【Confidential】

Subsidy Project of Decommissioning and Contaminated Water
Management in the FY2018 Supplementary Budgets

R&D for Treatment and Disposal of Solid
Radioactive Waste
Accomplishment Report for FY2019
December 2020

International Research Institute for Nuclear Decommissioning(IRID)All rights reserved. International Research Institute for Nuclear Decommissioning (IRID) ©International Research Institute for Nuclear Decommissioning【Confidential】

Contents of the Report
1. Approach to Research & Development
2. Project Detailsa. Storage and management(a) Study and evaluation of storage and management methods(b) Development of contamination evaluation technology for solid waste segregationb. Establishment of the concepts of treatment & disposal, and development of safety assessment
techniques(a) Establishment of techniques for selecting the preceding processing method(b) Presentation of disposal methods and development of safety assessment techniques① Study of disposal methods in accordance with classification of solid waste and collection &consolidation of information for establishing safety assessment techniques② Development of techniques for assessing impact of affecting substances, etc. on disposalc. Characterization efficiency improvement(a) Characterization efficiency improvement(b) Development of sampling technology
d. Integration of R&D results
3. R&D Management©International Research Institute for Nuclear Decommissioning1

【Confidential】

1. Approach to Research & Development- Background and Purpose of Research & Development Basic concept of waste management in the Mid-and-Long-Term Road-map◆

The characteristics of solid waste, such as nuclide composition, radioactivity concentration, etc.
need to be understood in order to study solid waste treatment and disposal.

◆

Generated solid waste is stored and managed by means of safe and practical methods basedon the characteristics of the waste.

◆

To ensure safety of solid waste storage and management, techniques for practically selecting
treatment (preceding processing) methods for waste stabilization and solidification are
established, and preceding processing methods are selected based on the establishedtechniques before determining the technical requirements for disposal.

◆

To promote efficient research and development (R&D) on solid waste treatment and disposal, close
cooperation between R&D projects related to the characterization, treatment, and disposal of solid
waste is required. R&D is promoted by sharing research and issues among the R&D teams,
taking an overview of all activities pertaining to solid waste management, and identifying
required R&D tasks.

By around FY2021,
Treatment and disposal measures, and their technical prospects pertaining to safety will be
indicated.

*“Mid-and-Long-Term Road-map towards the Decommissioning of TEPCO HD’s Fukushima Daiichi Nuclear Power Station (September 26, 2017)” 4-5 excerpts, partly reworde©International Research Institute for Nuclear Decommissioning2

【Confidential】

1. Approach to Research & Development
Goals for achieving the mid-and-long-term road-map milestones◼ The following specific goals for achieving milestones set for FY2021 are indicated in 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 goals.

Specific goals related to technical prospects
1.

2.
3.

4.
5.
6.

Safe and practical disposal concepts should be established based on the properties andamount of solid waste generated at the Fukushima Daiichi Nuclear Power Station and thetreatment technologies that can be used for that solid waste, and safety assessment
techniques reflecting the characteristics of the disposal concept should be provided, based on
examples from various foreign countries.
Analysis and evaluation techniques for characterization should be clearly specified.
Treatment technologies for stabilization and solidification considering disposal of various
important waste streams such as secondary waste generated from contaminated water
treatment, which are expected to be introduced in actual conditions, should be clearlyspecified.
Based on the above No. 3., before determining the technical requirements for disposal,
techniques for practically selecting treatment (preceding processing) methods for stabilizationand solidification should be established.
With respect to solid waste for which treatment technology that considers disposal is notclearly specified, establishment of a treatment and disposal plan using the series of techniquesdeveloped until FY2021 should be foreseeable.
The challenges pertaining to storage and management of solid waste until waste conditioning
and corresponding measures should be clearly specified.

# “Technical Strategic Plan 2019 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of “Tokyo Electric Power Company Holdings, Inc.” established by the Nuclear Damage Compensation and Decommissioning
Facilitation Corporation, September 09, 2019, 3.2.2.4.

©International Research Institute for Nuclear Decommissioning3

1. Approach to Research & Development- Goals (1/2) -【Confidential】

Second Term (Period until start of fuel debris retrieval)
Item / Year
2021 onwards
Main events in the current midand-long-term road-map[R&D related to treatment anddisposal of solid waste]

I. Storage and management
1. Study and evaluation of storageand management methods

2. Development of contamination
evaluation technology for solid waste
segregation

II. Establishment of treatment anddisposal concepts and developmentof safety assessment techniques
1.

Establishment of techniques forselecting the preceding
processing method

2. Presentation of disposal methods
and development of safetyassessment techniques△ Summarizing the basic concept related to treatment & disposal
Treatment & disposal plan and technical prospects pertaining to their safety △Evaluation of the type andamount of waste resulting
from debris retrieval
Investigation of cases related tostorage of high-level radioactive
waste within Japan and overseas,
study of applicability to 1F andconsolidation of challenges
Investigation and study of themethods for measurement andevaluation of α contamination,etc.
Technologicalinvestigation, testing,evaluation andpresentation ofcandidatetechnologies

Evaluation of the type and amount of waste accompanying debris
retrieval and revision based on the progress of studying retrieval

Study and presentation of storage and management
methods including hydrogen measures
Systematization for on-site
application

Mock-up test

Acquisition and evaluation of required data on high temperature
treatment technology and low temperature treatment technology,
using engineering scale testing equipment
Identification of technology that can beapplied to actual treatment (of some of
the waste), from the technologicalperspective

Collection and consolidation of items andinformation required for multi-faceted
evaluation, and presentation of challenges
Investigation of disposalplans within Japan andoverseas
Investigation of cases related tomaterials with impact on disposal,
study of analytical techniques

Study of the disposal concepts and safetyassessment techniques

Clear specification of the image of
waste conditioning, consolidation ofdata required for the safetyassessment technique for eachdisposal method

Evaluation study,
study ofmeasures inaccordance withthe on-site status
Technologicaldevelopment
pertaining to themeasurement andevaluation methods

Identification ofpreceding
processing
methodPreliminary safetyassessment

Acquisition of data for impact assessment related to the main materials with
impact, improvement in the impact assessment techniques[Source] Created by extracting R&D items from P.16 of the “FY2019 Decommissioning R&D Plan” of the Team for Countermeasures for Decommissioning and Contaminated
Water Treatment / Secretariat Meeting (63rd session, February 28, 2019)©International Research Institute for Nuclear Decommissioning4

1. Approach to Research & Development- Goals (2/2) -【Confidential】

Second Term (Period until start of fuel debris retrieval)
Item / Year
2021 onwards
Main events in the current midand-long-term road-map
III. Characterization
Characterization efficiency
improvement(1) Establishment of the method ofcharacterization by combining
analysis data and evaluation data
based on the migration models(2) Simplification and speeding upof analysis methods△ Summarizing the basic concept related to treatment & disposal
Treatment & disposal plan and technical prospects pertaining to its safety △Reflection of analysis data
pertaining to secondary waste
generated from contaminatedwater treatment, debris, felledtrees and soil

Development oftechnology forcollecting cesium
adsorbents andsamples from inside
the reactor building,and sampling ofsludge fromdecontamination
systems

Establishment of the method of characterization by combining analysis dataand evaluation data based on the migration models, simplification andspeeding up of analysis methods

Study on theapplicability oftechniques,
reflection into the
Okuma Analysis andResearch Facility
Building 1

Acquisition, evaluation and management, etc. of analysis data, which would
help in simplification and speeding up of analysis methods, study andproposal of techniques for providing manuals, overall needs of waste
management or establishment of consistency

Reflection into the
Okuma Analysis
and Research
Facility Building 1

Element tests and evaluation required for designing mock-up devices

Manufacturing
and sampling test

2. Development of samplingtechnology

IV. Integration of R & D results

Creation of draft proposal,
reflection of the outcome
and revision

Integrated progress in R&D, consistency, challenge evaluation

Evaluation based
on R&D progress

Study of waste streams[Source] P.16 of the “FY2019 Decommissioning R&D Plan” of the Team for Countermeasures for Decommissioning and Contaminated Water Treatment / Secretariat Meeting (63rd session, February 28, 2019)©International Research Institute for Nuclear Decommissioning5

【Confidential】

1. Approach to Research & Development
Reflection of results into the decommissioning of Fukushima
Daiichi Nuclear Power Station
Waste management[Project]

Generation of waste

StorageResearch & development

Storage[Waste]• Contamination
evaluation technology
for solid waste
segregation• Improvement ofcharacterization efficiency• Development of samplingtechnology

2019• Study andevaluation ofstorage andmanagement
methods

Stabilization• Development oftechniques forselecting thepreceding
processing method

Milestones

At present• Integration of
R&D results

2020

2021
Treatment

Disposal• Presentation of
disposal methods, anddevelopment of safetyassessment
techniques

Treatment anddisposal plan, andtechnical prospects
pertaining to its
safety will be
indicated.

©International Research Institute for Nuclear Decommissioning

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【Confidential】

1. Approach to Research & Development
Role-sharing among relevant organizations, and correlation between study items
Tokyo Electric Power Company (TEPCO)
Embodiment of the waste management
plan (from generation to storage)

Nuclear Damage compensation and Decommissioning
Facilitation Corporation (NDF)

Formulation of waste management plan(treatment and disposal)

Formulation of overall strategy

Information flow

IRID
Research and development
pertaining to waste streams

Research and development on
storage and management• Integration of R&D
results• Study and evaluation ofstorage and management
methods• Development of contamination
evaluation technology for solid
waste segregation[Potential risk reduction
measures]Research and development
related to characterization • Characterization
efficiency improvement• Development ofsampling technology

Research and development
related to treatment• Establishment of techniques forselecting the preceding processing
method(Coordination with other projects forhigh temperature treatment
technology)

Research and development
related to disposal•

Presentation of disposal
methods, and developmentof safety assessment
techniques[Comprehensive waste management measures]©International Research Institute for Nuclear Decommissioning7

1. Approach to Research & Development- Correlation with other research projects [Input into this project]
RequiredTime

ID

Implementation Items and Specific Contents(Use of Information)

Required Information

Source ofAcquisition

1

Understanding the storage conditions of highlevel radioactive waste that will be studied
under waste management

Specifications of the consumables in the equipment used
for fuel debris retrieval and of the equipment used in thewater treatment and air conditioning system, as well as
specifications of the retrieval containerWhenrequired2

Understanding the storage conditions of highlevel radioactive waste that will be studied
under waste management

Specifications of equipment used for collection, transfer,and storage of fuel debris canisters

WhenrequiredCanister project3

Understanding the fundamental knowledge
required for studying waste streams and forstudying the approach for evaluating
applicability of treatment technology

Information on high-temperature treatment technologies

WhenrequiredSome proposers

Debris Retrieval
project【Confidential】

Remarks
Information exchange asnecessary
4 times this year
Information exchange asnecessary
3 times this year

Coordinated via MRI[Output from this project]
Usage at Output Destination

RequiredTime

Understanding as fundamental data required for studying
high-temperature treatment technologies

WhenrequiredSome proposers

Sharing the results of storage and management, Confirmation of conformance with the debris retrieval
and regulating the conditions for coordination
processWhenrequired

Debris Retrieval
project

Checking the conditions for evaluating amountof hydrogen generatedWhen
requiredCanister project

ID

Implementation Items and Output Details

1

Research results of subsidized projects so far

2

3Comparison with evaluation of amount of hydrogengenerated by fuel debris

Destination

Remarks

Coordinated via MRI

Note) Debris Retrieval project: Development of Technology for Further Increasing the Scale of Retrieval of Debris and Internal Structures project
Canister project: Development of Technology for Collection, Transfer and Storage of Fuel Debris project©International Research Institute for Nuclear Decommissioning8

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1. Approach to Research & Development- Implementation Schedule (1/4) Implementations details

PlannedActual

FY2019First half

FY2020
Second halfFirst half

Second half

Storage and management(a) Study and evaluation of the storage and management
methods① Investigation and evaluation of waste information, andstudy on the method of storing high-level radioactive waste

Consolidation of the latest
information on high-level
radioactive waste

Study on the high-level
radioactive waste handlingprocess
Study on the measures for long-term
Study on the storage
storage of high-level radioactive waste method in accordance withthe properties of waste

Consolidation of preconditions, investigationof previous research and technologies② Study on the container and storage facility requirements
Study on the concepts
Study on a practical container shape (scenarios) of filter vent andstorage & transfer containers Study on the filter ventrequirement specifications
Study on the drying treatment facility requirements
Consolidation of measurement requirements
Study on measurement items(b) Development of contamination evaluation technology for Prototype improvement
solid waste segregation

Study on the basic functions of
the drying treatment facility
Study on the measurement
methods

Element tests,
systemization preliminary
tests and measurement
Fabrication of thesystem designingmeasurement systemb. Establishment of the concept of treatment and disposal,and development of safety assessment techniques

Study on the storage & transfer
container requirements
Study on the concept of thedrying treatment facility
Study on the measurement
facility

Mock-up tests, performance
evaluation(a) Establishment of techniques for selecting the preceding
processing method① Low-temperature treatment technology
i. Acquisition and evaluation of data on low-temperature
treatment technology, which would help in selecting thetechnology

Study on the techniques for examining the possibility ofsolidification by means of low-temperature treatment(Verification by means of carbonate slurry)
Acquisition of data on the properties of cement and AAM
solidifications in the slurry (carbonate slurry)
Investigation of special cement (bibliographic survey, base metal
performance evaluation)

Study on revision and streamlining of examination techniques(Verification by means of iron co-precipitation slurry)
Acquisition of data on the properties of cement and AAM
solidifications in the slurry (iron co-precipitation slurry)
Investigation of special cement (Identification of the scope of
application using simulated waste)©International Research Institute for Nuclear Decommissioning9

【Confidential】

1. Approach to Research & Development- Implementation Schedule (2/4) Implementations details

PlannedActualii. Investigation related to the degradation of solidifications

FY2019FY2020First half
Second halfFirst half
Second half
Investigation of degradation of solidifications due to heating, etc. (bibliographic survey, base metal performance evaluation
test)
Investigation of degradation of solidifications due to heating, etc. (acquisitionof missing data and verification of performance evaluation)
Evaluation of the relationship between the inventory of Cs, etc. and thetemperature of solidifications (OPC model trial analysis, etc.)
Evaluation of the relationship between the inventory of Cs, etc. and theInvestigation and evaluation of factors influencing long-term degradation(Study, etc. on the mineral phase change of the carbonate slurry)temperature of solidifications (AAM model trial analysis, etc.)
Investigation and evaluation of factors influencing long-term degradation(Study, etc. on the mineral phase change of iron co-precipitation slurry)② Study on the approach for evaluating the applicability oftreatment technologyi.

Investigation and study related to the impact ofwaste composition, etc. on the performance ofsolidificationsii.

Acquisition of data related to facility configuration,etc. for each treatment technologyiii.

Investigation related to Cs volatilization volume
during high-temperature treatment and its control(b) Presentation of disposal methods and development ofsafety evaluation techniques① Study on disposal methods in accordance with theclassification of solid waste and collection andconsolidation of information for establishing safetyassessment techniques② Development of techniques for assessing the impact ofaffecting substances, etc. on disposal

Acquisition of vitrification test data and analysis with the help of acharacteristic model (Evaluation of waste filling density)

Investigation and study related to the configuration, etc. of each treatment
facility

Bibliographic survey related to Cs volatilization volume and consolidation ofengineering test results

Verification of characteristics bymeans of glass melting test

Evaluation of filling density when
multiple types of waste are mixed.

Investigation and study related to the configuration, etc. of each treatment
facility (Supplementing data pertaining to economic efficiency)

Bibliographic survey related to Cs volatilization volume and consolidation ofengineering test results

Consolidation of waste characteristics and study on disposal concepts andsafety evaluation techniques (Waste selected in FY2019)
Study on requirements for the process from disposal to waste conditioning Consolidation of waste characteristics and study on disposal concepts andsafety evaluation techniques (Waste selected in FY2020)
Setting of data acquisition conditions, investigation and data acquisition(focusing on actinide nuclides)

・Investigation of fundamental data

Setting of internal test data acquisition conditions, data acquisition throughtesting (boric acid)・Investigation of techniques for impact assessment

Investigation and establishment of impact
assessment techniques

Setting of parameters
based on data acquired
in FY2019

Study of the image of waste packages

Setting of data acquisition conditions, investigation and data acquisition(focusing on transition metals)
Setting of internal test data acquisition conditions, data acquisition throughtesting (supplementing boric acid)

Investigation and establishment of impact assessment techniques(Supplementary investigation and trial run by means of data acquisition)©International Research Institute for Nuclear Decommissioning

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【Confidential】

1. Approach to Research & Development- Implementation Schedule (3/4) Implementations details

PlannedActual

FY2019FY2020First half

Second halfFirst half

Second halfc. Characterization efficiency improvement(a) Development of a method for estimating the inventory andacquisition of analysis data
Understanding the contamination mechanism (Study of target contaminants① Establishment of the method of characterization by
combining the analysis data and the evaluation data based on
migration models

and case studies)

Understanding the contamination mechanism (Study of applicability to target
waste, etc.)

Evaluation related to representativeness of the analysis data (The properties of waste
related data are consolidated and their correlation with the analysis data is clarified.)Evaluation related to representativeness of the analysis data (The method ofclassifying waste related data is revised and the representativeness isevaluated.)

Evaluation related to representativeness of the analysis data (The properties of waste
related data are consolidated and their correlation with the analysis data is clarified.)
Investigation related to the characteristics of analysis data (Contamination distribution is studied for all types of waste and every storage facility.)

Statistical inventory estimation method (Study on application of statisticaltechniques and typification of waste)

Investigation related to the characteristics of analysis data
(Study of contamination distribution for every storage
container)
Investigation related to the characteristics of analysis
data (Study on creating models showing uneven
distribution of contamination in every storage form)
Statistical inventory estimation method (Study of estimation
method and evaluation of applicability)

Evaluation and management of analysis data (Operation and maintenance of database)Evaluation and management of analysis data (Study on applicability of Bayesian estimation) Evaluation and management of analysis data (Proposal of analysis

planning technique for 1F waste)Collection of analysis data (Sampling, transportation and analysis)② Simplification and speeding up of analysis methodi.

Development of sampling technology

Testing for identifying optimal conditions and designing of mock-up equipment

Fabrication of mock-up testing equipment
Study on collection method

ii.

Streamlining of separation process

Test using simulated samples (Radioactivity measurement method, nonactinide)
Consolidation of separation techniques

Mock-up tests

Test using simulated samples (Radioactivity measurement method,
actinide)

Proposal of efficient analysis techniques
Proposal of calibration methods
Test using simulated samples (ICP-MS method, non-actinide)
Test using simulated samples (ICP-MS method, actinide)©International Research Institute for Nuclear Decommissioning

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【Confidential】

1. Approach to Research & Development- Implementation Schedule (4/4) Implementations details

PlannedActual

FY2019First half

FY2020
Second halfFirst half

Second half

Separation and search test (Non-actinide)
Separation and search test (Actinide)iii. Development of automation technology
iv. Establishment of standard analysis techniques

Feasibility evaluation test (Radioactivity measurement method,
non-actinide)
Feasibility evaluation test (ICP-MS method, non-actinide)

Feasibility evaluation test (ICP-MS method, actinide)(b) Development of sampling technology
d. Integration of R&D results

Feasibility evaluation test (Radioactivity measurement method,
actinide)

Designing of element testing equipment
Fabrication of element testing equipment
Fabrication of simulated adsorption towers, etc.

Study on the element test procedures

Element test
Study on options based on treatment and disposal (ideal storage
conditions, etc.)

Consolidation of the approach towards narrowing down treatment
options (Establishment of waste streams)©International Research Institute for Nuclear Decommissioning

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【Confidential】

1. Approach to Research & Development- Project Organization Chart Tokyo Electric Power Company Holdings,
Inc.
o Coordination from the perspective of
on-site application(Coordinating the reflection of
decommissioning status on site, etc.)
Japan Atomic Energy Agency
oManagement of technological development (a to d)oDevelopment (b, c)oOperation assistance (d, e)
96 persons

Nakayama Co., Ltd. (b.(a)①)
Inspection and calibration of hot disk device
Mitsubishi Materials Corporation (b.(b))
The work of investigating techniques forevaluating the impact of the huge variety ofaffecting substances included in accident waste
on nuclide migration, and nuclide migration data
Nuclear Development Corporation (b.(b))*
Work of acquiring data on sorption of actinide
nuclides into barrier materials under the presenceof substances that co-exist with accident waste

Tokyo Nuclear Services Co., Ltd. (b.(b))*
Work of acquiring fundamental data related tosorption of nuclides under the co-existence ofaffecting substances included in accident waste
Central Research Institute of the Electric Power
Company (b.(a))
Acquisition and consolidation of data used forstudying evaluation approaches (Contract
research)

Taiheiyo Consultant (b.(a)①)
Study related to evaluation of long-term behaviorof cement based materials (Contract research)

International Research Institute for Nuclear Decommissioning
FY2019o Formulation of overall plan and technological management (ato e)o Technological management such as progress of technological
As of March 31, 2020
development, etc. (a to e)*Determined through open bidding, etc.
o Coordination related to technological development (a to e)

Hitachi-GE Nuclear Energy, Ltd.
oDevelopment (a to d)
16 persons

Toshiba Energy Systems &Mitsubishi Heavy
Solutions Corporation
Industries, Ltd.
oDevelopment (a, b, d)oDevelopment (a)8 persons
3 persons

JGC Holdings Corporation (b)*Proposal of disposal methods and study on safety
evaluation techniques

General Engineering Co., Ltd. (d)Assistance in the work of consolidating
information related to the study of waste streams

Tokyo Power Technology Ltd. (c)*Consolidation and investigation of information
related to the waste list

Hamamatsu Photonics K. K. (a)Designing and fabrication of photoelectricdetectors for surface α contamination detectors

Hitachi Healthcare Systems, Inc. (c.(a)①)
Inspection and calibration of radiation
measurement equipment
Tokyo Power Technology Ltd. (c)*Consolidation and investigation of information
related to the waste list
Hitachi Healthcare Systems, Inc. (c.(a)①)
Inspection and calibration of radiation
measurement equipment
Hitachi Transport System, Ltd. (c.(a)①)
Transportation of analysis samples from
Fukushima Daiichi to the analysis facility andreturn transportation of the remaining samples
from the analysis facility
Nuclear Engineering, Ltd. (c(a)①)
Contract work of the operation and maintenanceof facilities related to the back-end technology
development building

Hokkaido University (b.(a)①)
Study related to evaluation of long-term behavior
of alkali activated material (AAM) (Joint research)

Mitsubishi Materials Techno, Co. (c(a)①)
Work of test and maintenance of the
instrumentation facility of the back-end
technology development building

Japan Chemical Analysis Center (c.(a)②)
Research on simple and speedy analysis method
for treatment and disposal of solid waste at the
Fukushima Daiichi Nuclear Power Station(Contract research)Nissin Giken Co., Ltd. (c.(a)①)
Work of measuring the comprehensive
maintenance efficiency of high performance air
filter devices in the back-end technology
development building

Mitsubishi Materials Techno, Co. (c(a)①)
Work of test and maintenance of the powerreceiving facility of the back-end technology
development building
Ascend Co., Ltd. (c(a)①)
Contract for work related to the analysis andtesting related to characteristics investigation of
secondary waste generated from contaminatedwater treatment at Fukushima

Showa Optronics Co., Ltd. (a)Designing and fabrication of optical systems forsurface α contamination detectors
Triterm Co., Ltd. (a)Designing and fabrication of the circuit for surface
α contamination detectors

Atox Co., Ltd.
oDevelopment (b,c)
20 persons

MHI NS Engineering Co., Ltd. (a)Assistance in creating the documents forstudying the requirements of containers andstorage facilities

Orano Atox D&D Solutions (c)Designing and fabrication of element testing
equipment

KK Kouyou (b)Fabrication of simulated adsorption towers

Utoc Corporation (c)Transportation of stored secondary waste
generated from contaminated water treatment(HIC)

Tepco Town Planning Co., Ltd. (U)*IRID symposium

Inspection Development Co., Ltd. (c.(a)①)
Contract for work related to analysis of β nuclides
in the Fukushima contaminated water and thesecondary waste from contaminated water
treatment, and consolidation of results

Operator E1 (c)*Sampling work at the 1F premises

Nuclear Development Corporation (c.(a)①)
Work of analyzing the radiation in accident waste,etc.
National Nuclear Laboratory (UK) (c.(a)①)
Investigation of analysis technologies in the UK(Contract research)
Japan Nuclear Fuel Development Co., Ltd. (c.(a)①)
Analysis and evaluation of highly contaminatedmaterial generated due to the Fukushima Daiichi
Nuclear Power Station accident (Contract research)
Central Research Institute of Electric Power
Industry (c.(a)①)
Study on improving the precision of analyticalevaluation techniques (Contract research)[Legend indicating the work for which the organization is
responsible for]a.Storage and management
b.Establishment of the concepts of disposal and treatment anddevelopment of safety evaluation techniquesc.Characterizationd.Integration of R&D results
e.R&D management©International Research Institute for Nuclear Decommissioning13

【Confidential】

2. Project Detailsa. Storage and Management(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste informationand study on the storage methods of high-level
radioactive waste② Study on container and storage facility
requirements(b) Development of contamination evaluation technology forsolid waste segregation©International Research Institute for Nuclear Decommissioning

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Status of the study on high-level
radioactive waste generated by fueldebris retrieval【Confidential】(a) Study and evaluation of storage and management methods

The FY2017 study wasstarted.
Preconditions for the
study were determined.
・ Waste related information
・ Process requirements

Examples of input required for study
・Information on waste properties → No direct

Waste related information
information ・Range of fuel debris / waste → Under discussion
・Cutting・method, size → multiple options are being
studied ・・・ Requirements for transfer inside the premises → TBD
・ Requirements of waste storage vault for receiving
Process
requirements waste → TBD

The collection, transfer and storage flow wasstudied based on the safety function
requirements and the primary proposal waspresented.
(Refer to the FY2018 Final Report)→The primary proposal on the scenarios is
considered as the starting point of this study.

・ Material accounting and management policy → TBD
・ Sorting → Under examination

・・・Study on last year’s process (example of lid closing and transfer inside thepremises)Work

3. Storage of inner container inside storage
container, and lid closing

6. Transfer inside the premises
Manned truck transfer

The FY2019 Subsidy Project started.

