the Fuel Debris Retrieval
Project*, etc. in order to exhaustively promote research & development related to treatment and disposal of solid waste
generated in association with fuel debris retrieval.
 Information was shared with partial proposers (IHI, KURION), and based on that the test conditions were set.
 In connection with the “Project for the Promotion of Nuclear Energy Science & Technology and Human Resource
Development through Concentrating Wisdom”, information was exchanged between Hokkaido University and The
University of Sheffield (Investigation on rational treatment and disposal of contaminated concrete waste consideringchange in properties due to leaching), and the symposium on the same topic was participated in (lectures were given
and young researchers were dispatched).
 Technical information was provided for research at Hokkaido University and University of Fukui (Quantitative evaluationof the changes in long-term state of contaminated reinforced concrete that considers actual equipment environment for
rational disposal).

*Canister Project: “Development of technology for collection, transfer and storage of fuel debris” Project
Fuel Debris Retrieval Project: “Development of technology for further increasing the scale of retrieval of fuel debris andreactor internals” Project©International Research Institute for Nuclear Decommissioning439

- Research management, project reports, information
communication, alternative plans ◼ Research management
 A meeting of all concerned parties of this project was regularly held every month (Waste treatment anddisposal project coordination meeting), to share the schedule of research & development and the statusof progress, and in addition, to determine the policy for resolving issues and implementing measures.
FY2021: April 6, May 7, June 4, July 2, August 6, September 3, October 1, November 5 December 3
FY2022: January 11, February 4, March 4◼ Project reports
 With regards to status of progress, the actual schedule chart was submitted to the MRI Secretariat at theend of every month. Also, the status of progress was reported during the Interim Meeting on October 25- 26.
◼ Enhancing communication of information
 The results of this project were posted on the IRID website and information was communicated such thatgeneral public can easily understand it. Also, information about the results of this project wascommunicated by presenting it at academic conferences, etc. held within Japan and overseas by AESJ
or other foreign institutions.
◼ Advanced preparation of alternative plans Throughout the duration of the project, alternative plans had been considered in the event that the
project does not progress as planned. However, alternative plans did not need to be implemented during
this project.

End©International Research Institute for Nuclear Decommissioning440