Maintenancecell

The prerequisites were reset based on the latest
information.
・The latest findings were collected from other projects andwaste information was updated.

Workflow

To prevent the contamination of carrying-in / out cell using
the double door system between the maintenance cell andcarrying-in / out cell
. Storage container specifications. Has shielding functionality. Has filter vent functionality

Identified
through studies
up to last year

Specifications. Capable of transferring into the transport cask using
remote operations. Capable of tightening the lid using remote operations

To list down multiple possible scenarios in the handling process ofhigh-level radioactive waste (waste generated by fuel debris
retrieval) based on the latest waste information.

.Transfer inside the premises. Has fall prevention function and structural strengththat prevents the leakage of contents of the waste
container in case it falls. Carrier. Has a function that prevents the storage containerfrom tipping over

. Is corrosion resistant (during storage). The transfer of waste storage container inside thecarrying-in / out cell is done using cart or conveyor belts(details under discussion)

Equipment and facilities
Maintenance cell /Storage
Carrying –in/out
containercell• Maintains theconfinement
Spread of• Confinement of function
contamin Confin fine particles using
ation /
ement anti-dispersing
internal• Controls theagent + filter vent
exposure
exhaust using
HEPA filter• Has a shielding
External Shieldi function using thickradiation ng
shielding

Major events

Safetyfunction

Updating and study on the collection,transfer and storage flow
・ Resetting the prerequisites and reflecting thespecifics on the concepts of equipment andfacilities
・ Verifying consistency throughout the process

Unmanned forklift transfer

Lid
Storagecontainer
Filter vent. Preventing the contamination of carrying-in / out cell

Specifics on the concepts ofequipment and facilities
・Inner container
・ Storage container with filter vent・ Transport cask
・ Drying equipment
・Measuring device

Carrying-in/ out cell

Fire

Hydrogenmeasures• Has a filter ventfunction

Industrial
accidentsOperability• Has operability suchthat it can be handledeasily and safely

Issues•••••Ventilated

Storage of cask using remote operations
Tightening of lid using remote operations
Prevention of contamination in carrying in/out cell bymeans of the double door system
Development of filter vent

Safetyfunction

Equipment and facilities
Transport
Storagecask
container

Structural
strength• Has acontainment

Confin function that
ement prevents thespilling ofcontents

Shielding
HydrogenmeasuresOperability•Transport
vehicle•Has astructural strengththat prevents
transport cask fromgetting damaged incase it falls.

• Has
sealingfunction•
Maintains
shielding function
using the shielding
thickness•
Performs time
management toprevent thehydrogenconcentration fromexceeding the
explosion limit• Has a fall
preventionfunction• Unmannedtransportation
for high-doseitems• Has a fall
preventionfunction

Increase in the throughput of unmanned
operations©International Research Institute for Nuclear Decommissioning

15

【Confidential】(a) Study and evaluation of storage and management methods<Results of studies so far ① An example of setting the prerequisite； waste information>
Example of an existing study
PCV head
RPV heatinsulationmaterial
Steam drying
machineShield plug
RPV head

Shroud headTop guide
Reactorcore
Bottom partof RPVInternal part ofthe pedestal

Waste#5

Classification#1

Shield plug

Solid radioactive wastes

PCV head

Solid radioactive wastes

RPV heat insulation material

Solid radioactive wastes

RPV head

Solid radioactive wastes

Steam drying machine

Solid radioactive wastes

Shroud head

Solid radioactive wastes

Top guide

Fuel debris

ShroudFuel debris

Jet pump

Fuel debris

Fuel debris inside the reactor core

Fuel debris

Fuel debris at the bottom of RPV

Fuel debris

Fuel debris adhering to lower part of
RPV / CRD housing

Fuel debris

Studied by carrying out discussions (conducted
23 times in FY2017-2019) with related projects,
with the existing study as the starting point.
Basic concept of what is treated as waste (provisional)① Removed items
Above the fuel loading position (shroud head - shield plug) andexternal structure of the pedestal② Other waste
Waste that retains its original form and is not likely to havecriticality

Internal structures of pedestal Fuel debris
CRD exchanger

Fuel debris

Fuel debris inside the pedestal Fuel debris
External structures of pedestal Solid radioactive wastes
External partof pedestal

Fuel debris outside the pedestal Fuel debris

Subsidy Project of Decommissioning and Contaminated Water Managementin the FY2014 Supplementary Budgets
Project for Advancement of Retrieval Method and System of Fuel Debris

Waste information is identified from the study
results of related Pjs and consolidated.

Consolidation of waste information (processed based on the FY2018 Final Report)
Estimated generation volume
Waste classification

Generated waste

Estimated dose
(Per reactor)
・Top access debris retrieval: 670 t

Removed items (Core internals, etc.)

Shield plug and PCV head
RPV insulating material and RPV head
Steam dryer and separator
Etc.

Retrieval equipment

Drills, manipulator, camera, etc.

TBD
(Depends on the retrieval method)ーAir-conditioning and water treatment
system waste

HEPA filter and water treatment filter
Waste adsorbent
Etc.

ー

ー

・Side access debris retrieval: 67 t

3.4E+12 -1E+16 Bq/t

Needs to be updated along withthe progress of study of related
projects
In this subsidy project, the waste
information is updated based on
the report on the latest
techniques.

Examples of other prerequisites (temporarily set based on the current waste handling information from 1F site)① Requirements for transfer inside the premises
・ Surface dose of transport cask 30 mSv/h or less・・・ Set with reference to the current waste transfer within 1F # site premises
・Transport cask should be a sealed type cask
・・・ From the viewpoint of preventing the spread of α contamination②Requirements for receiving water at the storage vault
・Surface dose of storage container 10 Sv/h or less・・・ Set with reference to the operation of the solid waste storage building No. 9・ Storage container weight (including shield and contents) 7.5 tons or less・・・ Set with reference to the operation of the solid waste storage building No. 9(# In the following slides, “1F" refers to "Fukushima Daiichi Nuclear Power Station“.)©International Research Institute for Nuclear Decommissioning

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【Confidential】(a) Study and evaluation of storage and management methods<Results of studies so far ② Study on safety functions / hydrogen measures>
Study on safety functions(Excerpts from the FY2018 Final Report)
Assumedsevereaccidents

Safety function
requirements

Spread ofcontamination

Confinement
Corrosion
resistanceInternal
exposure

Structural
strength

External
exposure

Safety functions
are studied basedon theprerequisites.

Investigation on the cases of storing high-level radioactive wastes within Japan and overseas(processed based on the FY2018 Final Report)
Hydrogenmeasures

Storage method

Examples of main
application

Issues while applying

Dry type /Without ventDry + sealed storage
container・ Zion (USA)・Development of drying
technology
・Scale of drying equipment
and drying time

Dry type /With vent#Exhaust throughfilter vent・TMI-2 (USA)・Development of suitablefilter vent

Wet type /Without vent

Full of water + open
storage container

・Shroud replacement
work, etc. (within
Japan)

・Contamination of the pool・Requires securing the pooland a water treatment system

Wet type /With vent

Vent pipe /Compensator・Paks-2(Hungary)

・Development of vent pipe
・Securing the pool

ShieldingFire

HydrogenmeasuresIndustrial
accident

Operability
Main factor
SecondaryfactorFocusing on dry storage that does not require the securing and management of pools,
and based on the results of overseas case studies on filter vents,
the “dry type / without vent” and “dry type / with vent” storage methods were selected as candidates.

Illustration of storage method (left: dry type/ with vent, right: dry type / without
vent)
Filter vent

Identification of the main issues
・Prevention of nuclide dispersion(ensuring sealability)

Hydrogen

Dust (including α)・Release of hydrogen gas(ensuring permeability)

Dust (including α)

Dry waste

Waste
These are conflicting safety function
requirements. Measures to makethese requirements compatible are
required.

Control ofhydrogen

Storage container

Storage container# When transporting inside the premises, it is over-packed with sealedtransport casks and the amount of hydrogen generation is time controlled.

In this subsidy project, the specifics of the concept of filter vent system anddrying methods suitable for fuel debris retrieval waste will be studied.

©International Research Institute for Nuclear Decommissioning17

【Confidential】(a) Study and evaluation of storage and management methods<Results of studies so far ③ Study on collection, transfer and storage flow>
Collection, transfer and storage flow of waste generated by fuel debris retrieval(Example of implementing hydrogen measures using a container with a vent filter in side-access retrieval method PLAN-B)(Studied based on the FY2018 Final Report)
LegendRetrieval - Storage in container - delivery - transfer inside premises

Study on practical shapeand usage of innercontainer

Reactor Building
Lifting
machineLifting
machineExpanded building

Study on the transport cask
and storage container withvent filter

Waste treatment
cell

Fuel debris
retrieval device

Waste container
handling cellSurfacedosemeasurem
ent

Dose levelmeasurement

Inner
container：Highly contaminated area
：Medium contaminated area
：Less contaminated area
：Processes, the necessity ofwhich is under study

Delivery cell
Delivery
confirmation Transport cask(includes
lid closing
surface dose
equipment
measurement)

Storagecontainer lid
closingequipment

Manipulator

Transfer inside thepremises

Fuel debris

Innercontainertransfer
equipment

Draining
equipment

Scope of study for Retrieval projectTransport
cask

Carry-in area

Receiving
confirmation
Transport
cask lid
openingequipment

Transport cask
transfer equipment

Device formeasuring theamount ofhydrogen (gas)generated

Study on drying
treatment

Receiving at the storage vault treatment - storage
Lifting machine

Facility formeasuring
Dryingamount ofradioactivity, equipmentetc.

Samplecontainer(Lifting
machine)Storage containertransfer equipment

Storagecontainer
lid opening
equipment
Manipulator

Sample container

Dryingequipment

Storage container
with filter vent

Storagecontainertransfer
equipment

Study on measurement
items and measurement
methods

Drying area

Storagecontainer
receiving
confirmation

Transfer toanalysis facility

Device for
Storagemeasuring container lid
the amountclosingof hydrogen equipment Gate
(gas)
generated

Transport cask

Transport cask transfer
equipment

Transfer bogie

Transfer vehicle<Illustration of storage container>
Vent

Storage area

Area
monitor・ 4 levels (0, 60, 140, 180 mm)
of shielding are used based on
the dosage of the waste.
・ Maximum weight 7.1 ton
・Estimated number ofcontainers
Approx. 2000 units (for 3
reactors)

Ventilation
systemTransfer toanalysis facility<Concept>●To use a storage container with a vent filter as a hydrogen measure during storage.
●For transportation inside the premises, a sealed transport cask is used to prevent the spread of α-contamination. Hydrogen concentration iscontrolled by means of time management during transfer.
● To prevent the spread of contamination and reduce work exposure, the areas are classified into three or more areas (red, yellow, green)
based on the level of contamination.

In this subsidy project, theconceptual study of major
equipment and facilities will be
carried out, and the process flow will
be updated to reflect the study
results.

©International Research Institute for Nuclear Decommissioning18

【Confidential】(a) Study and evaluation of storage and management methods
Table 1 (a) FY2019 and FY2020 (planned) results on the study and evaluation of storage and management methods
Items

FY2019 Results

FY2020 Results (planned)① Investigation andevaluation of waste
information andstudy on the storage
methods of highlevel radioactive
waste

Investigation andevaluation of waste
information

The latest information such as the typeand quantity of high-level radioactive
waste was consolidated based on thestatus of study of fuel debris retrieval
methods.

(Will be reviewed if necessary)

Study on the scenarios

The possible scenarios (collection,transfer and storage flow) related to thehandling process of high-level radioactive
waste were presented.

The feasibility of multiple scenarios
pertaining to the process of handlinghigh-level radioactive waste will bepresented.

② Study on
container andstorage facility
requirements

Study on the shapeand usage of innercontainer*

Practical inner container dimensions andoperation methods were presented.

(Will be reviewed if necessary)

Study on gas
generation measures

Filter vent application results were
investigated, and the concepts of filter
vent, storage container, transport cask
were presented.

The requirements for filter vent,storage container, transport cask will
be presented.

Study on drying
treatment

The existing drying treatment
technologies were investigated, and theprimary draft on drying treatment
equipment requirements was presented.

Equipment requirements for drying
treatment, etc. will be presented.

Study onmeasurement
methods andmeasurement
equipment

The measurement requirements wereorganized, and the measurement items
were listed based on them.

Measurement method (measurement
items, measurement location,measurement time, etc.) and themeasurement equipment draft will bepresented.

* Inner container: A container for collecting waste from the reactor. To be handled by the retrieval robot and stored in the storage container.
©International Research Institute for Nuclear Decommissioning

19

【Confidential】(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste (Overview)◼

Results up to FY2018



Methods for safely collecting, transferring, and storing high-level
radioactive waste (core internals) generated during fuel debris
retrieval were studied, and the collection, transfer and storage
flow until storage was proposed.



Based on the proposed collection, transfer and storage flow, thefunctions required for the storage container were organized, andstorage container candidates that satisfy the required functions
were listed.

◼

Goals



To list down multiple possible scenarios for the handling processof high-level radioactive waste based on the latest waste
information.

◼ Implementation details
 In FY2019, the latest information such as the type and quantity ofhigh-level radioactive waste was consolidated based on thestatus of study of fuel debris retrieval methods. Also, multiple
possible scenarios for the handling process of representative
high-level radioactive waste were studied.
 In FY2020, the latest information such as the type and quantity ofhigh-level radioactive waste will continue to be consolidated
based on the status of study on fuel debris retrieval methods, andbased on that information, the handling process up to the storageof high-level radioactive waste will be studied. Moreover, theeffects on throughput, storage capacity, etc. with respect to themultiple scenarios studied will be evaluated and their feasibility
will be confirmed.

① Pre-treatment ofwaste and storage incontainers
Surfacecontamination site

Maintenance cell
Anti-dispersing agent coating
Measuring device
device② Storage of the storage
containers in the transport
cask and verification ofdelivery
Carrying-in/out cell
Transport cask
Lid

Conveyor belt
③ Transfer inside thepremises and carrying
into the building
Temporary waste storage
building

Carry-in area

④ Removal of the storage
containers and temporarystorage
Temporary waste storage
building

Temporarystorage area

RadiationmonitorFig. 1 Example of collection, transfer and storage flow for high-level radioactive waste (Overview)Vent

1.00 m◼ Indicators to determine goal achievement
 (FY2019) Latest information such as the type and quantity ofhigh-level radioactive waste should be listed. The multiple
possible scenarios for the handling process of representative
high-level radioactive waste should be listed.
1.42 m (FY2020) The latest information such as the type and quantity of
1.42mhigh-level radioactive waste should be listed, and based on that
Fig. 2 Illustration of the external appearance of waste storage container
information, the handling process up to storage should be(proposed for light weight container#)presented. The feasibility of the multiple scenarios presented
should be listed.
# The dimensions are temporarily set such that the current limit of the 1F waste container
weight is not exceeded when shielding and waste weight are included.
©International Research Institute for Nuclear Decommissioning

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【Confidential】(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Investigation and evaluation of waste information (1/3)< Investigation and evaluation of waste information>● Reports from related Pjs investigated・Subsidy Project of Decommissioning and Contaminated Water
Management in the FY2016 Supplementary Budgets
Project for the Development of Fundamental Technology for
Retrieval of Fuel Debris and Internal Structures
・Subsidy Project of Decommissioning and Contaminated Water
Management in the FY2016 Supplementary Budgets
Project for Advancement of Retrieval Method and System of Fuel
DebrisCollecting information mainly
related to removed items andretrieval equipment
Collecting information mainly
related to liquid and gaseous
secondary waste● Investigation items
・Waste classification (removed items / consumables for retrieval
equipment / liquid / gaseous)・Name of equipment and waste
・Waste properties (material / shape)
・ Amount generated
・ Dose, etc.

● Main items updated based on the investigation
・Organizing the estimated amount of removed items for each method
・Shape of removed item after cutting・Specification of waste names for retrieval equipment (identification of equipment that will become
waste)
・Specification of waste names for liquid-/ gaseous secondary waste
・Amount of liquid secondary waste (partial) generated● Evaluation items
・ Rate of generation of hydrogen from the removed items©International Research Institute for Nuclear Decommissioning

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(a) Study and evaluation of storage and management methods【Confidential】① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Investigation and evaluation of waste information (2/3)
Organizing the information on the waste generated during debris retrieval

Common preconditions for the studies mentioned from the next page onwards
Table 1 Overview of results of organizing information on waste generated during fuel debris retrieval#1
Important updates are shown in red.
Classification ofgenerated waste
Items
removedfrom floor

Removed
items

Items
removedfromcontainment vessel /pressure
vesselMain waste generated

Estimated dose

Method ofgeneration

Estimated generation
amount#3(Per reactor)

Shape of the waste#2

Items removed from 1F LーSide-accessretrieval□260 mm to □500 mm / plate

Shield plug
DS slot plug

4 Sv/hTop-accessretrieval□1800 mm × thickness 306 mm /plate

1 ton (PLAN - C)
24 ton (PLAN - A, B)

91 ton (Route A)
792 ton (Route B)

PCV head / RPV heatinsulating material
RPV head / steam dryer
Separator

4 to 400 Sv/ｈ

Top-accessretrieval・□108 mm to □500 mm / plate・Φ150 mm / cylinder

Pedestal externals

ーSide-accessretrieval・□108 mm to □500 mm / plate・Φ150 mm / cylinder

38 ton (PLAN - A)
157 ton (PLAN - B, C)

ー

ー

ー

ー

ー

ー

Retrieval equipment

Pit / cutter
Wire / grip claw
Cable / sealing materialー

Liquid secondary waste

Filter material of the filter /packingーGaseous secondary
waste

Filter material of the filter /packingーTop-accessretrieval,Side-accessretrievalTop-accessretrieval,Side-accessretrievalTop-accessretrieval,Side-accessretrieval#1 Corresponds to “waste generated during fuel debris retrieval” in the waste stream classification.
#2 “□260mm to □500mm” refers to squares with lengths/widths from 260 × 260 mm to 500 ×500 mm. “Φ150 mm” refers to “a circle with a diameter of 150 mm”.
#3 Route A, B method is debris top-access retrieval method, PLAN-A, B, C method is debris side-access retrieval method (Refer to the FY2018 Final report of the
Project on Upgrading of Fundamental Technology for Retrieval of Fuel Debris and Internal Structures)©International Research Institute for Nuclear Decommissioning

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(a) Study and evaluation of storage and management methods【Confidential】① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Investigation and evaluation of waste information (3/3)

Evaluating hydrogen generation rate for representative waste(a) ② Used as input conditions for gas generation measures## In FY2019, "(a) ② Gas generation measures" were studied based on the results of this evaluation.
Based on the study results, the evaluation model will be reviewed in FY2020 as necessary.

Table 1 Estimated hydrogen generation rate (Shows the maximum value of metal / concrete waste from
top-access retrieval / side-access retrieval method)Top-access retrieval
metal waste

Top-access retrieval
concrete waste

Side-access retrieval
metal waste

Side-access retrieval
concrete waste

1.90E+15

3.40E+12

3.40E+12

3.40E+12

0.6

2.08

1.3

2.08

1.1E+15

7.1E+124.4E+127.1E+12

Energy emitted(MeV/s)1.3E+15

7.9E+12

4.9E+127.9E+12

Hydrogen generation rate (m3/d)

1.8E-02

1.1E-04

7.1E-05

1.1E-04<Reference>
Anticipated waste

Dryer

ShieldPlugRemoved items fromoutside the pedestal

Removed items fromoutside the pedestal

Items
Dose (Bq/t)

Cs-137

Storage capacity per storage
container (t)Source (Bq)

Cs-137<Reference> Evaluation conditions (set with reference to the results of hydrogen generation measures in FY2017 FY2018)
Calculation
parameters
Set value
Evaluation formulaHydrogen generation rate = ∑ i, j (Decay heat of nuclidei x G value of ray typej x absorption efficiency of ray typej)
Decay heat of nuclidei = ∑ j (Inventory of nuclidei × emission rate of ray typej × emission energy of ray typej)

G value (β, γ rays)

0.45 (Set with reference to the results of FY2018 hydrogen generation measures)

Absorption efficiency

1.0 (Set to the most conservative value)

Energy released per disintegration of Cs-137

1.11 MeV(= 0.662 × 0.851 + 0.514 × 0.944 + 1.176 × 0.056)[Probability of occurrence per disintegration of γ and β rays × energy.
Energy of β rays calculated at the maximum value (conservative) ]©International Research Institute for Nuclear Decommissioning23

【Confidential】(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Study on the scenarios (1/3)▸ Based on the study results, the collection, transfer and storage flow was updated.
▸ It was confirmed that significant changes were not required in the updating of the collection, transfer andstorage flow.
Work
作業

1. Filling the waste inside the inner container andtransferring it within the cell
１．廃棄物を内容器に詰め込み・セル内移送
High surface dose
/ Contaminatedarea

Workflow

作業フロー

3. Storing the inner container inside the storage
container and closing its lid
３．内容器を保管容器へ収納・蓋締め

2. Various measurements
２．各種計測High surface
dose /
ContaminatedareaHigh surface dose
/ Contaminatedarea

Retrieval cell
Waste retrieval
device Inner container(wastecollection
container)

Measuring
deviceRetrieval cell
(Samplingdevice)Medium surface
dose / Contaminatedarea

Storagecontainer
handling cell(maintenancecell)
Measuring

Retrieval cell

•Inner container (means of transfer within the cell)•Prevention of contamination in storage container handling cells
・保管容器取り扱いセルの汚染防止
・各種計測（詳細検討中）
・内容器（セル内移送手段）•No shielding (The inner container is usually handled•Various measurements (detailed study in progress)•The double door system or other equivalent substitute between
・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
・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）
through remote operations)•Measured while stored in inner container
the retrieval cell and the storage container handling cell controls
品により、保管容器取り扱いセルの汚染を抑制
・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• Study on a practical shape for the inner container, taking ・測定の候補•Candidates for measurement
the contamination in the storage container handling cell.
・重量 •Weight
への収納高率等を踏まえて合理的な形状を検討
into consideration its contact with the retrieval equipment,
・保管容器の仕様of storage container
・表面線量率・概算必要個数（上取り出しルートＢ工法1号機あたり）
storage efficiency in storage containers with multiple•Surface dose rate
•Specification
・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, 180
・推定の候補shielding thicknesses,
etc.
•Candidates for estimation•Has shielding function (multiple shielding thicknesses①内容器□500
mm×300 mmの場合→10016個mm）depending on the contents: 0, 60, 140, 180)
•Estimatednumberof containers(For 1 unit by means of the•Amount of radioactivity (γ)
・放射線量（γ）②内容器□335
mm×272mmの場合→27216個
top-accessretrieval
Routemmの場合→14612個
B method)•Amount of radioactivity (α and β)•Has filter vent function
・フィルタベント機能を有する・放射能量（α・β）③内容器□450mm×272① In the caseof a 口mmの場合→10180個
inner container with dimensions 500•Amount of nuclear material•Can be transferred to the transport cask through remote
・遠隔操作により移送容器へ収納可能
・核物質量④内容器□535mm×272operations
mm × 300 mm -> 10016 containers
・遠隔操作により蓋締め可能
・発熱量 •Amount of heat generated② In the case of a 口 inner container with dimensions 335•Hydrogen generation rate
•Lid can be closed through remote operations
・保管時の腐食を考慮する
・水素発生速度・ベルトコンベア（セル内移送手段の例）
mm × 272 mm -> 27216 containers•Considers corrosion during storage
・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・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•Implementation of sampling (detailed study in progress)•Considering transportation by an unmanned forklift in the③ In the case of a 口 inner container with dimensions 450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・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
mm × 272 mm -> 14612 containers•The waste is sampled and transferred to the analyticalstorage vault, the dimensions
arem×1.0
set som）
that the weight of the
寸法を設定する。（暫定外径：1.4m×1.4・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④ In the case of a 口 inner container with dimensions 535
facility, as needed.
container after storage does not exceed the current unmanned
・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
・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。
forklift weight limit. (Provisional outer diameter: 1.4 m x 1.4 m x
mm × 272 mm -> 10180 containers
・サブプロセス
する可能性がある。
•Sub-process
1.0 m)
例）測定値が基準値を上回った場合
・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・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、•Conveyor belt (example of means of transfer within the cell)
Example) If the measurement values exceed the standard •Somelarge removed items may be cut to the size of the storage
される可能性がある。
またはベルトコンベアを用いる•To improve throughput and simplify remote operations, a→2.a 保管容器あたりの内容器収納数変更
valuescontainerand stored directly in the storage container.
・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・必要に応じて保管容器の各種測定を行う
conveyor belt is used for transfer inside the retrieval cell. →2.b 撤去物の詰め直し-> 2.a Change the number of inner containers stored•The
transfer of waste container inside the storage container
収納される可能性がある。
例）表面線量測定、表面汚染
inside each storage container
handling
cell is done using cranes, bogie or conveyor belt.
・サブプロセス•The dimensions, appearance, surface condition, and materials-> 2.b Repack removed items•Variousmeasurements of the storage container are taken if例）蓋締め不具合が生じた場合→（3.a蓋締め直し）
are checked using a camera during retrieval.
necessary.
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）•Some large sized removed items may not be stored in an inner
Example)Surface dose measurement, surface contamination
container and may be delivered from inside the PCV.
•Sub-process•Some large sized removed items may be cut to match the size
Example) When a problem with lid closing arises -> (3.a Close the lidof the inner container and stored in the inner container.
once again)
Example) When surface contamination
exceeds the receivable dose in
装置・設備
装置・設備
装置・設備
the
storage vault (3.b Decontaminate)
安全機能
安全機能
安全機能
取出しセル / 保守セル

閉じ込め

Major events

Spread ofcontamination / Confinement
重大 Internal
外部被ばく
遮へい
事象 exposure
External
火災
exposure

Fire

作業事故

Industrial
accident
課題

Issues-Hydrogen• The top portion is
・容易に、かつ、安全に取り
measuresalways open.
扱える操作性を有する

Safety functions
閉じ込め• Maintains theconfinement
function
Confinement• Manages the・遮へい厚により遮へい機能
exhausts using HEPA遮へいを有する
filter

水素対策
Shielding ・上部は常に解放-

操作性▸ It was confirmed that there were no issues
in the flow of the updated waste information.

Inner container▸ Updated to store multiple inner containers
inside the storage container▸ The number of storage containers required
was organized based on the shape of theinner container

Study on measurement
methods▸ A primary draft on the confirmation items
and measurement items was added to eachprocess.

• Ventilates--

Hydrogen・容易に、かつ、安全に取り
measures

操作性

Equipment and facilities

Safety functions
・閉じ込め機能を確保する
・閉じ込め機能を確保する
・フィルタベントによる微粒子
Maintenance cell /・HEPAフィルタ等で排気を閉じ込め
・HEPAフィルタ等で排気をInner containerの閉じ込め
Maintenancecell
Storage container
管理
管理
Carrying-in / out cell

• Has a shielding
・換気する
function using thick水素対策
Shielding ・上部は常に解放 shielding• Has operability toenable easy and safe
Operability
・合理的な内容器形状の検討
handling• Study on the practical shape of inner container

保守セル / 搬出入セル

保管容器

Equipment and facilities

・閉じ込め機能を確保する。
Handling cell /Innercontainer ・HEPAフィルタ等で排気を管理
Maintenance
cell

-

保守セル

内容器

Equipment and facilities
Safetyfunctions

Waste information

Conveyor belt

内容器

汚染拡大
/内部被ばくUpdate details

Lid
Filter vent

Conveyor belt

仕様

Main update items

deviceStoragecontainer
Bogie

Specifications

Table 1 Main locations of update and update details

扱える操作性を有する• Maintains the• Maintains theconfinement
confinement• Confinement of finefunctionfunction
Confinement
particles using filter• Manages the• Manages the・遮へい厚により遮へい機能
・内容物に応じて複数の
exhaustsvent
遮へい- exhausts using
を有する
遮蔽厚
using HEPAHEPA filter
filter• Has a• Multiple thicknesses of・フィルタベント機能を
shielding
・換気する
水素対策
shielding depending on・換気する
Shielding
有するfunction using
the contents
thick shielding• Ventilates-

Hydrogen• Has the filter vent・容易に、かつ、安全に取り
measuresfunction

操作性

扱える操作性を有する

For the filter vent, storage container, transport cask, and drying
equipment, the study results are planned to be reflected in FY2020.

• Ventilates-• Has operability to• Has operability toenable easy and safe
Operability enable easy and safe
Operability
・各種測定内容の検討
・遠隔操作による容器収納・サンプリング方法の検討 handling
・遠隔操作による蓋締め handling• Study on the various types of measurements of the inner ・ダブルドアシステムによる保管容器取り扱いセルの汚染防止•Storageincontainers
using
remoteoperations
・フィルタベントの開発
container•Closing of lid using remote operations
・各種測定の検討• Study on the sampling methods•Prevention of spread of contamination in storage container handlingcell using double door system•Development of filter vent•Study on various measurements

Fig. 1 Updated collection, transfer and storage flow (partial excerpt)©International Research Institute for Nuclear Decommissioning24

【Confidential】(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Study on the scenarios (2/3)

Updated collection, transfer and storage flow (1/2)
1. Filling the waste inside the inner container and transferring it within the cellWork

2. Various measurements

High surface dose /Contaminated site

Waste retrieval
deviceWorkflow

3. Storing the inner container inside the storage container and closing its lid

High surface dose /Contaminated site

Retrieval cell

Retrieval cell

Retrieval cell

4. Storing the storage container inside the transport cask and closing its lid

Medium surface dose
/ Contaminated site
Storagecontainer
handling cell(maintenancecell)

Sampling device

Measuring device

Inner container(waste collection
container)

Medium surface dose
/ Contaminated site

High surface dose /Contaminated site

Low surface dose /Contaminated site

Conveyor belt

Equipment and facilities•Prevention of contamination in storage container handling cells•The double door system or other equivalent substitute between the
retrieval cell and the storage container handling cell controls thecontamination in the storage container handling cell.

Lid
Waste
transport cask
Bogie

Equipment and facilities

Maintenance cell

Storage container

Maintenance cell / Carrying-in /out cellStorage container

Major events-• Maintains the confinement
function• Manages the exhausts using
HEPA filter

Confinement-• Maintains the confinement
function• Manages the exhausts using
HEPA filter

Confinement• Confinement offine particles using
filter vent• Maintains the confinement
function• Manages the exhausts using
HEPA filter

Confinement• Confinement offine particles
using filter ventExternal exposure

Shielding-• Has a shielding function using
thick shielding

Shielding-• Has a shielding function using
thick shielding

Shielding• Multiple
thicknesses ofshieldingdepending on thecontents-

Shielding• Multiple
thicknesses ofshieldingdepending on
the contents

Fire

Hydrogen measures• The top portion isalways open.

Hydrogen measures• Has the filter ventfunction

Hydrogenmeasures• Has the filter
vent function

Industrial
accident

Operability• Has operability toenable easy andsafe handling

Operability• Has operability toenable easy andsafe handling

Hydrogen measures-

Operability• Has operability toenable easy and safe
handling••• Ventilates-

Study on the various types of measurements of the inner container
Study on the sampling methods

Equipment and facilities

Safetyfunctions

Confinement• Ventilates

Delivery confirmation• Delivery confirmation (surface contamination confirmation, surface
dose rate confirmation, seal check, external appearance
confirmation (including ID confirmation)) of the storage container
is required before transportation.
Surface dose measuring device (detailed study under progress)• Measuring the surface dose and determining the method fortransfer within the premises• Transferring of waste within the 1F premises (at present)• Surface dose rate > 30 mSv/h: Remote (unmanned) transfer• Surface dose rate < 30 mSv/h: Manned truck transfer

Equipment and facilities

Safetyfunctions

Safety functions
Inner container

Handling cell / Maintenance cell

•Prevention of spread of contamination in storage container handling cell•
• A double door system or equivalent substitute between the retrieval
cell and the storage container handling cell controls the contaminationin the storage container handling cell.

Spread ofcontamination /Internal exposure

IssuesPre-delivery test
device•Specification of storage container•Has shielding function (multiple shielding thicknesses depending on the•Specification of transport cask•
contents: 0, 60, 140, 180)
• Has shielding function•Has filter vent function•Can be transferred to the transport cask through remote operations• Has confinement function•Lid can be closed through remote operations• The storage container can be stored through remote operations.
•Considers corrosion during storage• The lid can be closed through remote operations.
•Considering transportation by an unmanned forklift in the storage vault,
the dimensions are set so that the weight of the container after storage• Has a structural strength that prevents the contents from spilling out
does not exceed the current unmanned forklift weight limit. (Provisional
even if the container tumbles during transfer
outer diameter: 1.4 m x 1.4 m x 1.0 m)• Has a spatial volume that maintains the hydrogen concentration below • Sub-process•Some large removed items may be cut to the size of the storage container andthe lower explosive limit
Example) When there is a problem with closing the lid -> (5.a Close the lid
stored directly in the storage container.
•The transfer of waste container inside the storage container handling cell is
once again)
done using cranes, bogie or conveyor belt.
• Make it a returnable container method where it can be used repeatedly. Example) When surface contamination exceeds the standard transfer dose
•Various measurements of the storage container are taken if necessary.
• The waste containers are transported within the carrying-in/out cell byrate (5.b Decontaminate)
Example) Surface dose measurement, surface contamination•Sub-processmeans of cranes, bogie or conveyor belt.
Example) When a problem with lid closing arises -> (3.a Close the lid once again)
Example) When surface contamination exceeds the receivable dose in the
•
Some
of
the
large
removed
items
may
be
storeddirectlyin
specialstorage vault (3.b Decontaminate)containers.

Equipment and facilities
Safety functions

Study on the practical shape of inner container

Carrying in/ out cell

Conveyor belt

Safety functions•

Carrying
in/out cell

Measuring device
Lid•Inner container (means of transfer within the cell)•No shielding (The inner container is usually handled through remote•Various measurements (detailed study in progress)
operations)•Measured while stored in inner container• Study on a practical shape for the inner container, taking into consideration•Candidates for measurement
its contact with the retrieval equipment, storage efficiency in storage•Weight
containers with multiple shielding thicknesses, etc.
•Surface dose rate
•Estimated number of containers (For 1 unit by means of the top-access retrieval•Candidates for estimation
Route B method)•Amount of radioactivity (γ)① In the case of a 口 inner container with dimensions 500 mm × 300 mm ->•Amount of radioactivity (α and β)
10016 containers•Amount of nuclear material•Amount of heat generated② In the case of a 口 inner container with dimensions 335 mm × 272 mm ->27216 containers•Hydrogen generation rate
③ In the case of a 口 inner container with dimensions 450 mm × 272 mm ->14612 containers•Implementation of sampling (detailed study in progress)④ In the case of a 口 inner container with dimensions 535 mm × 272 mm ->•The waste is sampled and transferred to the analytical facility, as needed.
10180 containers•Sub-process•Conveyor belt (example of means of transfer within the cell)
Example) If the measurement values exceed the standard values•To improve throughput and simplify remote operations, a conveyor belt is-> 2.a Change the number of inner containers stored inside each storage
used for transfer inside the retrieval cell.
container-> 2.b Repack removed items•The dimensions, appearance, surface condition, and materials are checked usinga camera during retrieval.
•Some large sized removed items may not be stored in an inner container and may
be delivered from inside the PCV.
•Some large sized removed items may be cut to match the size of the innercontainer and stored in the inner container.

Inner container

Building

Storagecontainer
handling cell(maintenancecell)Filter vent
Storagecontainer
Bogie

Specifications

5. Delivery confirmation and loading of transport cask

Low surface dose /Contaminated site

• Ventilates-

Operability• Has operability
to enable easyand safe
handling

Carrying-in/out cell

Transport cask

Storage container• Maintains theconfinement• Confinement of• Has sealing
functionfine particles
Confinement• Manages the function
using filter ventexhausts using
HEPA filter• Multiple• Maintains shielding
thicknesses ofShielding shielding depending
function using thickshieldingon the contents• Ventilates-• Performs time
management toprevent the hydrogen
Hydrogenmeasuresconcentration fromexceeding the explosion
limit• Has operability toenable easy and safe
handling•Storage in containers using remote operations• Storage of containers using remote operations••Closing of lid using remote operations• Closing of lid using remote operations••Prevention of spread of contamination in storage container handling cell using• Prevention of spread of contamination in storage container handling cell by •double door system•Development of filter ventmeans of double door system•Study on various measurements

Operability• Has the filter
vent function• Has operability
to enable easyand safe
handling

Carrying-in/out cell

Transport cask

• Has sealing
function--• Ventilates-• Maintains
shielding function
using thick shielding• Performs time
management toprevent thehydrogenconcentration fromexceeding the
explosion limit• Has operability
to enable easy andsafe handling

Study on the delivery test items
Reduction of exposure while loading in transfer truck
Confirmation of sealing performance©International Research Institute for Nuclear Decommissioning25

【Confidential】(a) Study and evaluation of storage and management methods① Investigation and evaluation of waste information and study on the storage methods of high-level radioactive waste
Study on the scenarios (3/3)

Updated collection, transfer and storage flow (2/2)
7. Acceptance test of transport cask and carrying-in inside thestorage building

6. Transfer within the premises

Work
Workflow

Low surface dose / Contaminated site

Low surface dose /Contaminated site

Mannedtransfer truck

Temporary waste storage building

8. Removing the lid of transport cask and carrying-in ofstorage container

Medium surface dose /Low surface dose /Contaminated site
Contaminated site
Temporary waste storage building

Carryingin area

Carrying-in area•

Transfer within the premises•
The transport cask has a fall prevention function
and structural strength that prevents the contentsof the waste container from being spilled if it falls.
Transfer vehicle•
The storage container has a fall prevention•function.
•
It has a shielding function.
••

Specifications

Spread ofcontamination /Internal
exposure

Structural
strength

Confinement

Equipment and facilities
Storagecontainer-

Transport cask
• Has a structural
strength that prevents
transport containerfrom getting damaged
if it falls.

Major events• Confines
Has sealing
fine particles •function
using filter ventExternal
exposure

Shielding• Multiple
thicknesses ofshieldingdepending on
the contents

Fire

Hydrogenmeasures• Performs time
management to• Has a filter prevent the hydrogenvent function concentration fromexceeding the
explosion limit

Industrial
accident-

Acceptance test•
Carrying out measurements to check whether or not
the transport cask meets the acceptance criteria of
the storage building•
It is assumed that the acceptable dose limit of thestorage vault is tentatively equivalent to that inbasement 2nd floor of the solid waste storage building
No.9.
•
Acceptable dose limit of solid waste storage
building No.9••
Surface dose < 30m Sv / h: basement 1st floor•
Surface dose < 10 mSv / h: basement 2nd floor-•

Transport cask

Transfervehicle-• Has astructural strengththat prevents
transport containerfrom getting
damaged if it falls.

-

Confines fineHas sealing
particles using •function
filter vent-Hydrogenmeasures• Has a filter
vent function

Operability• Has operability • Has operability • Has a fallthat enables easy that enables easy prevention
and safe handling and safe handling function.

Increase in the throughput of unmanned operations ••• Maintains• Has ashieldingshielding function
performance using (unmanned
the shielding
transportation for
thicknesshigh-dose items)• Performs time
management toprevent thehydrogenconcentration fromexceeding the
explosion limit

Study on acceptance test items
Methods for measurement management••

Operability

Confinement••

Monitoring
Hydrogen concentration, radiation, etc., are monitored.

Estimated total required storage floor area (per unit by means of
the top-access retrieval Route B method)
In the case of a 口 inner container with dimensions 500 mm ×
300 mm -> 2100 m2②
In the case of a 口 inner container with dimensions 335 mm ×
272 mm -> 1200 m2③
In the case of a 口 inner container with dimensions 450 mm ×
272 mm -> 1300 m2④
In the case of a 口 inner container with dimensions 535 mm ×
272 mm -> 1400 m2①

Equipment and facilities

SafetyfunctionsTemporarystorage building•

Has the filter ventfunction• Has operability thatenables easy and safe
handling• Confinement offine particles using
filter ventConfinement• Manages theexhausts using HEPA
filterHydrogenmeasures• Multiple
thicknesses ofshielding dependingon the contents• Installation of
exhaust or hydrogen
treatment facility

Operability•-

Confinement• Has operability
that enables easy andsafe handling

Opening of lid using remote operations
Retrieval of storage container using remote
operations
Prevention of contamination in carrying-in area bymeans of double door system
Study on various measurements•

Equipment and facilities
Safety functions

Storage container• Maintains the• Multiple thicknesses
function
Hydrogen measures of shielding depending shielding
though shieldingon the contents
thickness••

Hydrogen measures
Prevents the increase in hydrogen concentration by meansof ventilation in the temporary storage building or through thehydrogen treatment facility•• Maintains theconfinement function• Confinement of fineparticles using filter vent-Shielding• Has
operability that • Has operability• Has a fallthat enables easy and prevention
safe handling
function.
and safe
handling

Storage container

Confinement• Multiple
thicknesses ofshieldingdepending on thecontents•

Moisture is removed to reduce corrosion duringstorage.

Various measurements of the storage container are
taken as necessary.
Equipment and facilities

Storagecontainer•

Drying treatment•

The transfer of storage container within the drying
area is done using cranes, bogie, conveyor belt or
unmanned forklift.

Safety functions

Structural
strength

Confinement

Prevention of contamination in carrying-in area•
Contamination in carrying-in area is controlled bymeans of the double door system between the
carrying-in area and the drying area
Carrying-in of storage container•
Removal of transport cask lid using remote
operations•
Carrying-in of storage container using remote
operations•

Equipment and facilities• Has ashielding function(unmannedtransportation for
high-dose items)

Operability enables easy•

Issues• Maintains
shielding
performance using
the shielding
thickness

TransfervehicleTemporarystorage area
Ventilation Monitoring
equipment equipment

Drying device•

Safetyfunctions

Drying area

Measuring
device•Safetyfunctions

10. Temporary storage
Medium surface dose /Contaminated site
Temporary waste storage building

Drying area

Drying area

Acceptancetest device

Unmannedtransfertruck

9. Drying treatment
Medium surface dose /Contaminated site
Temporary waste storage building

Temporary storage
building• Maintains theconfinement function

Temporary storage
building

Corrosion
measures• Shield thickness is
set considering thecorrosion margin.

-

Confinement

Shielding• Maintains theconfinement function• Confinement of fine • Manages theparticles using filter ventexhausts using HEPA
filterHydrogenmeasures• Multiple thicknesses • Maintains theof shielding depending shielding function though
on the contents
shielding thickness

Operability• Has the filter ventfunction• Manages the exhausts
using HEPA filter• Maintains the shielding
function though shielding
thickness

Has the filter vent • Installation of exhaust or
functionhydrogen treatment facility

Study on drying treatment

Storage container-••• Installation of
exhaust or hydrogen
treatment facility

Study on long-term storage measures
Study on monitoring©International Research Institute for Nuclear Decommissioning

26

(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Overview【Confidential】◼
Results
 Overseas hydrogen gas countermeasures for high-level radioactive waste were surveyed, andthe hydrogen generation countermeasures, methods and concepts in each country wereconsolidated.
◼
Goals
 To list down the requirements for containers and storage facilities required for collection,transfer, and storage of high-level radioactive waste.
◼
Implementation details
 (FY2019) A practical inner container# that can be stored in a storage container with multiple
shielding thicknesses depending on the dose level was studied. The concept of transport cask
and storage container with filter vent suitable for countermeasures against gases generated by
the waste generated during fuel debris retrieval was organized. The concept of facilities suchas drying equipment, etc. suitable for measures against corrosion during storage of waste
generated during fuel debris retrieval, and for measures against gas generation assuming thatthe waste will be sealed at the time of disposal, was organized. The measurement methods
and the concept of measuring equipment considering their priority and applicability to the site,
for the measurement items required for the storage in the storage building, were organized.

Fig. 1 Example of transport cask used for high-level
radioactive waste(B type transport cask, USA: TRUPACT- II)(FY2020) Requirements of transport cask and storage container with filter vent suitable forcountermeasures against gases generated by the waste generated during fuel debris retrieval
will be studied. The requirements of facilities such as drying equipment, etc. suitable formeasures against corrosion during storage of waste generated during fuel debris retrieval, andfor measures against gas generation assuming that the waste will be sealed at the time ofdisposal, will be studied. Measurement methods, requirements of measuring equipment,
measuring locations and measurement duration considering their priority and applicability to
the site, for the measurement items required for the storage in storage building, will be studied.

◼
Indicators for determining goal achievement
 (FY2019) A practical inner container that can cope up with multiple shielding thicknesses
should be identified. The concepts of transport cask and storage container with filter ventshould be presented. The concepts of equipment for drying treatment, etc. should bepresented. Measurement methods and the concepts of measurement equipment should bepresented.
(FY2020) The requirements of transport cask and storage container with filter vent should bepresented. The requirements of equipment for drying treatment, etc. should be presented. The
measurement methods, requirements of measuring equipment, measuring locations andmeasurement duration should be presented.

# Inner container: A container for collecting waste from the reactor. To be handled by the
retrieval robot and stored in the storage container.

Fig. 2 Remote controlled type filter vent attachment device(DVS: Drum Venting System) External appearance(Used at the Waste Isolation Pilot Plant (WIPP) in Carlsbad)©International Research Institute for Nuclear Decommissioning

27

(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on efficient shape and usage of inner container (1/3)

【Confidential】<Background and Purpose>

Reactor building● BackgroundLifting machine
Expanded building Lifting machine
Inner container ・・・ Primary container for storing the waste cut from inside
PCV
Waste treatment
cell
Example of functions ・・・ ① Cut waste is stored.
Dose level
Fuel debris
Innermeasurement② Water can be drained due to the mesh
retrievalcontainer
devicestructure.
Manipulator
AccesstunnelFuel debris③ Can be efficiently stored in storage
Device forcontainer.
SampleInner
Radioactivity
Transfer tomeasuringcask
StorageDrying the amount ofcontainer Draining, etc.
analyticalcontainertransfer equipment measuring equipment hydrogenfacility
Studies so far ・・・ In the flow studied until FY2018, it was tentatively
equipment
equipment(gas)
generatedassumed that one inner container will be stored in one storage container.
Fig. 1 Collection, transfer and storage flow
(Fig. 1)
(Excerpts of only sections related to inner container)● Issues① As the storage container uses different inner lining shields depending on
the amount of waste, it becomes necessary to use inner containers ofdifferent sizes.
Table 1 Dimensions of inner containers used in② Also, the Retrieval project reported the dimensions of inner containers
each process of the Retrieval projectsuitable for the equipment used in each retrieval method. (Table 1)
These values were smaller than last year’s assumptions,
Inner container
and it was assumed that multiple inner containers can be stored inside one
Methoddimensions (mm)storage container.

● Purpose
To present a primary proposal for the shape and usage method of the innercontainer suitable for waste storage and management .
Evaluation from the viewpointof storage and management

Side-access retrieval Plan A

Side-access retrieval Plan B

Side-access retrieval Plan C● Study procedure ・・・ As an evaluation index, the number of storage containers
generated is compared.
Top-access retrieval Route① Estimate the number of storage containers generated based on the latest retrievalA
method.
Top-access retrieval Route② Study shapes of inner containers so as to efficiently reduce the number of storage
B
containers generated.
③ Study usage methods so as to efficiently reduce the number of storage containers* “φ400” is a circle with diameter 400 mm, andgenerated.
④ Based on the above studies, study efficient inner container shapes and usage methods. “□500” is a square with length/width 500 mm.
©International Research Institute for Nuclear Decommissioning28

【Confidential】(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on efficient shape and usage of inner container (2/3)< Study Results ①: Study on the inner container dimensions>
Case

1

2

Name of study caseRequired number of Required number of
inner containers
storage containers(containers / unit)(containers / unit)

Details of the study#

Provisional proposed method(An example of top-access retrieval
Route B method)

When top-access retrieval Route B and the innercontainer for the latest retrieval method (□500 x
height 300 mm) are used.

<Study on the inner container
dimensions>

Study on the cases wherein the length, width and heightof the inner container are changed (Following are theexamples of 3 patterns that minimize the required
number of storage containers.)① □335 x height 272 mm② □450 x height 272 mm③ □535 x height 272 mm

10016

1425

① 27216① 807② 14612② 867③ 10180③ 956

With the assumed
method, the storage
efficiency is less and it
requires a large numberof storage containers.

Reduced by at most
43% by changing thedimensions.

# The cask dimensions indicated refer to external dimensions. “□500” indicates a square of 500 mm length and width.
Size of inner container
Provisional proposed method: □500 x height 300 mm
When the size of inner container is changed：□○○ mm x
height XX mm

Table 1 (Reference) Method for calculating the number of storage
containers and the number of inner containers
STEP①

Container thickness(6.4 mm)＋shielding thickness#1＋ gap (25 mm)#1: 4 types of 0, 60, 140,180 mm

It can be filled up to 85%
in the vertical direction
considering a gap (25
mm).

②

③

Calculation details

Calculation formula

Study image1
The inner dimensions are calculated bysubtracting the shielding thickness, container
thickness, and gap from the outer dimension of
the storage container (it is assumed that withthe lid closed, the container can be used up to
85% in the direction of its height.)

Inner length and width of storage container (m) = outer length and width of storage container
Calculating the inner dimensions of – 2 x (thickness of shielding + thickness of storage container + gap)
the storage container
Inner height of storage container (m) = { outer height of storage container – 2 x (shielding
thickness + storage container thickness) – gap } x 85%
Case 1: Value obtained by subtracting the thickness of inner container from the innerdimensions of storage container
Calculating the inner dimensions of Case 2, 4, 5: Value obtained by subtracting the thickness of inner container from the outerthe inner container
dimensions of the collection cask used in top-access retrieval method Route B
Case 3: Decided based on the evaluation results where the dimensions of the inner container
were used as parameters
Along the length: inner length of storage container ÷ outer length of inner container = X(rounded-off)Along the width: inner width of storage container ÷ outer width of inner container = Y
Calculating the number of inner(same as above)
containers stored inside each
Along the height: inner height of storage container ÷ outer height of inner container = Z
storage container(same as above)

Differs on a case-by-case basis, but basically the
inner dimensions are obtained by subtracting
the thickness of inner container (assumed to be2 mm) from the outer dimensions of the innercontainer.

The number of inner containers that can be
stored along the length, width and height iscalculated by dividing the inner dimensions of
the storage container by the outer dimensionsof the inner container.

Number of inner containers stored inside one storage container = XYZ containers

Waste storage rate in the
inner container
Metal waste: 10%
Concrete waste: 50%

Fig. 1 Illustration of the study on the provisional proposed method andthe case when the size of inner container is changed④

Calculating the amount of waste Amount of waste that can be stored per inner container (m3 / unit) W = inner length of innerthat can be stored per innercontainer (m) x inner width (m) x inner height (m) x waste filling rate (metal: 10%, concrete :container (m3 / unit)
50%)⑤

Calculating the amount of waste
Amount of waste that can be stored per storage container (m3 / unit) = Amount stored insidethat can be stored per storage
the inner container W (m3 / unit) x XYZ containers
container (m3 / unit)

Metal fillingrate : 10%

The amount of waste stored per innercontainer is calculated by multiplying theinner volume of the inner container by thefilling rate for each material.

The amount of waste stored per storage
container is calculated from the number of
inner containers that can be stored in astorage container and the amount of waste
stored per inner container.

⑥

Amount of waste (m3)⑦Calculating the number of storage
containers

⑥÷⑤

-

⑧

Calculating the number of innercontainers⑦x③-

The value set for top-access retrieval method Route B is used.

-©International Research Institute for Nuclear Decommissioning

29

【Confidential】(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on efficient shape and usage of inner container (3/3)< Study results ②: Study on the usage method>
Required number
Required number ofof innerstorage containers
containers(containers / unit)(containers / unit)

Case

Name of study case

Details of examination

1<For comparison> Provisional
proposed method(An example of top-access retrieval
Route B method)・ When the inner containers for the top-accessretrieval Route B and for the latest retrieval method(□500 x height 300 mm) are used.
・ No refilling

10016

2

Case where refilling is done

・Waste is refilled into storage container.
・All of the inner containers are discarded.
・It is assumed that the waste inner containers are
stored in a storage container with 0 mm shielding
thickness and are discarded.

10016

1786

3

Case where waste is refilled andcompressed

・It is assumed that the waste inner containers canbe compressed up to 30% along the height under
the same conditions as in Case 2.

10016

952

4

Case where refilling is done, andinner container is reused

・ It is assumed that the waste container is returnedto the PCV and reused 10 times under the same
conditions as in Case 2.

1002

660

1425

The number ofstorage containers
increases in the casewhere only refilling isdone.

Compression orreuse depends onthe method.

Waste

Shre
dding

Stored ininnercontainer

Waste innercontainer# “□500” indicates a square of 500 mm length and width.
Waste

Stored in storage
container(refilling)

To the transfer
and storage
process

Case 2Compression

Reuse
Case 4

Case 3Inside PCV
Legends for the process of waste inner container: Case 2: Case 3: Case 4

Fig. 1 Illustration of the study on the case where refilling is done(compression / reuse of inner container)

Outside PCV: Common< Summary and considerations>▸ As preconditions (such as the amount of waste per dose) arecontinually reviewed, quantitative comparisons of optimal
dimensions and the number of storage containers generated
are not considered to be very important.
▸ The following are the two important findings.
① The number of storage containers is likely to fluctuate greatly
depending on the size of the inner containers, which has alarge impact on storage. However, it is not practical to fix theinner container size at this point. Hence a design that can bechanged flexibly to some extent is desirable.
② In cases where optimization of dimensions is difficult, it could
be replaced from the usage perspective. However, the study
on refilling equipment and reuse routes (return routes to PCV)
becomes necessary and ensuring the traceability of waste also
becomes necessary.
▸ Opinions were exchanged with the Retrieval project and thestudy results were recorded.

©International Research Institute for Nuclear Decommissioning

30

(a) Study and evaluation of storage and management methods(b) ② Study on container and storage facility requirements
Study on measures against gas generation (1/3)

【Confidential】< Study results ③: Required number of filters and volume of void in transport cask>
When storing the storage container, the number of filters that need to be installed to keep the hydrogen concentration in the storage
container below 4 vol% and the volume of void required at that time to keep the hydrogen concentration in the transport cask (sealed) below
4 vol% for 7 days, was evaluated.
Evaluation conditions• The type of vent filter was tentatively selected from those that have a proven track record in the United States and that satisfy thehydrogen diffusion performance.
• It is known that the hydrogen diffusion performance of the vent filter depends on the temperature, and the lower the temperature, thelower the performance. Therefore, the temperature was corrected at 2℃, which is the minimum design temperature of the 1F solid
waste storage vault.
• The maximum transfer period of waste inside the premises was 7 days (in accordance with
The hydrogen release rate required to maintain thethe assumed conditions of the Canister project).
Table 1 Results of evaluating the required number of vent filters and the required volume of void in
the transport cask
Assumed cases

Metal waste topaccess retrieval

Concrete waste
top-accessretrieval

Metal waste sideaccess retrieval

Concrete waste
side-accessretrievalMaterial

Metal

Concrete

Metal

Concrete

Hydrogen generation rate
(m3/d)(Restated)

1.8E-02

1.1E-04

7.1E-05

1.1E-04① Minimum hydrogen
release rate @ 4 vol%(mol/s/mol fraction)

2.3E-04

1.5E-06

9.2E-07

1.5E-06

Type of vent filter② Hydrogen diffusion
performance of vent filter(mol/s/mol fraction

NF016S

NF016LNF016LNF016L

1.43E-04

3.22E-06

3.22E-06

3.22E-06

2

1

1

1

3.25m3 or more

20L or more

10L or more

20L or more

@ 2℃)

Minimum number of ventfilters required
Volume of void intransport cask

hydrogen concentration in the storage container below 4
vol% is calculated using the following formula (based on
concept used in the US) (①)

𝑇𝑑：Hydrogen release rate (mol/s/mol fraction)
Q：Decay heat (Joule/s)
𝐺 ：G value (molecules/100 eV)
𝑁𝐴：Avogadro constant (molecules/mol)
N： Conversion constant (Joule/eV)
XH2：Hydrogen concentration inside the cask (4vol%)The hydrogen diffusion performance (catalog value *) of thevent filter was temperature-corrected by the following
formula (based on concept used in the US) (②)

TD,T = Hydrogen diffusion performance at temperature T
T0 = Temperature at known hydrogen diffusion
performance* Test results at 21℃ (294K)

A number where ② > ①.
The volume of void required to keep thehydrogen concentration in the transport
cask below 4 vol% for 7 days.
©International Research Institute for Nuclear Decommissioning

31

(a)Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on measures against gas generation (2/3)

【Confidential】< Filter events that need to be considered and their countermeasures (1/2) >The filter events that need to be considered and their countermeasures were studied by referring to the processflow.
During transfer (including during transfer
Work step
During temporary (long-term) storage
within the premises)
Medium surface dose rate /contamination site

Low surface dose rate /contamination site

Temporary waste storage building
Building
Drying area

Illustration of
the operation

Carrying-in/out
cell

Ventilationequipment

Monitoring
equipment

Pre-delivery test
device(1) Clogging of the filter
The contents rise during the transfer, clog
the filter, which prevents the release ofhydrogen.
Filter events
that need to beconsidered

Temporary storagearea(2) Contamination due to scattering ofcontents
Fine particles or gas components released
from the filter contaminate the inside of thebuilding and the inside of the transport cask.

(3) Corrosion and deterioration of the filter
The filter corrodes or deteriorates due tolong-term storage and does not satisfy thehydrogen diffusion performance anymore (theside where the opening narrows). Or thecontents are more likely to scatter (the side
where the opening widens).
(4) Performance degradation due to filter
condensation
Moisture inside and outside the container
condenses on the filter part, blocks the filter
and the filter does not satisfy the hydrogen
diffusion performance anymore.
©International Research Institute for Nuclear Decommissioning

32

(a)Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on measures against gas generation (3/3)

【Confidential】<Filter events that need to be considered and their countermeasures (2/2)>
Work step

Proposedmeasures#

During transfer

During temporary (long-term) storage(1) Clogging of the filter•Install multiple filters (multiplexing) so that even if one
clogs, hydrogen can be released through the others.
•Make the mounting structure of the filter like a labyrinth
so that the contents do not reach the filter directly.
•As it is assumed that the filter gets clogged due to thecontents flying during transportation, use a lid with afilter, which is exclusively used during transportation(during storage, replace it with a lid with a filter, which is
exclusively used during storage).
•Evaluate the effect on storage container in the event ofan explosion at the expected hydrogen concentration (4vol%).

(3) Corrosion and deterioration of the filter•Select SUS316L or carbon composite
materials that are corrosion-resistant in
atmospheric environment for the filter media(sintered porous filter).
•Check corrosion resistance with test piecesof the same material under the same
conditions (or harsh conditions).

(2) Contamination due to scattering of contents•For fine particles, a particle collection efficiency
equivalent to that of HEPA (99.97% particle collection
rate for 0.3 μm particles) is required.
•Gas components are taken care of with a ventilation
system.
•Set up areas classified (red, yellow, green) based on theassumed scattering from filters to control contamination.
•Carry out surface contamination test (decontaminate if
necessary) of the storage container before transferringinside the premises and before storage.

(4) Performance degradation due to filter
condensation•Control the temperature and humidity bymeans of the ventilation system of thestorage facility to prevent condensation.
•If the ventilation system of the storage
facility stops functioning and condensation
occurs, it is necessary to restore theventilation system and eliminate
condensation, before the hydrogenconcentration in the storage container
exceeds 4 vol%. Ensure that the hydrogenconcentration inside the storage container
does not exceed 4 vol% during the time thatis required for restoring the ventilation
system. If it exceeds (the time margin cannot
be secured), ensure the functioning of theventilation system at all times bymultiplexing the ventilation system or byarranging for emergency power supply.

# The applicability of the vent filter and the extent of countermeasures depends on the negotiations with the regulatory authorities.
©International Research Institute for Nuclear Decommissioning

33

(a)Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on drying treatment (1/3)

【Confidential】[FY2019 Implementation details](1) Investigation of the existing drying treatment technologies (including consolidation of conditions)(2) Study on the requirements of drying treatment equipment(１) Results of investigating the past drying treatment technologies➢ The preconditions for drying treatment were studied. (Table 1)➢ Based on the basic principle of drying, 8 methods were identified by combining heat transfer forms (4 methods) and waste transportation (2methods). (Table 2)➢ The above methods were compared#1, and it was confirmed that the static (convection, conduction) method may be applicable to existing
technologies, and multiple heat transfer forms (convection, conduction, radiation) can be used together.
#1: Significant survey results were obtained showing that the methods can be applied to bulky metal and concrete waste (with due consideration of shape, moisture,radiation), interaction with other studies (container, storage building), items to be studied for container corrosion countermeasures and the impact (dust, hydrogen)during the drying process.

Table 2 Results of investigating existing technologies related to drying treatment(Results of comparative study of the eight identified methods)

Table 1 Preconditions for the drying treatment
Major
items

Minor
items

Propertiesof thewaste
targeted

Status of
raw
waste

・Items removed during debris retrieval(Wet due to the cooling water in thereactor, draining assumed, bulky
shape)

Treatment fortemporary storage

・The raw waste generated during debris
retrieval is cut inside the PCV into the sizeof the inner container or smaller, it getsdrained by its own weight, and notreatment other than drying is performed
before primary storage.

Investigation items

Compatibility withother studies

Items compared

Inner container
Shape

Investigation results
Static: Can be static or stationary inside the drying equipment
Conduction - transfer: Cannot be stationary inside the drying equipment

Bulky shape waste collected

Convection - static / transfer, conduction / radiation - static: Does not require additional processing and applicable to bulky shape.
Conduction-transfer: In addition to retrieval from inner container, it requires additional processing.
Radiation-transfer, microwave- static / transfer: If necessary, additional processing is required after removing it from the innercontainer.
No water penetration / applicable to slightly bulky shape

Reduction index

Absence of moisture: Possible up to equilibrium moisture content
Presence of moisture: For each method, there is a limit to the moisture content and a depth from the surface from which it can be
irradiated.

Moisture

Static, convection radiation / microwave-transfer: No limitation
Amount brought-in
Conduction-transfer: Has limitation

Considerations related to radiation

Container

Storage building

Study items required for container corrosion
measures

Impact during thedrying treatment

・For hydrogen measures, there are theoptions of completely dry (sealedcask) or vent type (unsealed cask).
・Drying in the storage building mainly
contributes to the confinement of thecask (corrosion countermeasures).
(Refilling is made possible within thecontainer, until container corrosion
measures during storage become
feasible.)

Applicability to bulky metal or concrete waste

Application
to retrieval
flow

Conditions

Static: Requires shielding depending on the amount handled, and the number of sites to be maintained is less as there are only few
drive parts and interior parts inside the equipment.
Transfer: Requires shielding depending on the amount handled, there are a large number of sites to be maintained as there aremany drive parts and interior parts inside the equipment.
Static: Dimensions and structure of inner container, transportation method of inner container
Convection/radiation/microwave-transfer: Those said above +Transfer method inside the drying equipment
Conduction-transfer: Method of retrieving wastes from the inner container, method of storing inside the inner container after
dryingCommon methods: Air supply / exhaust system, power (electric power) system

Common methods: Status of adhesion of free water on target waste (metal, concrete)
Corrosion margin based on the amount of residual water after drying treatment

Dust (solid)

Common methods: Requires the confinement of air supply / exhaust systems (filter), including the generation of airflow in thedrying equipment

Hydrogen

Convection – static / transfer, conduction / radiation - static: scavenged to the outside of the system by exhaust (including vacuum)
Conduction-transfer: Requires confirmation on the complexity of the structure of the drying equipment (difficult to scavenge) andthe possibility to control temperature of the heat transfer part (hydrogen ignition prevention).
Radiation-transfer, microwave-static/transfer: Though it is scavenged outside the system through the exhaust, requires confirmationon the possibility to control the temperature of the waste (hydrogen ignition prevention)

Blue notation: Details that meet the conditions

Red notation: Details the are difficult to or can’t meet the conditions.

©International Research Institute for Nuclear Decommissioning34

【Confidential】(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on drying treatment (2/3)(２) Results of studying the dry treatment equipment requirements (1/2)➢ The Atomic Energy Society of Japan Standard "Technical requirements and test method for manufacturing sub-surface industrial waste package:
2015”#1 was referred to for the requirements for burial disposal of radioactive waste packages within Japan, and the concept of drying treatment for thewastes in this study were compared and organized#2.
➢ As a result of the above study, it was confirmed that among existing technologies, the convection (static) and conduction (static) methods are
expected to be applicable to the requirements for drying treatment, response to drying treatment procedure of each container, and to the concept ofoperation management indicators.
#2: It was confirmed that the following 7 points could be dealt with.
・Conditions of target waste to be considered
・Method for removing free water・Method for drying free water
・Evaluation of appropriate removal or drying method
・Requirements that control the absence of a large amount of free water・ Handling of the drying treatment procedure for each container
・ Operation management indicators ortreatment conditions(Free water described in the Atomic Energy Society of Japan Standard refers to the water that is brought into the container together with the radioactive waste when storing the radioactive
waste in the container.)
Table 1 Comparison of the Atomic Energy Society of Japan Standard
and the concept of drying treatment of the waste generated during fuel debris retrieval (excerpts)

Table 2 Comparison of existing technologies and drying treatment requirements #5

Form of heat
Concept
related to drying
treatment in the Atomic Energy Society of デブリ取出し時の発生廃棄物
日本原子力学会標準「余裕深度処分対象廃棄体の製作要件及びConvection
Conduction 伝導
Radiation 放射
Microwave
伝熱形態transfer
対流
マイクロ波
Concept of drying treatment of
方Method＊1
Japan
Standard “Technicalrequirements and test method forの乾燥処理の考え方
検査方法：2015」における乾燥処理に関連する考え方
Transport
of式
waste generated during debris
StaticTransfer
Static静置Transfer 移送 Static
Transfer
StaticTransfer
廃棄物の搬送
静置
移送
静置
移送
静置
移送
*1
waste
manufacturing sub-surface industrial waste package: 2015”
retrieval
項目
PCV内に存在する対象廃棄物の条件のうち、種類、形状、貯蔵場所(現状は取出し前のためPC
V内にある)
Items
Conditions of Among the conditions of the target waste existing inside the PCV, it is possible to create a plan for information such as type, shape, storage
対象廃棄物の条件
1 適用範囲事業者が余裕深度処分対象として、廃棄体の製作 Treatment
取出し時の発生～一時に関する情報については、計画を作成することが可能である。しかし、自由水の状態は不確かである。
target waste location (currently
it is inside the PCV as it is yet to be retrieved). However, the state of free water is not certain.
from the time of
Methods used by operators to review conformance with
における要求事項や技術基準への適合性検保管するまでの処理●方式に従い、自由水が除去可能●同左●方式に従い、自由水が除去可能
generation
during retrieval up to
●Free
water
can
be
removed based on the removal ●Same asthat●Free water
can be removed based on the removal
requirements
and technical standards during waste
自由水の除去方法
Method of
1. Scope of●除去時間は、水分持込量に依存●制約有methodon the left
method ●除去時間は、水分持込量に依存
査の方法
conditioning
for marginal depth disposal.
removing free
temporary storage
●The
time
requiredfor
removaldepends
onthe●Presence
of
●The
time
required
for removal depends on theapplication (Does not include technical requirements from waste disposal
water
乾●既往技術で適●機器内の ●既往技術で●追加加工要等
amountof water brought-in
restrictions
amount of●廃棄物温度の遠隔測定要等
water brought-in(埋設施設や輸送からの技術要件を含まない)(廃棄体化に向けた処理は該当(Not
applicable to the treatment
燥
保守部位多●Existing
用可
適用可facilities or pertaining to transportation)●Larger number●Requiresforしない)waste conditioning)の 自由水の乾燥方法 technologies
can●同左 ●Existing●同左 additional●条件及び保持●移送条件が複of maintenance●条件及び保持時間設定可技＊4
Method for be applied.
technologies can●Requires
remote measurement of waste
4.1 要求事項
一時保管に密閉容器を用いる
From余裕深度処分報告書
the marginal depth ＊3
disposalreport*3 andburial雑化
sites within theprocessing
時間設定可，埋設事業告示(表D.1)から、
drying the free ●Conditions andbe applied.
temperature術
Technological
The
free
water
is
removedbeforeequipment*4 (Table D.1), it is●Transfer(附属書D(参考))disposalproject
declaration
場合の水素ガス発生や腐
water
retention
●Same
as
that
●Conditions
and retention duration can be set.
requirementsfor
要
自由水除去は「廃棄体の十分な強度」と「汚染temporary storage to prevent●既学会標準の考え方が適用可●乾燥機器内の●既学会標準への温度の考え方
4.1
conditions
get●Same as that
適切な除去方法又 duration
can beon the left
件 drying
が適用難等
on the left
complicated
移送モデルが複雑that the removal of free water contributes
Requirements interpreted
食発生防止（容器の密閉拡大の防止措置(容器の密閉性)」に資すると解set.
は乾燥方法の評価
hydrogen gas generation and化(Attachment D to the "sufficient strength of waste packages" andEvaluation of●The transfer
釈性）のために一時保管前に自corrosion(container sealing)
appropriate(reference))model inside the
伝導●運転条件(温度)の設定が複雑化●The
concepts from the existing Atomic Energy●The temperature
concept from the Atomic Energy●運転処理条件及び必要保持時間が設定可●乾燥機器内の多量の自由水を含“measures to prevent the spread of contamination
removal methods
drying facility
由水を除去
静置
when
using a sealed container inSociety of Japan Standard can be applied.
●Society of Japan Standard is difficult to apply.
移送時間の設定
んでいないことの管and drying
gets(sealingofcontainers)”.
4.2.1 容器に封 廃棄体内に自由水が残留する場合、容器密閉前に自由(廃棄物の強度や処分に向けたが複雑化
complicated.
temporarystorge. (Does not
理要件 methods
Similar to と同じ
Requirements for●Setting of入する場合の自完全乾燥は含まない)
4.2.1 Removal If free水除去をする(「真空乾燥」及び「真空乾燥以外」に区分しincludethe complete drying forwater remains in the waste package, the free water isconduction- static
ensuring that a ●対応可
transfer time
●同左
●同左
●内容物の取出し●容器が照射へ与える影響確認●Operation conditions and required retention time
●The settingof operation conditions (temperature)of free water
由水除去
て規定)
large
amountofwithinthedrying要
removed before sealing the container (classified as “vacuum
the strength of waste packages or 容器単位の乾燥処理手(汚染拡大
要can be set.
gets complicated.
when enclosing
free water is notfacility gets
防止設備が
順の対応性
for disposal)
present
complicated.
in a container drying” or “other than vacuum drying”)＊1：日本原子力学会標準「余裕深度処分対象廃棄体の製作要件及び検査方法：2015」（原子力発電所、再処理施設から*1: The Atomic Energy Society of Japan Standard "Technical requirements and test
method for manufacturing sub-surface industrial waste package: 2015” (Standard for
operators to perform burial disposal of radioactive waste generated from nuclear power
plants and reprocessing facilities by means of waste conditioning.)*3: Safety regulations related to marginal depth disposal of low-level radioactive waste(January 2008, Advisory Committee on Energy and Natural Resources, Nuclear andIndustrial Safety Sub-committee, Waste Safety Sub-committee)*4: The Regulation related to the Project for burial of type 2 waste resulting from nuclear
fuel substances or from items contaminated by nuclear fuel substances. (Burial Project
Regulations)

運
運転条件
転
Handling of drying treatment
管procedures of each container
理
指 残水量
標 運転時間Operation
management
index

Operation
conditions●設定可●Can behandled.

●設定可●Can be set.

Amount of●Can be set.
residual water●Same大型化)
as thaton the left
(Increase in thesize of equipment ●Same as thatfor theon the leftprevention of
spread ofcontamination)●運転条件(温度)の設定が複雑化●Requires
retrieval of●移送時間の設
contents

定が複雑化●The setting oftransfer time getscomplicated.

●The impact of the container on irradiation needs tobe confirmed.

●運転時間(保持時間)に際し、容器が照射へ与える影響確認要●The setting of operation conditions (temperature)
gets complicated.
●Impact of container on irradiation during operation
time (retention time) needs to be confirmed.

*5： Below is the distinction between the colored characters shown in the table.
Blue notation: Details that meet the conditions
Red notation: Details the are difficult to or can’t meet the conditions.

©International Research Institute for Nuclear Decommissioning35

【Confidential】(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on drying treatment (3/3)(２) Results of studying the dry treatment equipment requirements (2/2)➢ The requirements for the drying treatment facility were created based on the results of investigating the existing technologies and by
consolidating the requirements.

Table 1 Requirements for the drying treatment facility
項 目 Items

対 象 廃 棄物 に 対 する適 用 性Applicability to thetargeted waste

形 状 、
性 状

stored in them).

デ ブリ取 出 しに 伴 い 発 生 廃 棄 物 に よ って 、持 ち 込 ま
れ る自 由 水 の 持 込 量 を一 時 保 管 可 能 に な る ま
で 低 減 可 能 で あ ること

Shape, properties

放 射 線 に対 す る考慮

廃 棄 物 の 取 扱 量 に 応 じた 遮 へ い が 可 能 で あ り、処理 設 備 内 の 駆 動 部 や 内 装 品 が な い /又 は 保 守 が 同 上
可 能 で あ ること

Applicability to retrieval flow

The raw waste generated during fuel debris retrieval should be cut inside
the PCV into the size of inner container or smaller, get drained by its own
Based on the collection, transfer andweight, transported and received at the drying treatment facility. In thestorage flow
drying treatment facility, only the drying before the temporary storage
should be performed.

水 素 対 策 は 、密 閉 容 器 又 は 非 密 閉 容 器 とし、保管 建 屋 で の 乾 燥 処 理 は 主 に 容 器 の 閉 じ こ め (腐食 対 策Hydrogen) * 1 のmeasures
た め にshould
実 施be classified
す る for sealed and non-sealed

対 象 廃 棄 物 に 応 じて 検
討

* 1 ： 保 管containers,
中 の 容 器and腐the
食 drying
対 策 treatmentの 成 立 可in the
能 性storage
目 途 buildingが 得 らshould
れ る まmainly
で
be carried out for the confinement of the container (corrosion measures) #1 Studied based on the targeted waste
は 、 容 器*1:内It should
廃 棄 物beをmade
詰 替possible
可 能 toと すrefillる waste
こ と in the container, untilcontainer corrosion measures during storage become feasible.

容 器 （ 寸 法 、 構 造 、 搬 送 方 法 ） 及 び 保 管 建 屋 (給容 器 や 保既 往 技 術 調 査 よ り、適気 / 排 気 の 系 統 、 動 力 の 系 統 )と の 取 合 い が 可
管 建 屋 へ用 可 能 性 の あ る技 術 を
能
で
あ
る
こ
と
Compatibility with
The applicable technology is selected byの container
取 合 いand storage The container (dimension, structure, transport method) and storage
抽 出 on the existing
investigatingbuildingbuilding (air supply / exhaust and power systems) should be compatible.

technologies.

汚 染 拡 大 防 止 の た め 、乾 燥 処 理 設 備 の 給 気 /排
気 系 統 で 放 射 性 物 質 (ダ ス ト )の 除 去 (フ ィ ル タ )を
乾 燥 処 理 ダ ス ト (固 行 う こ とTo prevent the spread of contamination, the radioactive substances (dust) 同 上
中 の 影 響 体 ) 、 水 素 乾 燥 処in the
理 air
中 supplyに 発/ exhaust
生 す るsystems
水 素of theは drying、 設 treatment
備 内 に facility
滞 留should
せImpact during drying
Dust (solid), hydrogen be removed (filtered).
treatment
ず に 系The外hydrogenへ 掃 generated
気 す るduring
こ と the drying treatment should be scavenged Same as aboveoutside the system without allowing it to remain inside the facility.

Items

乾 燥 処 理 の 前 後 で 容 器 及 び 乾 燥 処 理 設 備 で の
放 射 性 物 質 の 閉 じ 込 め が 可 能 で あ る こ と
材 料 ：炭 素 鋼 SS材
容 器 の 構 造 ：1F構 内 で 使 用 中 の 角 型 容 器 に 汚 染
拡 大 防 止 措 置 が 可 能 な フィル タベ ン トの 取 付 け を検討 中

The confinement of radioactive substances in the container and at thedrying treatment facility before and after the drying treatment should be
Measures for preventing possible.
the spread ofMaterial: Carbon steel SS material
contamination
Structure of the container: Attaching a filter vent capable of preventing the
spread of contamination on a square shaped container that is being used in
1F premises is being studied.

It should be possible to reduce the amount of free water carried in along
The applicable technology is selected bywith the waste generated during fuel debris retrieval, to a level where it caninvestigating the existing technologies.
be temporarily stored.

デ ブリ取 出 し作 業 時 に 発 生 した 原 廃 棄 物 は 、PCV内で 内 容 器 サ イ ズ 以 下 ま で 切 断 され 、自 重 に よ る
水 切 りをした 状 態 で 搬 送 され 乾 燥 処 理 設 備 が 受 け
一 時 保 管 に 向 け た 処
収 納 ・移 送 ・保 管 フ
入 れ る 。乾 燥 処 理 設 備 で は 、一 時 保 管 前 の 乾 燥理
ロー に 基 づく
以 外 の 処 理 は しな い

取 出 しフロー へ の 適
用乾 燥 処 理 設 備 の 要 求 事 項
備 考
Requirements of drying treatment facility
Remarks
種 類 、 形 状 、 貯 蔵 場 所 、 自 由 水 の 付 着 状 態 が設 定 可 能 で あ る こ と( 自 由 水Itofshouldの 付be着possible
状 態to setは the
、 現type,時dimension,
点 で のstorage
推 定 location
は 困 and
難 stateでfree water adhesion.
対 象 廃 棄既 concepts
学 会 標in the準 existingの 考 Atomic
え 方TheConditions of targeted(Estimation
of
free
water
adhesionstate
isdifficult
at
this
point
in
time,
あ り 、 既 学 会 標 準 ( 附 属 書 E ) の 考 え 方 に 倣 い 、 金 and Energy Society of Japan Standard can be
物 wasteの 条 件
が 応 用 可is set as metal: management category I, concrete: management category I
applied.
属 ： 管 or理 II (considering
区 分 Ⅰ ，theｺ moisture
ﾝ ｸ ﾘ ｰ ﾄinside)： 管 based
理 区 on分theⅠconcepts
又 は inⅡthe( Atomic
内 部の 含 水Energy
を 考Society慮 ) とof Japan設 定Standard))

既 往 技 術 調 査 よ り、適
用 可 能 性 の あ る技 術 を抽 出

It should be possible to have shielding depending on the amount of waste
Considerations related
handled, and there should either be no drive parts or internal parts inside Same as above
to radiation
the treatment facility OR their maintenance should be possible.

Treatment for temporary storage

項 目

乾 燥Requirements
処 理 設 備of dryingの 要treatment
求 事 項facility
備 考
Remarks
デ ブ リ 取 出 し に 伴 い 発 生 す る 廃 棄 物 (炉 内 冷 却 水 で濡 れ た 状 態 で 、水 切 りされ た バ ル ク形 状 の 金 属 及 想 定 廃 棄 物 の 性 状 設
び コ ン Itクshould
リ ーbeトpossible) が 、 to保handle
管 容waste器 generated
単 位 (during
内 容fuel器debris
収 retrieval納 )で 定 に 基 づ く(bulky metal or concrete, which got wet due to the reactor cooling water
Based on the properties set for the
取 扱 いand可was能drained)
で あ byる means
こ と of storage containers (with inner containers assumed waste

汚 染 拡 大防 止 措 置
乾
燥
の
技
術
要
件

Technologicalrequirements fordrying

The facility should dry free water that can impair confinement of sealedcontainers.
There should not be any material that significantly promotes corrosion(Waste acid: hydrogen ion index concentration of 2.0 or less, waste alkali:hydrogen ion index concentration of 12.5 or more) or that generates
腐 食 を 著significant
し く 促 進gasさ(pyrophoric
せ る 物 質substance,( 廃 酸 ： 水water
素 イreactive
オ ン 濃substances).
度 指 数 2 .0 以

Substances that may
cause damage to thesoundness

健 全 性 を損 な うおそ れ の ある 物 質

on the properties set for the
密 閉 容 器 の 閉 じ 込 め を 損 な う お そ れ の あ る 自 由 Basedassumed waste
水 を 乾 燥 す る 設 備 で あ る こ と
想 定 廃 棄 物 の 性 状 設
定 に 基 づく
下 、 廃 ア ル カ リ ： 同 1 2 .5 以 上 ) や ガ ス 発 生 を 著 し く 発 生 さ せ る物 質 (自 然 発 火 性 物 質 、 禁 水 性 物 質 )は な いMethod of removing
free water, drying
method and its
evaluation

The technology that can be applied to
the requirements of drying treatment
It should be possible to remove free water using the methods in existing
shall be identified through investigation
technologies, and it should be possible to evaluate the drying method.
of existing technologies.
(The time required for removal depends on the amount of water carried in
The concept described in the Atomic
along with the waste, and the conditions and retention duration can be set.)
Energy Society of Japan Standard
can be applied.

自 由 水 の 既 往 技 術 の 方 式 に よ り 、 自 由 水 が 除 去 可 能 か
除 去 方
つ 、 乾 燥 方 法 が 評 価 可 能 で あ る こ と
法 、乾 燥
方 法 、 評 (除 去 時 間 は 水 分 持 込 量 に 依 存 し 、 条 件 や 保 持 時
for
価 Requirements
間 が 設It should
定 可 be能possible)
ensuring that a large
to set the operation and treatment conditions andamount of free water is required retention time.
not present

既 往 技 術 調 査 に て 、乾
燥 処 理 の 要 求 へ 適 用
可 能 性 の あ る 技 術 を抽出
既 学 会 標 準 の 考 え 方が 応 用 可Same as above

多 量 の 自由 水 を含運 転 処 理 条 件 及 び 必 要 保 持 時 間 が 設 定 可 能 でん で い な同 上
あ る こ とい こ とのdrying treatment should be performed for individual containers after
管 individual
理 要 件containers The
Drying treatment forSame as abovestoring the radioactive substances (waste) inside the container.
容 器 単 位 の 乾 燥 処
容 器 に 放 射 性 物 質 (廃 棄 物 )を 収 納 後 、 容 器 単
同 上理
位 で 乾 燥 処 理 を実 施 す る こ と
残 水 量
放 射 性It should
廃 棄be 物possibleの 条to set件the( 管amount理 区of residual
分 ) にwater
応 (representative
じ 、 既 往運 Amount
転 条 of件residual
技 術 のvalue),
方 operating
式 ( 真 conditions
空 乾 燥 (pressure,、 真 空temperature,
乾 燥 以 amount外 ) でof exhaust
残 水 /Operation
運 転 管 理waterand operation time (retention time determined based on the
management
量 ( 代 ventilation)
表 値using) 、the運methods
転 条 in件existing( 圧 technologies
力 、 温 度(vacuum、 排 drying,
気 / other
通 風 Same
同 上as above
Operating conditions
tests)
指 標
indicators
運 Operation
転 時 間time 量 ) 、 運than転vacuum
時 間drying)( 試 depending験 に よonっtheてcondition
決 定 ofし radioactive
た 保 持 waste時(management classification) .
間 )が 設 定 可 能 で あ る こ と©International Research Institute for Nuclear Decommissioning

36

(a)Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on measurement methods and measurement equipment (1/3)

【Confidential】<Background and Purpose>● BackgroundStudies conducted so far ・・・ In the collection, transfer and storage flow studied up to FY2018, it was assumed that measurement would be
performed, however, the measurement items, measurement methods, measurement equipment, etc. were not studied.

● Issues① Inputs such as properties of waste, requirements for transfer within the premises, storage requirements, etc., were not determined.
② There are restrictions on measurement time, space for equipment installation, etc., for making the entire fuel debris retrieval process
feasible (quantitative design requirements were not determined).
③ It is not practical to measure all the confirmation items. A logic for selection is required.

● Purpose
To draw up specifications of the measurement equipment (listing the requirements of the measurement equipment(measurement method, measurement location, measurement time, etc.)).

● Study procedure
1. From the process flow studied in FY2018, organize
the confirmation items that are likely to be measured
and investigated.

4. Investigate measurement methods andmeasurement equipment for high-priority measurement
items.

2. Study the priority for measuring the organizedconfirmation items.
5. Review the measurement flowchart and organize therequirements for each measurement.
3. Create a primary draft of the measurement flowchart
based on the priority of each measurement item.
1 to 3: Implemented in FY2019, 4 to 5: Planned to be
implemented in FY2020
Fig. 1 Process for studying the measurement methods and measurement equipment©International Research Institute for Nuclear Decommissioning

37

(a)Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on measurement methods and measurement equipment (2/3)
1. From the process flow studied in FY2018, organize
the confirmation items that are likely to be measured
and investigated.

<Confirmation items>

2. Study the priority for measuring theorganized confirmation items.

<Purpose and intent of the confirmation items>● When stored in the inner container

Purpose and intent

Basic properties

Requirements for acceptance at thedelivery facility
Requirements for transfer within thepremisesRequirements for acceptance at thestorage vault
Incorporation in the process up to storage
Incorporation in the process after storage

Dimensions, capacity, weight, surface
condition

Waste properties (composition)
Material, water, organic matter, corrosive
substancesRadiochemical properties
Surface dose rate, the amount of radioactivity, amountof nuclear material

Heat generation characteristics
Hydrogen generation
properties● When storing in storage containers

External appearance
WeightFilling rate
Surface dose rate
Surface contamination
Surface temperature
Hydrogen generation rate
Amount of nuclide emission● When stored in transport cask
External appearance
WeightSurface dose rate
Surface contamination
Surface temperature
Sealing【Confidential】<Measurement plan for confirmation items>
Proposal for measurement
technology
Details and targets formeasurement
Implementation method
Implementation period
Place of implementation
Required value
Assumed issues

Priority based on the purpose
Are these items that require to be confirmed,
such as requirements for acceptance at thefacility, requirements for transfer within thepremises?
Difficulty in measurement

Is it technically likely to be measurable
during the process up to storage?
Can it be estimated from other measurements?

Is it relatively easy to estimate fromalternative measurement items?
Possibility of avoiding measurement throughoperation, evaluation, preliminary investigation, etc.

Is it possible to substitute themeasurement by operation, etc.?

<Policy for setting the measurement flowchart>
Based on each priority, the measurement items are set based on the following policy.
▸The confirmation of items up to the collection, transfer and storage such as requirements for
acceptance at the facility, requirements for transfer within the premises, etc., shall beconsidered essential, while the items required for the processes after storage shall not be
considered essential.
▸ Among the essential confirmation items, with respect to items for which measurement can
be avoided through estimation from other measured values, or through operation, evaluation,
preliminary investigation, etc., the policy shall be to avoid measurement, and the items forwhich measurement cannot be avoided shall be measured.
©International Research Institute for Nuclear Decommissioning38

【Confidential】(a) Study and evaluation of storage and management methods② Study on container and storage facility requirements
Study on measurement methods and measurement equipment (3/3)
3. Create a primary draft of the measurement flowchart
based on the priority of each measurement item.
Confirmed at the time of planning(Preliminary Investigation and evaluationof the objects to be removed)<Concept>●If the number of measurement items increases, there is concern that themeasurement time and equipment installation space will affect the overall plan for fueldebris retrieval.

Estimated from surface dose rate (Ifan estimation method cannot beestablished, it is confirmed throughmeasurement, etc.)

Presence of organic matter
Presence of corrosive substances
Amount of radioactivity (noble gas)

●Therefore, the policy shall be to avoid measurement as far as possible for items thatcan be estimated from other measurement results, items that are likely to be evaluatedin advance, and items that can be confirmed based on their external appearance(camera).

Calorific value / surface
temperature
Amount of radioactivity (αβγ)
Amount of nuclear material
Hydrogen generation rate

Confirmation based on external
appearance (camera)

Ensured by meansof vent filter design

Dimensions / external appearance
Surface condition
Material (concrete / metal)

Reactor building

Amount of nuclide
emission

Ensured through draining
Presence of moisture
content / amount ofmoisture content
Expanded building

External appearance
Amount of nuclide emission(Are the lid and vent properly
installed?)

External appearance
Sealing performance(Is the lid properly
installed?)

Confirmation using
measuring device

Confirmation using
measuring device

WeightSurface dose rate

WeightSurface dose rate
Surfacecontamination

WeightSurface dose rate
Surface contamination

Lifting machine

Waste container
handling cell

Dose levelmeasurement

Inner
container

Confirmation based on external
appearance (camera)

Confirmation using
measuring device

Waste treatment
cell

Fuel debris
retrievalequipment

Confirmation based on external
appearance (camera)

Delivery cell
Deliveryequipment(includes
surface dose
measurement)

Storage
Surface dose container lid
measurement
closingequipment

Transport
cask lid
closingequipment
Transfer inside thepremises

Manipulator
Fuel debris

Innercontainertransfer
equipment

Draining
equipment

Radioactive
substanceetc.,
measuringfacility

Device for measuring
the amount ofDryingequipment hydrogen (gas)generated

Sample cask

Transfer toanalytical facility

Storage container
with filter vent

Storagecontainertransfer
equipment

Transport cask

Transport cask transfer
equipment

Transfer bogie

Transfer truck

Fig. 1 Primary draft of measurement flowchart<Summary and future plans>▸ There is a possibility to estimate multiple confirmation items from the surface dose rate of thewaste (stored in inner containers). The challenge is to develop an estimation method.
▸ In the future, the measurement methods and equipment for the confirmation items with highmeasurement priority identified in this year's study will be investigated. In addition, themeasurement flowchart is planned to be updated based on the investigation results.

・・・Items for which it is believed that measurement can be avoided
through operation, evaluation, and preliminary investigation
・・・Items that could be estimated from other measurements
・・・Items that are believed to require confirmation using measuring
device©International Research Institute for Nuclear Decommissioning

39

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Overview Am-241
radiation source◼ Results
 In FY2018, a prototype (Fig 1) was developed as a measuring
equipment that can measure alpha contamination remotely (Alphacamera) keeping in view the site environment and an element test was
performed. Issues related to this method were identified based on theelement test results and scope of application was specified.
◼ Goals
 To develop a scanning mechanism such as pan-tilt mechanism etc. as ameasuring system for alpha contamination settled on the surface of theequipment and building concrete, and also develop a measuring systemcomprising image synthesis function that creates 3-dimensional
Synthesize
contamination distribution. To design and develop systems for on-site
contamination
distribution in camera
application and conduct functional verification testing. In addition, to
image
conduct mock-up test for alpha contaminated equipment on the site andReceiver of signals
carry out performance evaluation.
(ultra-violet rays)
produced by alpha rays
◼ Implementation details
Fig. 1 Prototype developed in FY2019
 In FY2019, the prototype was improved based on the results of FY2018
element test and element tests continued to be performed. In addition,Alphacameraアルファカメラpreliminary systemization test was conducted and the data required fordesigning the measuring system was collected. (Fig 2)
 In FY2020, a system based on the design will be developed andControlunit of
運搬台車装置の制御装置
transport trolley
functional verification test for the system will be performed. Also, aequipment
mock-up test will be conducted for alpha contaminated equipment on Tilt チルト機構
アルファカメラのPower
supply unit formechanism
電源ユニットthe site and performance evaluation will be carried out.
alphacamera◼ Indicators to determine goal achievement
Pan
パン機構
 The data required for design completion should be acquired. (FY2019) mechanism Performance evaluation should be conducted by means of a mock-up(Tilt drive(Wiretype potentiometer)(Pandrive motor) （チルト駆動モータ）
test. (FY2020)（パン駆動モータ）
ワイヤ式ポテンショメータ
motor)
Fig. 2 Sample image of alpha ray contamination
measuring system©International Research Institute for Nuclear Decommissioning40

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Principle of alpha camera Detector (PMT)

Measurement principle① Reaction takes place with nitrogen within alpha radiation range(several cm) to produce several 100 photonic ultraviolet rays.
② These ultraviolet rays are focused and imaged with a lens andthe distribution of alpha rays is measured from the distribution of
the number of photons reaching the light detector.
③ PMT (Photomultiplier Tube) is shielded as it is likely to beaffected by Cherenkov radiation when gamma rays enter PMT.

Visible light filter
MirrorShield for gamma rays

Lens
Ultraviolet rays

Alpha rays
Light Emission

Uranium etc.

Fig. 1 Pattern diagram showing principle

Light emission in theultraviolet region

Example of presentations made in connection with alpha 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 forremote measurement of alpha radiation
4. Autumn meeting of Atomic Energy Society of Japan in FY2014, Technology forremote measurement of alpha radiation–Application to lighting environment-

Emission spectrum
Fig. 2 Nitrogen emission spectrum
Cited from J.Sand. Remote Optical Detection of Alfa Radiation.
IAEA－CN-184/23.
©International Research Institute for Nuclear Decommissioning41

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Alpha camera development history ■Development history••

Survey was undertaken on the assumed site environment and requirement specifications were formulated in connection with “R&D
for Treatment and Disposal of Solid Wastes (R&D for Advanced Processing and Analytical Methods)” in FY2018.
Basic performance of development method (Alpha camera) was verified on the basis of requirement specifications through element
tests.

Table 1 Overview of requirement specifications and test results

Cited from FY2017 supplementary budget “Subsidy Project of Decommissioning and Contaminated Water Management (R&D for
Treatment and Disposal of Solid Wastes (R&D for Advanced Processing and Analytical Methods))” FY2018 Research Report
Items

Requirement Specifications

Test items

Result overview•

About 50 mSv/h (Max.150 mSv/happrox.)

Sensitivity is proportional to the square of
the distance.
• 100 mm lens, there are cases where 4
Bq/cm2 can be measured by increasing
the measuring time.
• There is less impact when measuring
direction (spherical surface) is 0 to 90°.
Element test: Complex radiation • When measuring the contamination in
concave portion at a depth of 25 mm,
source evaluation
sensitivity worsens to 71% and is 67% at a
depth of 50 mm.
Confirmed gamma-ray sensitivity at 3.3 [s1/mSv/h]Element test: Environmental*Identified the location at angle of partial
dose impact test
incidence where shielding effect is less. It
needs to be improved.

Environmental temperature-5℃-35℃

Element test: Temperature test

Environmental humidity

Max.100%

Element test: Humidity test

Lighting available depending on
the case

Basic performance test: Opticalfilter evaluation

Sensitivity reduces by 78% due to opticalfilter.

Above outdoor dust levels

Element test: Powdery dust
impact test

Change in sensitivity is max. 1% outdoors.

α/βγ=1-106-8

Element test: Impact assessment Acquired sensitivity relative to alpha rays.
Beta rays (Co-60): 0.36%
test for beta-gamma nuclide
Gamma rays (Cs-137): 0.009%
concentration

Measurement distanceAccessibility
Profile of object to be measured

WeightEnvironmental dose

Environment
conditions
Lighting
Powdery dust
Beta-gamma nuclide
concentration

Max. 3 mm

Spherical surface, uneven, planesurface

Depends on the trolley

Basic performance test: Lens
evaluation

No effect on noise by using heat control
mechanism
Sensitivity decreases by less than 3% even at
95% max.
*Sensitivity tends to decrease when water isadded to radiation source.

* Settings were specified based on the characteristics identified from FY2017 supplementary budget “Subsidy Project of Decommissioning andContaminated Water Management (R&D for Treatment and Disposal of Solid Wastes (R&D for Advanced Processing and Analytical Methods))”©International Research Institute for Nuclear Decommissioning42

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Consolidation of FY2019 implementation details Issues identified from FY2018 implementation results
1. It was noticed that ghost-peak was formed with φ100 mm lens and detector combination.
⇒ Countermeasures for ghost-peak
2. It was noticed that the sensitivity is proportional to the square of the lens diameter and a φ
100 mm lens (it was φ 50 mm in the past) was used. It was confirmed that there weredirections where the shield was ineffective, partly due to the increase in size.
⇒Improvement of prototype
Issues to be addressed from FY2019
3. Tasks up to the tests using stabilized and standard radioactive source in FY2018 arecomplete. However, material and shape simulation assuming on-site conditions has not been
carried out.
⇒Continuation of element test

4.

Tasks up to imaging at a fixed position in alpha camera unit test in FY2018 arecomplete.
It is necessary to design a system that can move the measurement location remotely andevaluate the data.
⇒Addition of functions to software, design of measuring system©International Research Institute for Nuclear Decommissioning43

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Current status of alpha contaminated waste management and issues thereof ⚫

Current status of alpha contaminated waste management

When smear measurement detects alpha nuclide contamination at surface contamination density 0.4 Bq/cm 2 or above, thewaste is segregated and stored in a storage container. However, decommissioning work will commence on a full-scale in future
and a large amount of waste will be produced. Hence, it is not realistic to carry out smear measurement and segregate all
waste after dismantling, considering the time required for measurement and radiation exposure to workers.

⚫ Requirements pertaining to waste segregation technology➢➢

Identify the surface contamination distribution of the target objects before dismantling.
Reduce the measuring time and reduce the exposure dose associated with measurement.

⚫ Purpose of this ProjectTo examine its applicability to site and develop a technology that can measure the object (or area) to be dismantled in an
exhaustive manner and also within a short period of time.

Αlpha contaminated area

Alpha surface contamination
density distribution onequipment surface (Metals etc.)

Alpha surface
contamination density
distribution on building
concrete surface

Illustration of distribution measurement

Surface contamination density (Bq/cm2) distribution

Contamination distribution
measurement

Below lower detection limit

Fig. 1 Image showing distribution measurement using alpha camera©International Research Institute for Nuclear Decommissioning44

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Application targets Investigate the measurement needs and determine the timing of application and purpose of measurement.
“Understanding of contamination distribution” was established as an application target (uses) that can be used in the studyof work planning such as preparatory works, equipment dismantling etc. pertaining to debris retrieval in future.

[Assumed decommissioning flow (Ex: R/B)]
・Dry-up of stagnant water・Removal of spent nuclear
fuel・Debris retrieval

・Preparatory workfor the above

・Dismantling ofequipment(RI, RPV, etc.)

・Dismantlingof building

Fig. 1 Image showing application targets for alpha camera

Applicable targets for surface alpha contamination measurement➢Timing: Before the preparatory work for debris retrieval and prior to equipment dismantling work
➢Target: Surface of structures and equipment within the scope of the work
➢Use: Understanding of contamination distribution©International Research Institute for Nuclear Decommissioning45

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Application timing for alpha camera Investigate the latest needs (Purpose of contamination measurement) of TEPCO HD and determine the direction to aim forwith respect to this technology.

Purpose of contamination measurement:
・“Qualitative perspective on surface contamination density distribution” on the surface of theequipment and structures before dismantling
Utilization of the measured contamination distribution (Proposed):

・Development of work plan without increase in contamination and wastes

・Study on radiation control methods in accordance with contamination concentration and work content
・Study of the plan and the need for detailed measurement such as sampling etc.

Target object of
this project

Distribution
measurement byalpha camera

Primary storage
Study on detailed
measurement(Sampling)

Evaluation points
・Assessing the presence of contamination anddistribution of qualitative contamination density
・ Assessing the presence of localized contamination

Work plan

Work

Waste
generation

Storage andmanagement bydisposal classification

Study ondecontamination
and radiation
control methods

Fig. 1 Assumed work flow and application timing for alpha camera©International Research Institute for Nuclear Decommissioning46

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Overview of FY2019 development results Table 1 Implementation details and achievements in FY2019
Items

Implementation details

Verification test for
countermeasures forghost-peakCountermeasures for ghost-peak (peaks other thanradiation source location)• Cause for ghost-peak was identified and devices were improved.

Improvement of shielding structure① Review of shielding structure② Review of optical system in accordance withshielding structure• Optical system and shielding structure were determined by ray tracing analysis and shielding
calculation.
• Optical system: Evaluation was completed for angle of view and resolution performance of
this structure.
• Shielding structure: Performance equaling or surpassing the analysis results was verified
through irradiation test.

Verification of the performance of improved alphacamera• It was confirmed that there is no formation of ghost-peak.
• Basic performance was re-evaluated and specifications of alpha camera were consolidated.

Evaluation of the performance of measurement
using unsealed radioactive source etc.
① Simulation of adhesion to concrete, etc.
② Measurement of surface source, etc.

• Surface roughness of the object to be measured and its contamination profile were identified
as the causes affecting the measuring performance of alpha camera and were reflected in
test items.
• Evaluation of characteristics based on surface roughness and contamination profile of theobject to be measured was performed using unsealed Am-241 radiation source.

Addition of measurement results mapping function• Function was added for synthesizing the results acquired at multiple locations.
• It was confirmed that image synthesis and 3D space mapping was possible by scanning thealpha camera.

Study of on-site applicability• Applicability was set.

Study of specifications for measurement system• Pan/tilt/raising and lowering functions were identified and reflected into the device as
essential functions.

Study on mechanism specifications and control
methods• Behavior of essential functions (pan, tilt, raising and lowering) was verified using theexamined device.
• Combination with mapping function was also confirmed.

Basic verification of measurement system• Basic verification was performed on contamination distribution measurement and issues
concerning measurement system were identified.

Improvement ofprototypeContinuation ofelement test

Addition of functions
to software

Designing of
measuring system

Achievements©International Research Institute for Nuclear Decommissioning47

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Countermeasures for ghost-peak (1/2) -■Purpose of implementation
To implement cause investigation for ghost-peaks identified in FY2018 and determine the structure thatcan solve the problem. To perform a test to ensure that there is no formation of ghost-peak.

■Implementation details (Overview)
It was found that the lenses in optical system were not the only
cause but the internal structure of photomultiplier tube PMT
may also cause the formation of ghost-peak.

③ Emitted light returns
to light receiver forming
ghost-peak⇒Study the countermeasures against PMT structure.

Ghost-peak

Original peak

Ion feedback② Slight
emission of light(Cause of ghostpeak)

MirrorSlight emission of light

Lens

Incidence ofultraviolet rays
① Light converts
into electrons andamplifies.

FilterShield(Side face)

Measuringvisual field

Fig. 1 Internal structure of the device

Shield(Front face)Fig. 2 Sectional diagram of PMT©International Research Institute for Nuclear Decommissioning48

(b) Development of contamination evaluation technology for solid waste segregation- Countermeasures for ghost-peak (2/2) -【Confidential】■Implementation Results
・ Test was performed using PMT with gaps removed to prevent the return of light emission and it was compared with previous
results.
・ Decrease in ghost-peak was noticed at the end portion using a commercially available single lens (φ 50 mm).
Before countermeasures

After countermeasures

There is gap
and light
returns partiallyNo gapPicture showing PMT imaging area (Picture of PMT taken from lens side)

Ghost-peak

No ghost-peakDistribution when measuring Am241 (φ 15 mm thin disc)

Fig. 1 Comparison before and after the countermeasures for ghost-peak©International Research Institute for Nuclear Decommissioning

49

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Improvement of shielding structure (1/5) ■Purpose of implementation
To re-examine the design for shield and housing as it was noticed that shielding effect is weak in some directions in the
prototype. Also, to perform the test again for the prototype after taking countermeasures and confirm the effects.

■Implementation details (Overview)
Re-design and evaluation were carried out for the shieldable structure without obstructing the ultraviolet rays from alpharadiation.

Lens
Shield(Surroundingarea)

MirrorDirection
where gamma
ray shielding
effect is poor

Shield structure•Improvement in shielding effect is possible by increasing
the size of the shield (front face).
• However, sensitivity decreases since measuring visual
field is partly missing due to the shield (shadow isformed).

PMT
Shield (Front face)

FilterMeasuring visual field

Shadow is formedby the shieldFig. 1 Shield structure in FY2018©International Research Institute for Nuclear Decommissioning50

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Improvement of shielding structure (2/5) ■Implementation results

・Light tracing analysis was performed to evaluate the minimum size that light rays emitted from a light source within the angle of view, passthrough. (Fig. 1 and Fig. 2)・Study was conducted on a shielding structure that can be installed without obstructing the light rays. (Fig. 3)
・Analysis and evaluation were carried out to confirm the effects of the shielding structure. (Refer to improvement of shielding structure (3/5))
・Dependence on incidence angle was confirmed by irradiation test using Cs-137 radiation source. (Refer to improvement of shielding
structure (4/5))
Lens

Mirror② Extension of distance between
lens and mirror(66 mm ⇒ 99 mm)

Lens
Calculate thearea size

Light
rays

MirrorPMT③ Minimizationof opening size
(φ 108 mm)

FilterFig. 1 Example of light ray
analysis①Shield added

Gamma
rays
Shield(Side face)

Shield (Front face)

Measuring visual field

Both shielding and visual field can be balanced by extending the
distance between lens and mirror

Fig. 2 Light rays passage area at the lower
part of the mirror (Example)

Fig. 3 Image showing improvements and shield after improvement©International Research Institute for Nuclear Decommissioning51

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Improvement of shielding structure (3/5)<Results of analysis and evaluation>••

Gamma ray sensitivity was analyzed and evaluated assuming uniform contamination of Cs-137 in the surroundings ofalpha camera.
It was confirmed that gamma ray sensitivity can be reduced by 1/6.5 with a redesigned structure when compared to thestructure used in FY2018.

Table 1 Analysis conditions
Distance
between lensand mirror

Shield aroundmirror7

1

66.4 mm

No shield

FY2018 design model2

66.4 mm3

96.4 mm

The opening size is evaluated with the
distance between lens and mirror being
126.4 mm, so that ultraviolet rays cannot
pass through.

4

126.4 mm5

96.4 mm

Opening size
120 x 120 mm
Opening size
120 x 120 mm
Opening size
120 x 120 mm
Opening size
108 x 108 mm

Improved model. Small opening with the
distance between lens and mirror being
96.4 mm② Extension of distance between
lens and mirror (66 mm ⇒ 99 mm)Lens① Shield added
Mirror③ Minimization of
opening size (φ 108
mm)

6

5.3

5.1

5
4.0
4
3
2
1.0
1
After improvement

0

＃2

＃3

＃4

＃5

Analysis conditions

FilterMeasuringvisual field

6.5

＃1

Gamma
rays
Shield(Side face)

FY2018 structure

Remarks

Relative sensitivity to gamma rays＃

Shield(Front face)Fig. 1 Alpha camera structure and points for improvement

Fig. 2 Evaluation of gamma ray sensitivity using Cs-137
when the measuring environment is uniformly contaminated©International Research Institute for Nuclear Decommissioning52

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Improvement of shielding structure (4/5) <Test Results>•••

The dependence of gamma ray noise on the incidence angle was evaluated by controlling the incidence direction of gamma rays
through rotation of alpha camera using an improved shield against Cs-137 radiation source (position fixing).
Amount of noise reduced to max. 1/10 in FY2018 test under 45o and 135o conditions wherein the shielding effect was low.
It was confirmed that the average value too can be reduced to 1/13.7, which is the same as the analytical value (Fig. 2). It was
assumed that this was because the light produced by the lens was reduced by the filter placed between the lens and PMT.

180°

1000
*Gray color indicates shield

225°

Lens

100

Sensitivity in2018年度感度[s-1/mSv/h]
FY2018Sensitivity after
改良後感度[s-1/mSv/h]
improvement

135°

Angle with lessshielding effect(FY2018)

101
Mirror

270°

90°

0.1

band pass filter

315°
Cs-137 radiation source(Positioned at 0o condition.
Moved for each test)

Fig. 1 Test system

45°

Average count rate[s-1]FY2018：27.8
After improvement：2.00°

Fig. 2 Dependence of gamma ray noise on the incidence angle©International Research Institute for Nuclear Decommissioning53

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Improvement of shielding structure (5/5)<Improvements in the prototype including the shield>
Table 1 Improvements in the prototype
FY2018FY2019 improvement details

PMT

Opening at the end portion of imaging area

Filled the opening at the end portion of imaging area

Shield・Opening in the surroundings of mirror
・Distance between lens and mirror 66 mm

・Used a shield to cover the areas other than the entrance for ultraviolet raysfrom alpha radiation
・Distance between lens and mirror 96 mm

Cooling mechanism of PMT

PMT was given a cooling structure and ashield was installed outside it.

Reduced the size and improved cooling performance by changing it to acooling structure (radiating fin) that cools the PMT and releases heat from theentire shield in the surroundings.

Remote control function

Did not function.

Added function that allows remote power ON/OFF operation.

Circuit for alpha camera

MountedChanged the profile in order to match it with the profile of the housing andshield (function is maintained).

Optical shutter

Installed at the bottom side of the mirror

Located in between lens and mirror

FilterFixed to the front face of PMT

Improved the jig to integrate the filter with the lensEnvironment
monitor#2

Illuminometer

Shutter
ToF camera*1

CircuitLens (Filter
unit)

ShieldMeasuring direction

Fig. 1 Improved shield in FY2019

PMT (Built inside theshield)Coolingmechanism(Radiating fin)

#1 ToF camera: Used in 3D mapping function. A
camera with ToF (Time of Flight) function and which
captures a normal camera image and distance data
simultaneously.
#2 Environment monitor: Collects temperature of
the cooling unit, values of illuminometer and the
gamma ray intensity inside the shield©International Research Institute for Nuclear Decommissioning54

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Verification of improved alpha camera performance (1/2) ■Purpose of implementation
To re-evaluate the basic performance of an alpha camera with improved countermeasures for ghost-peak and shielding
structure.

■Implementation details (Overview)・ Software developed in FY2018 was modified in accordance with the changes in optical system and power ON/OFF
operation. (Fig. 1 – Fig. 3)
・ Alpha radiation measurement sensitivity, angle of view, position resolution etc. were evaluated.
Movement,
preparations
Commencementof measurement

BG
measurement

Illumination Shutter

ON

Closed-Display

DisplayON

ClosedNoisemeasurement

RecordRecordOFF

Open

Actualmeasurement

Displays imageof BG
measurement

Displays imageof BG
measurement

Actualmeasurement

ON

ClosedActualmeasurement noise
measurement

Displays data of
BG measurement

Displays data of
BG measurement

ON

Closed-Display

Display

Result output
Movement,
synthesis
process

Alpha radiation
Background image Distance data
data

Visual field of alpha camera
Am-241 radiation
sourceFig. 2 Alpha camera software screen (During BG measurement)

Am-241 radiation
sourceFig. 1 Alpha camera measurement flow diagram
Fig. 3 Alpha camera software screen (During actual measurement)©International Research Institute for Nuclear Decommissioning55

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Verification of improved alpha camera performance (2/ 2) ■Implementation results

・Issues identified in FY2018 were resolved and evaluation of the basic performance of alpha camera was completed (Table 1)
Table 1 Overview of test methods and results

No.

Items

Test Methods

Results

Remarks

1 Sensitivity to alpha rays

Install 3 types of standard sources (10 cm square, thin plate) withdifferent intensities at a distance of 1 m and carry out measurement.

0.23 s-1/ (Bq/ cm2)

2 Measurement angular field

Install Am-241 source of Φ36 mm 1 m ahead and evaluate thesensitivity at each position within the visual field.

Confirmed that ideal measurement angular field: 27.2° Refer to Fig. 1.
x 27.2°
Area where sensitivity at 80% or above can be achieved is
Effective measurement angular field: φ25.2°
defined as an effective angular field.

3 Position resolution

Evaluate position resolution from measured distribution by simulating±21 mmAm-241 point source at the aperture.

4 Distance property

・ Confirmed that it is inversely proportional to the
Carry out measurement by changing the installed position of Am-241
square of distance.
Refer to Table 2.
source between 0.5 m – 3 m.
・Confirmed that it can be corrected with distance data.

5 Sensitivity to beta rays

Measure Sr/ Y-90 source (9kBq) and evaluate sensitivity to beta
rays.

0.05% of alpha ray sensitivity or less

Effect is reduced by an order of magnitude as compared to
FY2018 results based on the review of optical system.

Evaluate incidence angle dependence with irradiation facility.

2.0 s-1/ (mSv/ h)

Improved as compared to 3.3 s-1/ (mSv/ h) of the FY2018
results.
Refer to improvements in shielding structure.

6

Sensitivity to gamma rayssourcesourceIdeal measurement
angular field(Red dashed line)

Sensitivity is reduced by about 20% as compared to last
fiscal year based on the review of optical filter.

Major factor is the expansion of alpha ray-emitting region
and dependence on measurement distance is less.

Table 2 Distance property and correction results
Corrected counting
rate
[s-1]

0.5

Counting
rate
[S-1]
64.7

1

17.8

17.8

1.5

7.3

16.4

2

4.2

15.8

2.5

3.0

18.5

3

2.0

17.4

Distance[m]sourcesourceEffective measurement
angular field
4 corners are far from thecenter of the measurement
angular field, hence
sensitivity is reduced.
(Orange dotted line)-

Fig. 1 Measurement results for each source position©International Research Institute for Nuclear Decommissioning56

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Evaluation of measurement performance using unsealed source (1/ 5) ■Purpose of implementation
Test was performed in FY2018 using the metallic sealed sources, thinned flat plate and thinned disc.
However, it is assumed that irregular contamination occurs on non-metals such as concrete etc. on
the site. Hence, the purpose is to conduct a test assuming contamination closer to the site and clarify
the locations for application.

■Implementation details (Overview)

Multiple types of Am-241 #1 unsealed sources were developed and combined with sealed sources to
conduct the following tests. Further, in each test, the difference in measurement results wasquantitatively evaluated by comparing with the existing detector (ZnS survey meter etc.).
#1： Nuclide specified with JIS4329:2004 radioactive surface contamination survey meter

Table 1 Test items#

Test items

Test methods

Parameters

1

Performance evaluation for measurementof structures assuming the floor and walls
at the site

Measurement of unsealed source byapplying alpha radioactivity to concrete

Concrete surface painted ornot2

Evaluation of measurement performancein case of complex contamination profiles

Measurement by combining unsealed/sealed sources and changing thesurface contamination profile

Source profile©International Research Institute for Nuclear Decommissioning57

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Evaluation of measurement performance using unsealed source (2/ 5) ■Implementation results• Factors behind the impact on lower detection limit with respect to surface contamination density wereconsolidated. (Fig. 1)• Source profile and source surface profile were identified from these factors.
• Sources with different profiles and concentrations were created and measurement sensitivity wasevaluated.
Lower
detection
limit

Sensitivity

Amount of alpharays emitted

Attenuation oflight

Alpha ray energy

Proportional to energy in principle

Air composition

Assumed to be constant

Humidity

Implemented in FY2018

Powdery
dust
Distance

Noise

Temperature
Emission bybeta rays(gamma rays)

Air dose rate

Contamination profile (Local
contamination, distributed
contamination)
Contamination characteristics(Water contamination, surface condition,
contamination levels)

Implemented in FY2018
Lighting

Fig. 1 Analysis tree for the factors influencing the lower detection limit©International Research Institute for Nuclear Decommissioning58

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Evaluation of measurement performance using unsealed source (3/ 5) <Method for creating unsealed contaminants and test items>
Table 1 Test items and method of creating specimen
Evaluation items
Contamination profile

Contamination
characteristic

Method of creating specimen

Test method

Local
contamination

Put drops of Am-241 on concrete.
Use quantity of liquid dropped and concentration (Bq/ cc) as parameters.

Distributed
contamination

Change the profile by tiling solid concrete with various contamination levels.
(Fig. 1, Fig. 2)

Watercontamination

Put drops of Am-241 on solid concrete and create aqueous Am-241 source.
Change the moisture content with or without drying.

Surfacecondition

Put drops of Am-241 on concrete (blasting process, solid, epoxy painting)and on the top of flat metal plate. (Fig. 3)

Contaminationlevel

Use solid concrete with varying contamination levels.

Evaluation of sensitivity to alpha rays in
different conditions

Comparison of sensitivity to alpha rays
with surface roughness of each material

Measurement of sensitivity to alpha
rays

Solid

Epoxy paint

Blasting process ofapprox. 5 mm

Blasting process ofapprox. 1 mm

Concrete specimen

Scope of blasting process

Application
range

Drops of Am-241

Fig. 1 Concrete source (solid) for tilingFig. 2 Condition of tiling for
distribution measurement

Fig. 3 Appearance of each concrete©International Research Institute for Nuclear Decommissioning

59

(b) Development of contamination evaluation technology for solid waste segregation- Evaluation of measurement performance using unsealed source (4/ 5) -【Confidential】<Test Results Summary>Evaluation was performed on the effects of moisture content and surface roughness of the measured object on the sensitivity.
In future, a study will be conducted on the evaluation of alpha contamination distribution while checking the surface condition of
the measured object through the camera image.
Table 1 Test Results (Summary)
Test Name

Test Method

Results

Remarks

Contaminationlevel

Apply radioactivity of 40 to 4000 Bq (0.4 to 40Bq/ cm 2) to solid concrete and carryout measurement from a location 1 m away.

Confirmed the counting rate in correspondencewith concentration.

Refer to .Fig. 1

Local
contamination

Control the amount of Am-241 source dropped onto the specimen, create sourcewith a size ranging from φ 6 mm to φ 12 mm and carry out measurement from alocation 1 m away.

Confirmed that sensitivity tends to increase when
the size of the source is small.

It is assumed that emission
spreads over multiple pixels if thesize of the source is large.

Distributed
contamination

Lay solid concrete tiling of 10 cm x 10 cm and carry out measurement from alocation 1 m away.
Compare tile laying and counting rate.

Confirmed that distribution can be measured
approximately.

Refer to the next page.

Water
contamination

Take 20 ul of solvent containing Am-241 at approx. 400Bq, put drops of thissolvent onto solid concrete, and perform measurement before and after drying thesolvent. Compare the sensitivity.

Confirmed that sensitivity will reduce to 44%
when the cement contains moisture.

Refer to Fig. 2.

Surfacecondition

Apply Am-241 at 4 Bq/ cm2 in the form of mesh on solid concrete, painted
concrete and blasted concrete (2 types) with different surface roughness andcarry out measurement from a location 1 m away.

・The higher the surface roughness, the more thesensitivity.
・Blasted concrete was found to be 6 times moresensitive than solid concrete.

・It is assumed this it is becauseof the increase in surface volume
and increase in the rate ofrelease into air. (Refer to Fig. 3)

5
y = 0.0021x + 0.8048
R² = 0.9956
0
0

2000

4000

6000

Dripped amount (Bq)Fig. 1 Linearity at solid concrete source

8000

Dry state

Water-containing stateFig. 2 Comparison of presence of water in contaminated area

0.35
0.3

1 mm blastingprocess

0.25

5 mm blastingprocess0.2[s-1]

Counting rate with respect to 400 Bq

10

Relative ratio

Alpha camera counting rate [s-1]

15

0.15

Solid concrete/painted concrete

0.1
0.05
0
0

0.5

1

1.5

Deviation of specimen surface depth [mm]Fig. 3 Correlation of counting rate with surface roughness©International Research Institute for Nuclear Decommissioning60

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Evaluation of measurement performance using unsealed source (5/ 5) <Test Results (Effects of source distribution)>••

It was confirmed that distribution can be measured roughly in accordance with source radioactivity distribution.
It was confirmed that the counting rate tends to increase depending on the expansion of light emitting areasurrounding the pixels with high radioactivity.

Placement of source (Numbers specified in the squares represent radioactivity [Bq])Equivalent to 1 pixelof alpha camera

A Section

0.4

0.4

0.3

0.3

0.2

0.2

0.1

0.1

0

0
1

2

3

4

5

Pixel No.

6

7

8

3

6

2

4

1

2

0

0
1

2

3

4
5
6
Pixel No.

7

8

8
Experimental
実験値
value
Theoretical
理論値
value

4
3

6
4

2

2

1
0

0
1

2

3

4 5 6
Pixel No.

78

Source intensity relative value

0.5

5

8
Experimental
実験値
value
Theoretical
理論値
value

Experimental value [s-1]

0.5

0.6

4

Source intensity relative value

0.6

0.7

Experiment
実験値
al
value
Theoretical
理論値
value

Experimental value [s-1]Experimental value [s-1]0.7

Source intensity relative value

Blurring due toexpansion of emission

Fig. 1 Example of source placement and measured results (A section)©International Research Institute for Nuclear Decommissioning

61

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Addition of the function for mapping measurement results (1/ 2) ■Purpose of implementation
Lower detection limit in alpha camera depends on measuring distance and environment. Hence, the purpose is to study the method
for correcting the measurement results so as to enable evaluation with same standards using measuring distance and environmental
conditions.
To develop a software by integrating the measurement results at multiple locations (creating a panorama) based on the software thatsynthesizes contour diagram with camera image created in FY2018.

■Implementation Details (Overview)
1. Development of function for correction of measuring distance and environmental conditions
2. Development of software by integrating the results of measurement at multiple locations
1. Development of function for correction of measuring distanceand environmental conditions

・ It was confirmed in a FY2018 test that sensitivity [S-1/ Bq/ cm2] toalpha rays is inversely proportional to the square of the distance.
・ Conduct a study on the addition of function for sensitivity estimationand study the methods for data correction while including the deviation
in horizontal position within the angle of view.

Distance condition
Estimation of sensitivity at
different target locations
through analysis and test
(Inversely proportional to thesquare of the distance)

Environmental
condition
Measurement of theamount of noise throughBG measurement(Implemented in
FY2018)

Data correction

Fig. 1 Data correction flow

2. Development of software by integrating the results ofmeasurement at multiple locations

Develop viewer software that allows a bird’s eye view of thesynthesized data and the algorithm for synthesizing the sensor
information (image, alpha dose, distance) acquired at multiple
locations.

Synthesis of data
at multiple
locations

Take pictures by moving
the camera to fit theobject since the angle ofview is limited.

Image and alpha rays data

Distance

Viewer allows a bird’s
eye view confirmation of
the results beyond
angular field.

Data at one point

Fig. 2 Illustration of software with overlay of images©International Research Institute for Nuclear Decommissioning

62

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Addition of the function for mapping measurement results (2/ 2) ■Implementation results• Operation check was carried out by conducting study and test on the process flow pertaining to photographic datasynthesis process at different locations resulting from pan, tilt and raising/ lowering of trolley. (Fig. 1)• Anti-distortion measures assumed from the characteristics of 3D distance data (acquired by TOF camera), noise and color
correction methods were implemented (Fig. 2) and test evaluation was performed.
• 3D data for the photographs of wall, floor and structures taken in different directions can be expressed with an image
quality that is close to actual vision.

1st

Alpha camera data

Image

Distance data

Noise correction/
point grouping
process
Estimation of positional relationshipbetween adjacent point groups using
image

Pan/ tilt/ raising/lowering information

2nd

Estimation of
adjacency
relationship3rd

It will be a patchy patterndepending on the lighting
conditions, etc.

Colors in thesame positions

Difference incolorsDifference incolorsPhotographic data includedin the area
Determine the average ofcolor and alpha camera
measurement results from themeasurement results at
multiple locations.

High precision estimation of positional
relationship using point groups

Synthesis of point groups using
positional relationship

Smoothing process of point groups

3D data after processing

Fig. 1 Processing flow

Fig. 2 Sample image of smoothing process©International Research Institute for Nuclear Decommissioning63

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on the site applicability (1/ 2) ■Purpose of implementation••

To conduct characterization with respect to environmental factors based on the element test results obtained until
FY2018.
To conduct study on alpha camera specifications (size, weight, measuring distance) and options for measurement
objects. To clearly define the site applicability to design a system most suitable to these conditions.

■Implementation details (Overview)
1. Study on the location used (operation floor or 1st and 3rd floor in R/ B)
2. Study on objects to be measured
Study on the site
applicability

System design
Alpha camera specifications(Shield, optical system etc.)

Location used
Environmental conditions,size, level difference etc.

Specifications for pan/ tilt mechanism,
raising/ lowering mechanism

Determination of specificmeasurement procedure

Objects to be measured
Size, measuring distance

Software specifications

Fig. 1 System design flow
©International Research Institute for Nuclear Decommissioning64

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on the site applicability (2/ 2) -■Implementation results•••

Needs investigation was conducted in connection with the measurement of surface alphacontamination density. Need was identified by qualitatively assessing contamination distribution forestablishing work plans such as dismantling etc. rather than segregation of wastes.
The 3rd floor of Unit 1 R/ B, 1st floor of Units 2 and 3 R/ B and refueling floor of Unit 2 R/ B wereassumed as the usage sites with high likelihood of work related to debris retrieval.
Study was conducted on the configuration of the device to be used for distribution measurement andthis information was incorporated into device specifications.

Table 1 Applicability of alpha camera
Items

Applicable conditions

Remarks

Lighting condition at measuringlocation

Dark room(No ultraviolet rays in particular)

・Not applicable in the presence of light that contains
ultraviolet rays such as those of the sun, moon etc.
・Dark room at 0.1 lux if there are no ultraviolet rays

Dose rate in measuring environment

Max. 50 mSv/ h(Depends on low detection limit, evaluation resultsof last fiscal year)

Reflect the performance of the improved shield.

Surface βγ contamination density
within measurement range

Max. 100 MBq/ cm2(Depends on low detection limit, evaluation resultsof last fiscal year)

Reflect the performance of the improved alpha camera.

Material of object to be measured

Concrete (solid, painted), metal. However,sensitivity reduces if object to be measured
contains moisture.

Reflect the performance of the improved alpha camera.

©International Research Institute for Nuclear Decommissioning65

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on the specifications of measuring system (1/ 2) -■Purpose of implementation
To identify the functions required for the measuring system in conformance with site applicability. To determine thespecific measurement procedure for the site based on the site applicability and decide the system specifications.

■Implementation details (Overview)
・Study of the measurement methods in consideration of the information concerning angular field of alpha camera and changes in
the characteristics depending on the measuring distance・Evaluation of the necessary operating performance based on the studied measurement method

Determine target
object to bemeasuredWall, large panel, etc.
(Objects considered tobe perpendicular)
The edge of the wallmoves to a position at theedge of the angular field at
a distance of Xm.

Move to a position Ym
away from the object to bemeasured.

Tip movement

Align with the height
direction.

Direct it obliquely
downwards (Upwards).

Movement

Move to a position that
does not overlap with theprevious angular field.

Move to a position that
does not overlap with theprevious angular field.

Trolley movement

Perform
measurementfrom face side as
far as possible.

Floor, ceiling (Objects
considered to be
horizontal)

Measurement fromoblique direction.
Distance depends
on pixels.

Specification of the target objects
・Size・Profile
・Material etc.

Small objects that can beaccommodated into apart of the angular fieldsuch as pipes, etc.
Move to a position where
the edge of the object tobe measured matches withthe pixels
If the object is verticallylong, align the height
direction.
Move to a position that
does not overlap with theprevious angular field.
Move it so that
“Pixel” and theboundary of theobject are aligned.

Specification of moving range
・Moving distance・Types of movement(Horizontal, perpendicular,
rotation)

Alpha camera(Trolley is notshown.)

Fig. 1 Measurement procedures studied in FY2018
Excerpts from FY2017 supplementary budget “Subsidy Project of Decommissioning and Contaminated Water Management (R&D for Treatment
and Disposal of Solid Wastes (R&D for Advanced Processing and Analytical Methods))” IRID design review (G1#2) “Investigation anddevelopment of alpha contamination measurement methods”©International Research Institute for Nuclear Decommissioning66

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on the specifications of measuring system (2/ 2) ■Implementation results・Study on the measurement method so that the distance with the target object to be measured becomes
constant. (Fig. 1)
・The following functions were deemed mandatory or desirable in view of the measurement of floor surface,walls and ceiling in order to evaluate surface contamination distribution, which is the purpose of themeasurement. It was decided to go for specifications incorporating both the functions.
1)Mandatory function: Tilt (Vertical oscillation) and raising/ lowering function
2)Desirable function: Pan (Oscillation in horizontal direction) function
For floor surface measurement

For wall surface measurement

For ceiling measurement

Measurement by raising/ changing camera
direction so that the floor surface can bemeasured in the perpendicular direction as
far as possible.

Measurement by changing the height so thatthe wall surface can be measured in theperpendicular direction.
#Direction of alpha camera is controlled bytrolley or pan.

Maintain the direction of alpha camera
upwards and measure the ceiling. Raise/
lower the camera as required.

床面

Alpha camera

Alpha camera
Alpha camera

Raising/
loweringmechanism
Trolley

Carry outmeasurement bychanging the angle intilt mechanism andmoving the trolley inhorizontal direction.

Carry out measurement bymoving the trolley inhorizontal direction after
adjusting the height withthe raising/ lowering
mechanism.

Carry outmeasurement bychanging the heightwith the raising/lowering mechanism.

Trolley

Trolley

Fig. 1 Measurement image of alpha camera using raising/ lowering, pan/ tilt functions©International Research Institute for Nuclear Decommissioning67

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on mechanism specifications and control method (1/3) -■Purpose of implementation
It is necessary to control the measuring location and direction of alpha camera through remote operation since highradiation dose environment is expected at the site. Hence the purpose is to determine the specifications and control
method for the mechanisms such as pan/ tilt, raising/ lowering mechanisms of alpha camera (approx. 150 kg or above)
including the shield.

■Implementation details (Overview)

・Study the driving mechanism of the alpha camera to enable remote operation of pan/ tilt and raising/ lowering
mechanism.
・Study the system for sharing the data on measuring conditions (pan, tilt, height) required for mapping the measurement
results.
・Study the fall prevention measures for alpha camera in view of emergencies such as earthquakes etc.
チルト
Tilt

120°

Alpha
cameraアルファカメラ
運搬台車装置の制御装置
Control
device fortransport trolley device
Power supply unit for
アルファカメラのalphacamera
電源ユニットPan tiltパンチルト機構
mechanism
Shooting direction
⇐撮影方向

865

Wire

アルファカメラ
cameraAlpha
最小高さminimum height

Cable length(30m or above)
ケーブル長（30m以上）
Wire type
ワイヤ式ポテンショメータ
potentiometer（昇降位置検知）(Raising
lowering
position detection)
Pushhandle
手押しハンドル

Outrigger
アウトリガーPCfor alpha camera
アルファカメラ操作用PC
operation

PC 画像合成処理用PC
for image synthesis
process

Ethernet

ワイヤ

1170

±180°
パン
Pan

2035

cameraAlpha
アルファカメラ
maximum
最大高さ
height昇降andRaising
lowering

-45°

HUB

PC
for transport trolley
運搬台車装置操作用PC
operationPower
電源 supply
AC100V

Trolley consisting of
電動リフター付き台車
electric lifterFig. 1 Configuration of measurement system©International Research Institute for Nuclear Decommissioning68

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on mechanism specifications and control method (2/3) ■Implementation results<Trolley provided with pan/ tilt and raising/ lowering mechanisms>
・When measuring in the lower direction, the alpha camera was arranged in such a range, where the visual field of alphacamera is not obstructed by the trolley and also the alpha camera is arranged in such a position that it does not fall/ tilt
while operating pan/ tilt and raising/ lowering mechanisms.
・A system that allows remote control of pan/ tilt and raising/ lowering operations was finalized. (Refer to Table 1.)
Table 1 Specifications of investigation system
Classification

Item

Pan tilt range

Accessibility

アルファカメラ
Alpha camera

運搬台車装置の制御装置
Control unit of transport
trolley equipment

Axial direction
measurement method

Requirement specifications

Remarks

Pan: ±180[°]
Tilt: -45-230[°]
Pan: Front face 0[°](Operable upper tilt limit is till 90[°] when alpha camera Tilt: Horizontal 0[°]is mounted. Refer to test results.)

Raising/ lowering stroke

1170 [mm]Alpha camera height

Raising/ lowering at the lower end:
From floor surface 865 [mm]Raising/ lowering at the top end:
From floor surface 2035 [mm]Pan tilt angle

Motor-mounted resolver

Raising/ lowering position

Wire type potentiometer

Pan tilt axis

Stepping motor with brakes

Camera optical axis height

Driving system

Tilt
チルト機構
mechanismアルファカメラのPower
supply unit for
電源ユニットalphacameraControl system

Mobility

Pan パン機構
mechanism

Remote operation(Tilt drive motor)(Pan drive motor) （チルト駆動モータ）（パン駆動モータ）(Wire type potentiometer)
ワイヤ式ポテンショメータ

Raising and lowering axis

Motor pump driven hydraulic cylinder

Motor-operated lifter mounted

Pan tilt axis
Raising and lowering axis

Target angle control: JOG operation (operation whilepressing operation button)
Target position control: JOG operation (operation whilepressing operation button)

Acceleration/ deceleration control
Constant speed operation by ON/
OFF control

Movement method
Fall prevention

Movement by hand pushing
Installation of outrigger when alpha camera is raised to a Safety measures during earthquake
etc.
higher position by raising/ lowering operation

With/ without cableWith cable (AC100V cable and Ethernet)

Operating area

Communication system

WeightFig. 1 Appearance of pan/ tilt mechanism

Ethernet can be wireless in thecurrent system too.

Operation is allowed from a remote location of 30 [m] orabove using the PC available for operation.
Ethernet
Pan tilt mechanism: 42 [kg]Control device: 16 [kg]Lifter trolley: 183 [kg](Weight of each single unit)

Raising/ lowering at the lower end:

Weight/ contour
Contour

Raising/ lowering at the top end:

Width (W) without outrigger is 759[mm](Alpha camera, outrigger included)©International Research Institute for Nuclear Decommissioning69

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Study on mechanism specifications and control method (3/3) -<Image synthesis test>

・The pan/ tilt and raising/ lowering operations can be carried out with the alpha camera with shield mounted.
・It is confirmed that the measurement results (background image, distance data) for 45 conditions can be synthesized to 3
dimensional data.

Table 1 Synthesis data measuring conditions

Parameters

Raising/ lowering
amount [mm]Pan [°]

Tilt [°]

Acquired image
count

78 mm

0,20,40,60,80 (5conditions)-20,0,20 (3 conditions)15

493 mm

0,20,40,60,80 (5conditions)-20,0,20 (3 conditions)15

883 mm

0,20,40,60,80 (5conditions)-40,-20,0 (3 conditions)15

Improved alpha camera(Shield included)
TOF camera++ Synthesis
process…
Pan/ tilt, raising/lowering mechanism

Fig. 1 Picture of the equipment

Blue portion:
Range inwhich
distance datacannot beacquired

Red portion:
Alpha camera
measurementarea

45 conditions (15 x 3) image

Fig. 2 Example of 3D synthesis of the photographic data of multiple locations©International Research Institute for Nuclear Decommissioning70

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Fundamental verification of measurement system -■Purpose of implementation

・To measure alpha contamination distribution and identify the issues in measuring system as a part of fundamental verificationof the examined measurement system.

■Implementation details

・ Measurement was carried out by changing measuring direction (pan angle) with the standard alpha source arranged at multiple
points in the dark room to evaluate alpha contamination distribution.
(Table 1, Fig. 1)■ Implementation results• It was confirmed that distribution can be viewed roughly.
• However, alpha contamination distribution is not displayed in case of source ① which lacked distance data. (Issue)• Measurement could not be carried out for source ④220 Bq because of the distance and low surface density. (2 Bq/ cm2).

Table 1 Am-241 source used
Radioactivity

ProfileSource ①

2.96 kBq

100 x 100mm, flat plate

Source ②

9.4 kBq

Φ36mm, circular disc

Source ③744 Bq

100 x 100mm, flat plate

Source ④

220 Bq

100 x 100mm, flat plate

Dark
room
entrance

Source ②
Dark room

Alpha contamination
distribution is not
displayed due toabsence of distancedata.
Alpha camera

White board

White board

Fig. 1 Source arrangement inside dark room

Source ③
Fig. 2 Three-dimensional synthesis results©International Research Institute for Nuclear Decommissioning71

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Summary (1/2) -<Ghost-peak countermeasures verification test>• PMT was identified as the cause for ghost-peak formation and PMT was improved. As a result, it was
confirmed that there is no ghost-peak formation.
<Improvement of prototype device>• Shielding mechanism was designed by which shielding performance improved without blocking ultraviolet raysfrom alpha rays.
• It was confirmed from improved alpha camera test that alpha contamination distribution and necessary
shielding performance can be achieved.
<Continuation of element test>• The test was performed using unsealed source and the effects of source size, surface roughness of object tobe measured and evaluation of moisture amount were confirmed.
• It was confirmed that sensitivity changes by 0.4 to 6 times when the surface is rough, or if there is watercontent.
• It is necessary to study the evaluation method for alpha contamination distribution with combined use ofcamera image (Issue)<Functional addition to software>・ The function for mapping between 3D spaces by synthesizing the measurement results of multiple locations
was developed.
・ However, issues were identified in some objects to be measured, such as difficulty in acquisition of distancedata in a few target objects to be measured, missing display and 3 dimensional data creation taking longer
time (approx. 96 minutes are required for the synthesis of 15 images) etc. (Issue)<Designing of measurement system>
・ Designing of the system that can remotely control pan, tilt, raising/ lowering functions required for on-sitemeasurements with alpha camera mounted, was completed.

©International Research Institute for Nuclear Decommissioning72

【Confidential】(b) Development of contamination evaluation technology for solid waste segregation- Summary (2/ 2) <Proposal for FY2020 Implementation Plan>① Study on the methods for determining specific measurement procedure and measurement
results
・Study the methods for the evaluation of measurement methods and measurement results taking into
account the effects of surface roughness of the material and issues involved in mapping to 3
dimensional model, which have been identified in this fiscal year.
② Development of investigation system for performance evaluation test and test preparations
・Carry out necessary preparations for development of required devices and required tests for
performance evaluation.
③ Performance evaluation test・Confirm measurement performance for alpha contamination distribution by applying development
techniques in a beta and gamma nuclide contaminated environment.
④ Evaluation of development method for on-site applicability
・ Evaluate the applicability of development methods based on the performance evaluation test.

©International Research Institute for Nuclear Decommissioning73

【Confidential】

2. Project details
b. Establishment of the concept of treatment and disposal anddevelopment of safety assessment techniques(a) Establishment of techniques for selecting the preceding processing method① Low-temperature treatment technologies② Study on the approach for evaluating the applicability of treatment
technologies(b) Presentation of disposal methods, and development of safety assessment
techniques① Study on disposal methods in accordance with classification of solid
wastes and collection & consolidation of information for establishingsafety assessment techniques② Development of techniques for assessing impact of affecting
substances, etc. on disposal©International Research Institute for Nuclear Decommissioning74

【Confidential】(a) Establishment of techniques for selecting the preceding processing method◼ Accomplishments till FY2018
 All the pros and cons of evaluation axes were weighed in order to assess both waste characteristics and treatment technology together
and techniques (approaches) that allow diversified evaluation of technology were proposed.
 Out of the evaluation axes that were considered in connection with low-temperature treatment technology (Cement solidification, AAM
solidification), with respect to the data on characteristics of solidifications, which was insufficient, data on solidifications that were not mixed
with waste was acquired.
◼ Goals
 To acquire and evaluate data on high-temperature treatment technologies and low-temperature treatment technologies, required for
identification of stabilization and solidification technology expected to be applicable to actual treatment, so as to contribute towards
establishment of techniques for selecting the preceding processing method.
◼

Implementation items and overview
Implementation items① Lowtemperature
treatment
technologies② Study on theapproach forevaluating theapplicability oftreatment
technologies

Overview

i. Acquisitionand evaluationof datapertaining tolowtemperature
treatment
technologies,
which would
contributetowards
identification oftechnology(i) Study on verification method of low-temperature treatment
solidification

Evaluation of applicability of test methodsto be used for determining the possibility ofsolidification and for setting the conditions(ii) Acquisition of data on characteristics of cement and AAM
solidifications in the slurry

Preparation of simulated solidifications, measurement of each type of physical property andacquisition of data on characteristics(iii) Investigation of special cements

Investigation of useful special cements and acquisition and testing of data concerning Na2CO3
limit value, etc.

ii. Investigation
related todegradationofsolidifications(i) Investigation of degradation of solidifications due to heating, etc.

Acquisition of data based on the study and tests concerning degradation tendency due toheating, drying, etc.

(ii) Evaluation of the correlation between inventories such as Cs, etc.
and the temperature of solidifications

Clarification of the changes in heat conditions of solidifications depending on the difference inmaterials and different shapes of the containers(iii) Investigation and evaluation of factors having an impact on longterm degradation

Study and estimation of the application scope for calculation of thermodynamic equilibrium

i. Research and investigation concerning the impact of waste composition on theperformance of solidifications

Study on the impact of solid waste composition on the performance of solidifications

ii. Acquisition of data pertaining to facility configuration, etc. for each treatment
technology

Acquisition of data on economic efficiency and addition/updating of the technological survey
sheetiii. Investigation of Cs volatilization volume and its control during high-temperature
treatment

Investigation of Cs volatilization and control mechanism©International Research Institute for Nuclear Decommissioning75

【Confidential】(a) Establishment of techniques for selecting the preceding processing method◼ Indicators to determine the implementation details and goal achievement
Implementation items① Lowtemperatur
e treatmenti.
Dataacquisition

Presentation of proposed screening techniques to check the possibility of solidification

Identification of the scope within which various types of lowtemperature solidification are applicable

Verification of techniques and identification of scope of applicability

Acquisition of data on solidifications mixed in carbonate slurry

Management of data such as performance, etc. of solidifications when mixed withsimulated carbonate slurry

Acquisition of data on solidifications mixed in iron co-precipitation
slurryManagement of data such as performance, etc. of solidifications when mixed withsimulated iron co-precipitation slurry
Presentation of scope of applicability concerning solidification properties

Investigation on available special cements

Identification of options wherein the scope is likely to broaden and understanding of theproperties of base metal

Study on the scope within which the various types of lowtemperature solidification are applicable

Identification of the scope within which the various types of low-temperaturesolidification are applicable(i) Degradationofsolidifications
due to heating,etc.

Investigation into changes in the properties due to heating/drying
and study on the tendency of change in base metal properties

Presentation of the tendency of change in base metal performance

Acquisition of data from near the inflection point at which change
in properties occurs

Presentation of the scope of applicability based on the data from near the inflection
point at which change in performance occurs(ii)
Temperature ofsolidifications

Establishment of an analysis system and analysis concerning
cement solidifications

Presentation of the correlation between cement solidifications inventory and heat

Implementation of thermal analysis inclusive of AAM

Presentation of the correlation between cement solidifications inventory and heat, andestimated achieved temperature

Trial run of the collection of thermodynamic data and calculationof thermodynamic equilibrium

Presentation of estimated results pertaining to changes in (Cement) mineral phase
Presentation of the sufficiency of (AAM) thermodynamic data and applicability ofequilibrium calculation

Evaluation of the applicability of calculation for thermodynamic
equilibrium

Presentation of the applicability of thermodynamic equilibrium calculations for phasechange, and the results

Evaluation based on the collection of vitrification data and glass
property models

Collection of data on glass properties and presentation of evaluation results pertaining
to the filling density of secondary waste generated from contaminated water treatment
based on Japan’s National Glass Database

Evaluation based on the US Glass Database and evaluation of
the scope within which mixing is possible with respect to lowtemperature treatment

Presentation of the scope within which the main secondary waste generated fromcontaminated water treatment can mix with glass
Presentation of the scope of low-temperature treatment based on the data in ①i and ii

Data acquisition pertaining to facility configuration, etc. for eachtreatment technology

Presentation of data such as facility configuration, treatment parameters, etc.

Consolidation of investigation results for each treatment
technology

Presentation of consolidated information on each technology

Consolidation of knowledge concerning Cs volatilization andcontrolPresentation of consolidated knowledge concerning Cs volatilization and control

Measurement of the effect of controlling Cs volatilization duringhigh-temperature treatment

Presentation of the effect of CS volatilization controlling techniques based on the tests(iii) Special
cements(iii) Long-term
degradation②
Approach

Indicators to determine goal achievement

Study on verification method of low-temperature treatment
solidification(i) Inspection
technique(ii) Data on
properties ofsolidificationsii.
Degradatio
n ofsolidificati
ons

Implementation details

i. Impact on the performanceof solidifications

ii. System configuration, etc.

iii. Cs volatilization andrestraints©International Research Institute for Nuclear Decommissioning76

【Confidential】(a) Establishment of techniques for selecting the preceding processing method- Implementation details and reflection destination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

Technologicalinformation

セメント固化

評価軸Ａ

評価軸Ａ

Analysis of waste properties(Details)

Ａ

Ｂ

Ｃ Ｄ・・・・・・・・
溶融技術

Applicable candidate
technologies

Analysis items(Evaluation axis)

General

汎用性
versatility
Economic
・・・・・
経済性
efficiency

社会的Socialacceptance
受容性
・・・・・

Safety安全性・・・・・・
政策とのCorrelation
with the plan
関係性

Compatibility
互換性Ease of運転容
operation
易性
・・・・・

メンテナ
Maintainability
ンス性

Selection of treatment
technology

Identification of applicable candidate technologies[Treatment technology selection process] Contribution towards introduction into actual treatment equipment{Data on properties, etc.}⚫ i(ii) Acquisition of data on properties
of solidifications⚫ i(iii) Investigation of special cements[Impact of heat and long-term soundnessdata]⚫ ii Investigation related to degradationof solidifications(Temperature
conditions)[Process information, etc.]⚫ ② Study on the approach for evaluating
applicability[Impact of heat and long-term soundness data]⚫ ii(ii) Evaluation of correlation withtemperature of solidifications

Analysis of waste
properties(Simplified / Alternative)⚫ i(i) Study on verification
method of solidification

Test
methodVerification of possibility ofsolidification

Operation management
parameters

Analysis method

Setting actual treatment conditions][Treatment technology selection process] Contribution towards operation management after introduction into treatment equipment©International Research Institute for Nuclear Decommissioning77

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】① Low-temperature treatment technology
i. Acquisition and evaluation of data pertaining to low-temperature treatment technologies, which would
contribute towards identification of technology Waste[Goals]Visual test

Inorganic powder / resin, Crystallized / fiber-likeetc.
material, etc.

(i) Study on verification method of low-temperature treatment
solidification• To study techniques (Fig 1) for determining feasibility ofsolidification in advance, with respect to cement solidification andalkali activated materials (AAM) solidification, from among the
low-temperature treatment solidification technologies, along withmeasuring the water absorbency, etc. of waste.

Study in detail

Understand the chemical composition
by means of portable x-ray
fluorescence.
Add cement equilibrium water (until the water completely covers it).
Cement equilibrium water (Na+, K+, Ca+, OH-)
Covered completely bywater (visual test)
Amount ofwater
absorbed,
volume when
fully coveredby waterVolume, mass (bulk Volume, mass (bulkspecific gravity)
specific gravity)

Is the computed
volume of thesolidified cement
allowable?

Add waste to the cement equilibrium
water.
(Prepare a dilute solution.)→ Add chemical admixture, Ca(OH)2.
Reference(water)+ Ca(OH)2
addition+ chemical
admixtureaddition
Is high pH
maintained?

(ii) Acquisition of data on the properties of cement and AAM
solidifications in the slurry• To promote understanding of the solidification process
phenomenon and also to acquire and present data by preparing
samples (Fig 2) by mixing simulated slurries (carbonates, iron coprecipitation) individually and analyzing their strength, eluting
property, etc.

(iii) Investigation of special cements• To investigate cement based materials (Special cements) otherthan Ordinary Portland Cement (OPC) having characteristics
designated to the required properties, in order to study theirapplicability to specific wastes containing components withadverse effects on Ordinary Portland Cement (OPC).

pH (indicator paper) measurement,
property (gas formation)
Evaluation of
Turbidity (scattering)
dispersibility
Let stand measurementorcentrifuge.
Measurementof bed height

It is very likelythat the cement
does not solidify
efficiently ordoes not solidify
at all.

Consider adding
chemical admixture.
Deduce thesolidification formula.

Study the preparation ofsolidified cement indetail.

Figure 1 Example of cement solidification
screening flow

Figure 2 Example of samples(M: Metakaolin, MB (xx)： Blast furnace slag (xx)% substituted metakaolin)©International Research Institute for Nuclear Decommissioning78

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】① Low-temperature treatment technology
i. Acquisition and evaluation of data pertaining to low-temperature treatment technologies, which would
contribute towards identification of technology(i) Study on verification method of low-temperature treatment solidification[Implementation details]
Techniques for determining feasibility of solidification, with respect to cement solidification and AAM
solidification, from among the low-temperature treatment solidification technologies, are studied inadvance in combination with measuring the water absorbency, etc. of waste.
•

Main performance required during the treatment is fluidity and factors influencing it include physicalproperties such as moisture content and form of waste, and chemical properties such as solubility,
reactivity, etc.

•

The verification method for the possibility of solidification are studied so as to simplify operation
management.

•

The formulated determination techniques will be reflected in the process flow, etc. created when theapproaches were studied.
Year

Implementation details

2019

Study on verification method of low-temperature treatment solidification
・Study screening techniques to determine whether or not cement solidification is possible, while focusing on water absorption
rate and solidification properties.

2020

Identification of the scope within which the various types of low-temperature solidification are applicable・Verify the technique of evaluation through experiments pertaining to factors that impact the properties of solidifications andidentify the applicable scope.

©International Research Institute for Nuclear Decommissioning79

【Confidential】(a) Establishment of techniques for selecting the preceding processing method① Low-temperature treatment technology
i. Acquisition and evaluation of data pertaining to low-temperature treatment technologies, which would contributetowards identification of the technology(i) Study on verification method of low-temperature treatment solidification[Study flow][Waste properties measurement method (left side of Figure 3)]
①

②

③

④

Consolidate the properties that have an impact on lowtemperature solidification and set the properties that need
to be measured such as bulk density, absorption rate,chemical reactivity, etc.

Consolidation of
impacting properties andmeasurement methods

Select material that waste is likely to contain and which is
expected to lead to differences in the properties to beverified, as simulated material to be used in the trial run of
the test methods.

Selection of simulatedmaterial with different
properties

Conduct a test trial, wherein individual properties set in ①
are inspected in a small container, using information fromother domains such as agriculture, soil, etc. as well.
Study the flow by combining test methods for the various
properties mentioned above.

[Method of simplified measurement of properties of waste
during solidification treatment (right side of Figure 3)]⑤

Waste properties
measurement method

In addition, study the methods for understanding fluidity
and compressive strength in a much simpler manner
through external measurements such as that of turbidity,
ultrasonic propagation, etc.

Method for simplified
measurement of properties ofwaste during solidification
treatment〇 Fluidity
Experiment pertaining tocreation of models forunderstanding
dispersibility〇 Compressive strength
Tests with small
samples, trial run ofmeasurement methodthat uses ultrasonicpropagation velocity,
etc.

Trial measurement of
individual properties

Selection of
measurement
methodPreparation of proposed
primary screening flow

Preparation of proposed
secondary screening flow

Indication of thepossibility ofsolidification andprovision of atechnique that canbe verified

Fig. 3 Study items for test methods©International Research Institute for Nuclear Decommissioning80

[Study results]- Selection of simulated material with different properties-【Confidential】➢ A combination of properties affecting solidification and physical properties having an impact
on solidification based on the investigation and findings till now, were studied. (Table 1)
Table 1 Correlation between properties of waste and properties of solidifications influencing thepossibility of low-temperature solidification treatment
Properties of waste
廃棄物の性状Properties of solidifications to be considered
考慮すべき固化体特性影響する
Influencing
processes
プロセス

特性

Properties

可使時間
Fluidity
処理プロセ
流動性
Treatment process
ス自体 Homogeneity
均質性Work safety
作業安全性
Strength of solidifications
固化体強度
Dimensional stability
固化後
寸法安定性
After solidification(Solids)
Degradation after solidification（固化体） 固化後変質
Nuclide fixation
核種固定性

形状・
Form/size
大きさ

粒度
Particle size
Density密度
distribution
分布

◎
◎

〇
◎
◎

○
〇

◎

〇

Working life

処理プロセス成立性・固化
Properties that have a major adverse impact on thefeasibility of treatment process and performance of the
体性能に多大な悪影響を
solidifications
及ぼす性状

◎
〇

液体（混練水等）添加時の性状
Properties at the time of liquid (Kneading water, etc.) addition
かさ化学
Chemical
Bulk density Moisture
content
Radiation
含水率
線量doseガス
composition
Water
密度
組成
吸水性
膨潤性
発熱性
溶解性 液性(pH)Expansibility Heatgeneration Gas generation 分散性pH
absorbency
発生 Dispersibility Solubility
◎
◎
◎
◎
〇
◎
〇
◎
－
◎
◎
◎
◎
◎
◎
◎
〇
△
〇
◎
◎
〇
◎
◎
◎
◎
◎
◎
◎
◎
◎
◎
〇
◎
◎
◎
〇
◎
〇
◎
◎
◎
〇
〇
〇
〇
◎
◎
◎
✓（他性
✓Indirect
状で間接
evaluation
using
other
的に評
properties
価）

✓

✓

✓

✓

•*
(Legend) Extent of impact: (Major), (Medium), (Minor), (None), Blank (Impact unknown)※（凡例の意味）影響度：◎（大）、〇（中）、△（小）、－（無し）、空欄（影響が不明なもの）

Select 12 types of simulated material (Silica sand, barium sulfate, etc.) wherein it isassumed that the properties affecting low-temperature solidification (Cement, AMM) are
different, while keeping in mind the waste generated at Fukushima Daiichi Nuclear
Power Station.
©International Research Institute for Nuclear Decommissioning81

[ Study results]- Examples showing trial measurement of individual properties 〇 Water absorbency【Confidential】〇 Dispersibility

After 10minutesVolume of montmorillonite(8g) is approx. 10cc

Status after 10 minutes since
its addition to 40g pure waterFig 4 Confirmatory results for absorbency(Expansibility) (Montmorillonite)➢

5 types of powders namely calcium carbonate(light, heavy), silica powder, silica sand andmontmorillonite are mixed at a liquid to solid
ratio 5 for 2 minutes.
➢ Only montmorillonite absorbs water andexpands inside the entire centrifuge tube. (Fig 4)⇒ As kneading water reduces, waste expands aswell.
⇒ Even if cement solidification treatment is carriedout, it is highly likely that filling rate isconsiderably low.
Cement solidification can be determined to beinappropriate.

Calcium carbonate(heavy) after 180 minutes

Fine powder of silica after
180 minutes

Fig 5 Confirmatory results for solvent based dispersibility(Bulk density)➢

Different solvents are added and stirred and the solution is
let to stand for 180 minutes.
Left side: Pure water; Right side: Saturated Ca(OH)2 withsimulated cement kneading water➢ It is confirmed from the uniformly dispersed condition thatthe extent of dispersion varies depending on the
differences in material and solvent.
➢ Calcium carbonate shows the same behavior in both,water and Ca(OH)2 solution, whereas silica stone
continues to disperse in pure water rather than Ca(OH)2.
⇒ Such changes in behavior due to the combination of solid
powder and solvent affect the success/failure andregulation of conditions during solidification of waste.
©International Research Institute for Nuclear Decommissioning82

【Confidential】[ Study results]- Study of the proposed screening flow (1/2)•

A test method (proposed) consisting of primary screening, which is the method of measuring waste properties bycombining the measurements of individual properties (Step 1 to 5) (Table 2), and secondary screening involving
the method of simplified measurement of properties during solidification treatment (Step 6 to 7) ) (Table 3), wasestablished.

•

From amongst the investigated test methods, primary screening involved testing whether the waste does nothave a detrimental adverse effect on the feasibility of low-temperature treatment solidification process or on theperformance of the solidifications, using a small container (50mL small centrifuge tube).

Table 2 Primary screening (proposed)
Step

Goal
(Property evaluated)

Operation

Tools used

Criteria to move on to the next Step

1

Form, bulk density

Put certain amount of sample in a centrifuge tube and after visuallyinspecting the form and size, tap 20 times and calculate bulk density.

Electronic balanceCentrifuge tube

Reference value not known

2

Absorbency,
expansibility,
impact of addition of kneading
water (heat generation, gas
generation)

Add droplets of simulated kneading water in the centrifuge tube used in
Step 1 and check for the amount of water absorbed, volume change
and gas generation. Check for generation of heat using radiation
thermometer.

PipetteRadiation
thermometer

There shouldn’t be significant expansion, excess
gas generation and heat generation.

3

Dispersibility and particle size
distribution in kneading water

Add simulated kneading water till the upper part of the centrifuge tube,in which Step 2 was performed, and shake it. Thereafter visuallyconfirm the sedimentation status of the sample. (Method using turbidity
meter may also be considered.)

Shaker(Turbidity meter)

Reference value not known

4(Only cement)

Impact of adding cement
suspension type simulated
solution (heat generation, gas
generation)

Add calcium hydroxide in the centrifuge tube in which Step 3 was
performed and check for gas generation. Check for generation of heatusing radiation thermometer.

Radiation
thermometer

There shouldn’t be excess gas generation and heatgeneration.

pH confirmation

Perform centrifugal separation for the cement using centrifuge tube
after the operation in Step 4 and perform centrifugal separation for
AAM using centrifuge tube after the operation in Step 3, and measure
the pH of the supernatant. (It is simple to perform visual test by pouring
BTB solution drop by drop.)

pH test paper, etc.
BTB solution

Should exhibit alkalinity.

5

Confirm from the operation till Step 5 that the waste does not bring about any phenomena (significant expansion, gas formation, heat generation) detrimental to the feasibility (applicability) of treatment
process.

©International Research Institute for Nuclear Decommissioning83

[ Study results]- Study of the proposed screening flow (2/2)•【Confidential】

Fluidity and compression strength are evaluated in a simplified manner in secondary screening (Table 3) but
study will be conducted to identify a method for screening on as small a scale (small quantities) as possible,
since radioactive waste is used.

Compressive strength・ With respect to the strength of a small specimen of approx. 2cm
square (1 inch cube), linearity is the same as Φ5×10 and it will be
helpful in evaluating the strength easily.

angle

Average compressive
strength of target
specimen

AAM

Average compressive
strength of target
specimen

Cementof 2cm square

angle(Reference)

Average compressive
strength of φ5X10 specimen

Required amountof sample
Φ5×10：200mL Form-work made of silicon
2×2×3：12mL rubber with inner dimension

angle

Average compressive
strength of φ5X10 specimen

Fig 6 Correlation between φ5X10 and 7 days compressive
strength of a small specimen

A proposal for secondary screening was madeafter confirming that both fluidity andcompressive strength can be evaluated with
about several dozen mL.

Table 3 Secondary screening (proposed)
Step

Objective(Property evaluated)

Operation

Tools used

Criteria to move on to the next
Step

6

Fluidity
at
the
time
solidification treatment

ofKnead the sample on a small scale (using a hand mixer,etc.) and evaluate the fluidity during solidification
treatment by conducting a small scale flow test. (Performthe test by changing the composition.)

Hand mixer, etc.
Small flow cone

Should satisfy the target flow.

7

Solid strength and improper
condensation (working life) after
solidification treatment

Knead the sample on a small scale (hand mixer, etc.),
place it in a small form-work and cure it; thereafter
evaluate the strength for short term i.e., 7 days, etc.
Further, evaluate coagulation defects.

Small form-workStrength tester

Should satisfy the target strength.

©International Research Institute for Nuclear Decommissioning84

【Confidential】[ Study results]- Study on the technique for predicting fluidity and compressive strength (2/2) Trial screening was performed by means of the proposed test method
using several materials.
Table 4 Example of primary screening verification
using simulated carbonate slurry (Dried and groundproduct)(Cement (Cement)
Propertyevaluated

Form, size, bulk density

Impact of cement
Expansibility, absorbency,
uspension type
Dispersibility, particle
heat generation, gas
simulated solution
pHsize distribution
generation(heat generation, gas
formation, etc.)

Table 5 Primary screening verification results formagnesium sulfate (AAM)
Property evaluatedForm, size and bulk density

Expansibility, absorbency, heat generation,
gas formation

Fluidity
Working life (condensation)

Quick-setting
Operation status
Operation status

After tapping
20 times

Conditions,operation, etc.

Sample 8g

Form: Powder-like
Determination
Bulk density 0.8g/cc

Whether or not
to move on to thenext step○

Without any
liquid passing
through

Adding dropsof pure water

Amount of pure water
Sample 8g
Prepare cement
added (Total) 40cc
Amount of pure watersuspension and add
60 minutes after
added drop by drop 10.7cc
simulated solution.
completion of shaking

After shaking andcentrifugal separation,
add BTB to thesupernatant drop bydrop.

No expansion
No heat generation
No gas generation
Amount of pure water
added: 0.54cc/g

Alkalinity (Reference)
pH=10.8○

No heat generation
No gas generation

○

○

○

Study based on secondary screening

Operation conditions,etc.

Determination

Whether or not tomove on to the next
step

Sample 8g

Sample 8g
Amount of AAM kneading water added dropby drop 5.7cc

Composition: Filling rate 40wt%

Form: powder-like
Bulk density: 0.8g/cc

Right after dripping, wet portion undergoes
quick setting and expands. Severe heatgeneration (approx. 60℃). No gas
generation.

Since the wet portion underwent quick setting rightafter dripping of kneading water, kneading was notpossible. Severe heat generation (approx. 50℃). No
gas generation.

☓(Quick setting and severe heat generation)☓(Quick setting and severe heat generation)○

Determined as inappropriate

Certain decision can be taken regarding the applicability of cement
solidification and AAM solidification.

In addition, study on estimating fluidity and compressive strength, which are the measurement items insecondary screening, was conducted along side as well, using the information obtained from primary
screening.
©International Research Institute for Nuclear Decommissioning85

【Confidential】(a) Establishment of techniques for selecting the preceding processing method① Low-temperature treatment technology
i. Collection and evaluation of data on low-temperature treatment technology to contribute to theidentification of technology(i) Study on the verification method for low-temperature treatment solidification[Summary/ Plan for the following year]➢ Summary to date•

The physical properties that have impact on the feasibility of treatment wereorganized and simulated wastes with those properties were selected.
• Trial methods were performed using 50mL container for the evaluation of physicalproperties and proposal for screening methods was formulated by combining thesemethods.
• Trial screening was conducted using simulated wastes to confirm the possibility of adecision.
➢ Plan for the following year
• The validity and usability of screening flow (proposed) will be verified by using simulated
dry iron coprecipitation slurry and new simulated waste materials.
• Based on the information obtained from the screening flow (planned), estimation methods
for fluidity and strength will be studied. (Study on the information that would be analternative to density information, review and study free water quantity calculation
methods, etc.)• Trials for quantification of data will be conducted.

©International Research Institute for Nuclear Decommissioning86

(a) Establishment of techniques for selecting the preceding processing method【Confidential】① Low-temperature treatment technology
i. Collection and evaluation of data on low-temperature treatment technology to contribute to the identification oftechnology(ii) Collection of data on properties of cement and AAM solidifications for the slurry[Goals]
Prepare specimens by mixing two types of simulated slurries (carbonate, iron coprecipitation) individually to analyze thestrength, eluting property etc., and promote deep understanding of solidification phenomenon as well as acquire and present
the data.
••

The simulated slurry to be used will be prepared using two slurries, carbonate slurry and iron coprecipitation slurry,
adjusted assuming dehydration treatment during storage.
Conduct tests in FY2019, by mixing the carbonate slurry, and in FY2020, by mixing iron coprecipitation slurry and collectthe data.

•

Incorporate the collected data in the study on the applicability of OPC and AAM (approach study).
⇒ Evaluation axis: Solidification performance (leaching rate, hydrogen generation etc.), operability (water content in thewaste, process) etc.
*Based on the results in FY2018, the plan is to conduct a study on the composition that reduces the fluidity of AAM and redu
ces the water content.

FY

Implementation details

2019

Collection of data on properties of cement and AAM solidifications for the slurry
・Collect data on each type of properties pertaining to simulated solidifications mixed with carbonate slurry.

2020

Collection of data on properties of cement and AAM solidifications for the slurry
・Collect data on each type of properties pertaining to simulated solidifications mixed with ironcoprecipitation slurry and conduct a study on the applicability of OPC and AAM based on the data collected.

©International Research Institute for Nuclear Decommissioning87

(a) Establishment of techniques for selecting the preceding processing method① Low-temperature treatment technology
i. Collection and evaluation of data on low-temperature treatment technology to contribute to theidentification of technology(ii) Collection of data on properties of cement and AAM solidifications for the slurry[Tests to be conducted]【Confidential】

Preparation conditions for solidifications and hardening
properties
・Adjust the composition to prepare
the specimens and conduct tests
for fluidity and condensation
performance.

Simulated slurry evaluation

Elution properties
・ Measure the components that are immersed
and eluted in water. Plan for balanced elution
using powder and non-balanced elution using
solidifications.

Gamma radiation properties
・ Evaluation of gamma radiation properties.
Acquire information concerning the measurementof amount of hydrogen and evaluation of alterationof solidifications.

ImmersionliquidSpecimen©International Research Institute for Nuclear Decommissioning88

(a) Establishment of techniques for selecting the preceding processing method【Confidential】① Low-temperature treatment technology
i. Collection and evaluation of data on low-temperature treatment technology to contribute to the identification oftechnology(ii) Collection of data on properties of cement and AAM solidifications for the slurry[Test items and parameters]
漏斗まParameters
パラメーター

Selection
Funnel or
配合 of たはフ
composition flow
test
ロー

選定

Viscosity
粘性
test

試験

Vicat
needle
ﾋﾞｶｰ針
test

試験

Strength
強度
test

試験

MIP

XRD

TG/DTA

EPMA SEM/EDS

Liquidphase
液相analysis

分析

NMR

試験

XRF

Particle
粒度sizedistribution

●
●

●
●

分布

Hardening properties

Simulated slurry●●
炭酸塩スラリー
Carbonateslurry
模擬スラリー評価
evaluation
Iron coprecipitation slurry●●
鉄共沈スラリー●M
Study of conditions for
z
固化体作製
●
preparation of
MB20
solidifications
条件検討
●
MB40
●
OPC
●
●
M
●
●
MB20
Fluidity
流動性
硬
●
●
MB40
化
●
●
OPC
特
●
M
性
●
MB20
固化時間time
Solidification
●
MB40
●
OPC
Curingperiod (Days)
養生期間（日）
●
●
3
●
●
7
圧縮
Compressive
●
●
28
強度
strength●●
91
Long material
age 1
●
●
長期材齢1
Long material
age 2
●
●
長期材齢2
Liquid-solid
ratio: 5 levels
液固比：5水準固●●●10Balanced
化
平衡
●
●
●
●
100
solubility
体 溶
溶解性
●
●
●
500
特 出
●
●
●
●
1000
性 特
●
●
●
2000
Immersion
period (Weeks)
性
浸漬期間（週）
非平衡
●
●
●
●
●
1
Non-balanced
溶出性
●
●
●
●
●
4
solubility
●
●
●
●
●
13
Ray
intensity
線強度
dose 1
γ線照射特性 Integrated
●
●
●
積算線量1
Gamma radiationdose 2（水素発生等）●●●
積算線量2
property (Hydrogen Integrated
generation etc.) Integrated
●
●
●
積算線量3
dose 3・硬化特性、固化体特性の各種試験は、配合選定で決定した代表配合にて実施•Each type of test for hardening and solidification properties is performed in accordance with representative mix determined through mix selection.

Hydrogengas
水素ガ
measureme
ス測定
nt

Eluting property

Properties of solidifications●●●©International Research Institute for Nuclear Decommissioning89

[Study Results]- Production of simulated wastes (Carbonate slurry) ••【Confidential】

Compounding condition (Table 1) for the simulated dehydrated slurry (320 kg) used in the test was thesame as filter press test.
Slurry compressed by filter press, which was to be used at Fukushima Daiichi Nuclear Power Station
was used (Fig. 1) while taking into consideration the storage conditions of actual waste.
Table 11 Compounding condition of carbonate slurry (Per 1m 3)

Name

Additiveamount (kg)RemarksA

Sodium chloride

28.5

Seawater component

B

Magnesium chloride hexahydrate

180

Seawater component

C

Calcium chloride dehydration product79.5

Seawater component

D

Magnesium sulphate heptahydrate

6.82

Seawater component

E

Sodium carbonate

152

F

48% Sodium hydroxide130

Added so that slurry is maintained at
pH12Fig. 1 Slurry before compression (Left) and slurry after compression (Right)

Useful knowledge obtained
through tests

Foaming occurred whilewashing when the filter cloth
for the dehydration treatment
was replaced.
It is likely to happen evenwhen it is implemented on
actual equipment.

Perform forecasting test using
washing waste water in order
to confirm that there is no
obvious impact onsolidification.
©International Research Institute for Nuclear Decommissioning90

【Confidential】[Study results ]- Study on mixing methods for simulated wastes ➢➢Properties of dehydration products• It is like papier mâché which gets deformed whenpressed firmly by hand, and has no fluidity. (Fig. 2)⇒ When kneaded as is, it is not homogenized. (Fig. 3)• Water/ solid ratio in the dehydration product is approx.
0.47.
⇒ Extra water needs to be added in order to mix it in thedehydrated state and it is likely to produce excess water.

Fig. 2 Appearance of
dehydration productFig. 3 Dehydration product whenkneaded as is

Mixing method for simulated wastes
Since the purpose of this research is to obtain fundamental data, such as properties during solidification and reaction
between solidified materials and waste and so on, which could form an evaluation axis between treatment
technologies, slurry particles and solidified particles were mixed uniformly for the test in order to be able to obtain dataon properties.

After drying (105℃, 24h) (Fig. 4)
Pulverized, and pulverized and sieved product(Fig. 5) is mixed with the powder of solidified
material in advance.

Fig. 4 Drying in progressFig. 5 Pulverized and sieved product after drying

As a result of the study, homogenous solidified sample was produced by mixing the dehydration product withdry powder in order to secure fluidity that will enable kneading without producing excess water.
©International Research Institute for Nuclear Decommissioning91

[Study Results]- Newly identified issues in actual implementation
based on the study on mixing methods -【Confidential】➢ It is necessary to establish a policy at the earliest in connection with theimplementation of low-temperature treatment for dehydrated slurry sincetreatment process/ facilities, waste filling quantity, etc. are likely to varyconsiderably depending on the decision on whether to go for homogeneous
solidifications or filled solidifications.
➢ If filled solidifications are used and if the solid needs a certain strength, it isnecessary to separately confirm the strength by using the strength of thecompressed material itself and filled quantity as parameters.
➢ In case of homogeneous solidifications, a study on appropriate pre-treatment
methods such as re-fluidization of slurry after compression and so on must be
conducted.
➢ Depending on the storage status of compressed substances, it is expected thatdrying and compaction of compressed substances will take place due to drying
by contained radioactive decay heat and compression by laminated layers.
Hence the above study that takes this into account, is necessary.

©International Research Institute for Nuclear Decommissioning92

【Confidential】[Study Results]- Solidified cement preparation conditions:
Test method and criteria for the evaluation of composition adjustment -•Criteria for determining the composition are set as follows.
Fluidity

Test method

J14 Funnel flow time
(JSCE-F-514
Test method for fluidityof filled mortar)

J14 funnel flow time
=2 - 4 seconds

Criteria

Condensation

Compressive strength
Compressive
strength(JIS R 5201
Physical testing
methods for
cement)Initial setting time
(JIS R 5201
Physical testing
methods for cement)

Hardening took place without
quick setting and without
bleeding within 24 hours

Φ 5 x 10 cm
20℃ Sealed
curing1.47 N/mm2 or above(Material age 28 days)

Definition of filling rate

Filling rate (wt%)＝

Wastesx 100

Base material powder + Base material liquid + Waste

Filling rate 30wt%
Water/solid ratio 45wt% (Water-cement ratio wt80%)

Filling rate 50wt%
Water/solid ratio 45wt% (Water-cement ratio 164wt%)

Water
31wt%

Water
31wt%

Cement39wt%

Cement19wt%

Dried and crushed
carbonate

30wt%
Dried and crushed carbonate

50wt%*When water/solid ratio is constant, the ratio of cement and AAM powder reduces as the filling rate increases.

©International Research Institute for Nuclear Decommissioning93

[Study Results]- Solidified cement preparation conditions (1/ 2) -【Confidential】

If simulated waste is filled into the solidified cement composition (water-cement ratio=45%) that was studied
in the FY2018 project, there will be a considerable decrease in fluidity and a kneaded product without fluidity
or without kneadability may be obtained. Hence, the properties data on fluidity, next day-hardenability(including the presence of bleeding) and compressive strength was obtained by using a composition withincreased W/C and changing the filling rate of the simulated waste, to identify a composition that meets theestablished criteria.

Composition studied

Fluidity

Table 2 Composition studied
Filling rate of

模擬廃棄物充填率（％）
simulated wastes

W/C

Remarks
備考

0

15

20

25

30

35

40

45

50

100

－

●

●

●

●

－

●

－

－

150

－

－

●

－

●

●

●

－

●

200

－

－

●

－

●

●

●

●

●

45

●

－

－

－

－

－

－

－

－ 2018年度配合
composition

4.5

rate

3.5
3.0
2.5

2.0
1.5
Filling rate
充填率15Filling rate
充填率25Filling rate
充填率35Filling rate
充填率45

1.0
0.5

充填率20
充填率30
Fillingrate
充填率40Fillingrate
充填率0Filling rate
Filling rate

•

0.0
10
FY2018

Not•Physical properties「－」：未実施、「●」：物性取得を実施
implemacquired
ented
Strength expression is extremely

low and the composition may notsolidify. Hence, not implemented.

•

Filling←W/C=45、充填率04.0•J14 funnel flow time
J14漏斗流下時間(秒)(seconds)

W/C was set at 3 levels (100, 150, 200wt%) andeach physical property was acquired by changing
the filling rate of the simulated waste.

Fluidity not achieved

60
110
Water/ 水/固体比（％）
solid ratio

160

Fig. 6 Relation between water/solid
ratio and flow time

If the relationshipbetween water/solid
ratio and flow time isorganized in connection
with the level at which
fluidity that makes
funnel test possible isobtained, the flow timeincreases as water/solid
ratio decreases.
All compositions thatmade funnel test
possible were within thescope of criteria.

*Water/solid ratio = Water weight / (Cement weight + Dry weight of
the waste)©International Research Institute for Nuclear Decommissioning94

【Confidential】[Study Results]- Solidified cement preparation conditions (2/ 2) -Compressive strength

Condensation• Bleeding occurred on thenext day in the case ofcomposition with W/C=150
and 200 and filling rate at
20wt%.
• In other compositions,
bleeding did not occur on
the next day and thesurface hardened.

45
40•Filling rate
←W/C=45、充填率035Decrease in strength

30
25

20
15

Expansion

10
5
0
10

60
110
Water/
solid
ratio
水/固体比（％）

充填率15Filling rate
Filling rate
充填率30Filling rate
充填率45

充填率20
Filling rate
Filling rate
充填率35Filling rate
充填率50

160

28 days compressive
28日圧縮強度(N/mm2)
strength

28 days compressive
28日圧縮強度(N/mm2)
strength• Strength decreases
significantly if simulated
waste is filled.
• Strength decreases aswater/solid ratio increases.
• Strength decreases as thefilling rate increases.
• It is assumed that decreasein the amount of cement inunit volume is the cause fordecrease in strength as thefilling rate increases.

10
9
8
7
6
5
4
3
2
1
0

Increase infilling rate

1.47N/mm2

10

60

110

160

Water/
solid ratio
水/固体比（％）

充填率25Filling rate
Filling rate
充填率40Filling rate
充填率0Fillingrate
充填率15Fillingrate
充填率25Fillingrate
充填率35Filling
rate
充填率45

充填率20
Fillingrate
Fillingrate
充填率30Fillingrate
充填率40Fillingrate
充填率50Fig. 7 Relation between water/solid ratio andcompressive strength

The composition “W/C=100, waste filling rate 30wt%” wasselected from among the compositions that met the criteria taking
into consideration the fluctuations in each physical property.
©International Research Institute for Nuclear Decommissioning95

【Confidential】[Study Results]- Hardening time, compressive strength -Compressive strength

Hardening time

Solids containing carbonate
slurryCompressive
strength

Compressive
strength

Solids containing carbonate
slurryOnly OPC

End

Time (minutes)

Start

Only OPC
Quantity of retarding
agent added

The start and end of the solidified cement
containing carbonate slurry (30 wt%) isfaster than that of OPC alone, but the startand end times can be adjusted by the
combined usage of a retarding agent.

Material age

Material age

Strength of solidified cement containing carbonate slurryshows a significant decrease in strength when comparedto OPC base material only.
Strength for material age of 7 days and 28 days is almost
the same and there is no increase in strength.

Retarding agent: Frolic T (Super retarding agent for concrete)©International Research Institute for Nuclear Decommissioning96

【Confidential】[Study Results]- XRD, TG/ DTA -TG/ DTA

Qualitative analysis of minerals

NB

40000
30000

N

20000

N

10000

N
H
P

P
B
P

C NN

C C CCB
C
N
N

20
10

0
-10

-10

Ca(OH)2分解
Decomposition

-20

-30
CaCO3Decomposition
分解-40

Mg(OH)2分解
Decomposition

-20

-50

B

-30

-60
TG(%)

200

P P P

0

10 15 20 25 30 35 40 45 50 55 60

DTA(μV)

-70

-40

0

P
5

Solidified cement containing
炭酸塩スラリー含有セメント固化体
セメント固化体
carbonate slurry

0

400

600

800

-80
1000

Temp［℃］

Rate of decrease in mass of solidified cement containing
carbonate slurry at 50 -100℃ is approx. 32%.

2Θ［°］

10

30

Dried carbonate slurry

20

Dried
carbonate slurry
炭酸塩スラリー乾燥品

OPC固化体Solidified cement containing carbonate slurry

100

Ca(OH)2Decomposition
分解

The OPC solidifications and minerals constituting
the carbonate slurry were observed in thesolidified cement containing carbonate slurry.

0
-10

-10

-20
-20

Mg(OH)2Decomposition
分解-30
CaCO3Decomposition
分解-40

DTA［μA］

OPC solidifications
OPC固化体セメント固化体TG［%］[count]Strength
強［count］

50000

P:Portlandite (Ca (OH)2)
C:Calcite (CaCO3)
B:Brucite (Mg (OH)2)
H:Halite (NaCl)
N:Northupite (MgNa3 (CO3)2Cl)30

DTA［μA］

C

TG［%］

60000

10

-50
-30

-60
TG(%)

DTA(μV)

200

400

-70

-40
0

600

800

-80
1000

Temp［℃］©International Research Institute for Nuclear Decommissioning97

【Confidential】[Study Results]- Preparation conditions for AAM solidifications (Basic conditions of composition) •Study on the adjustment of AAM composition with a view to identifying a typical composition that meets the evaluation
criteria of fluidity, condensation performance, and compressive strength, which are the basic properties of solidifications.

•

Specifically, fluidity, condensation performance and compressive strength were evaluated by changing the “Design index(water/ total powder ratio, base material Na/ Si molar ratio) to design the composition of AAM base material
solidifications” and the “Waste filling rate” which were identified in the FY2018 project.
MK/ Basematerial
powder[wt%]

BFS/ Basematerial
powder[wt%]

Si/ AI molar
ratio in basematerial[mol/ mol]Wastefilling rate[wt%]

Water/ solid ratio[wt%]

Na/ Si molar ratio
in base material[mol/ mol]

M

100

0

1.8

0

150

0.84

MB20

80

20

1.8

0

110

0.50

MB40

60

40

1.8

0

100

0.63

M

100

0

1.8
30,
40,
50.

Verification ofcomposition
impact by making
changes

Verification ofcomposition
impact by making
changes

Types

MB20

80

20

1.8

MB40

60

40

1.8

M：
Metakaolin

BFS：
Blast furnace slag fine powder

FY2018FY2019

Solid = Base material powder + waste solid phase

M: Metakaolin only, MB20： 20wt% of metakaolin replaced with blast furnace slag, MB40：40wt% of metakaolin replaced with blast furnace slag
©International Research Institute for Nuclear Decommissioning98

[Study Results]- Preparation conditions for AAM solidifications:
Changes due to the composition of properties of solidifications Legend:

MB20

M

Fluidity

MB40

Condensation

Compressive strength

Filling
rate
30wt%

M
M

MB系

Filling rate
40wt%

Compressive strength

MB system
Start time

J14 funnel flow time

Filling rate 50wt%
Fillingrate
30wt%

Na/ Si ratio

Water/ solid ratio (wt%)••

Increase in filling rate (Water/ solid
ratio constant) ⇒ Increase in fluidity
Increase in water/solid ratio (Fillingrate constant) ⇒ Increase in fluidity••【Confidential】

Changes occurring due to basematerial Na/ Si molar ratio which
may vary depending on powder
type (trends same as base material;
impact of filling rate is minor)
Increase in water/solid ratio ⇒Delayed setting

Fillingrate
50wt%

Fillingrate
30wt%

Fillingrate
40wt%
Water/ solid ratio (wt%)••

Increase in filling rate, water/solid
ratio⇒ Decrease in strength
Blast furnace slag replacement:
Decrease in Na/ Si molar ratio ⇒Decrease in strength(AAM composition properties)➢ If water/solid ratio is increased, fluidity is secured but strength deteriorates.
➢ If filling rate increases (when water/solid ratio is constant), fluidity increases, but strength deteriorates.
➢ Condensation performance is strongly influenced by Na/Si molar ratio. The effect is reversed depending on the inclusion ofslag.
⇒ Since the changing trend based on the composition varies for each property, it is important to identify the composition that
fulfills each property requirement.
©International Research Institute for Nuclear Decommissioning99

【Confidential】[Study Results]- Preparation conditions for AAM solidifications: Selection of composition that meets the evaluation
criteria -Compressive strength

Compressive
strength for thecomposition thatmeets theevaluation criteria of
fluidity andcondensation
performance

Out of the compositions
that meet the evaluation
criteria, the following
composition was selectedtaking into consideration
the fluctuations in eachphysical property.

Metakaolin only

Strengthdeterioration

Blast furnace slag 20%
replacement
Blast furnace slag 40%
replacement

Filling rate

Fig. 8 Filling rate and compressive strength for dry carbonate slurry

Waste filling
rate
[wt%]

Water/ solid
ratio
[wt%]

Na/ Si molar
ratio in thebase material[mol/ mol]

J14 funnel
flow time[sec]

Start [min]After 24h

M

30

92

1.2

3.6

470

End

5.3

MB20

30

86

0.7

2.8

430

End

8.3

MB40

30

80

0.8

3.0

460

End

7.0

Type

Setting

28 daysCompressive
strength[N/ mm2]

Like cement, composition “Waste filling rate 30wt%” was selected for AAM as well.
©International Research Institute for Nuclear Decommissioning100

【Confidential】[Study Results]- Data acquisition: Hardening properties ○Hardening (Changes in viscosity)○Irradiation test (During hardening)
G value: Number ofmolecules produced
per 100ev of energy
absorbed by the matter(During hardening)

Filling rate
Filling rate

(MB40)(M)

G value(During hardening)

・ Viscosity at initial stage is of same value.
・ Viscosity tends to start increasing earlier.
・ Change in the gradient for viscosity increase is slow (graphicalform of time variation).

It also demonstrated the occurrence of complex
changes in physical properties which affect thetreatment.

(During hardening)

Integrated dose

Fig. 9 Effects of viscosity during AAM solid hardening
and waste filling rate

When 30wt% of slurry is included(During hardening)

Fig. 10 G value during hardening•

Gamma rays were irradiated on the paste
immediately after kneading to investigatehydrogen generation behaviors while hardening
was in progress.

•

G value during hardening was bigger in the orderof M<OPC<MB20<MB40.

•

Efforts are being continued towards cutting ofsolidifications and analysis of internal bubble
condition.
In addition, collection of data such as elution test, irradiation test (simulated solidifications) etc. also continues.
©International Research Institute for Nuclear Decommissioning101

[Study Results]- Data acquisition: Gamma radiation properties -【Confidential】

Table 3 Irradiation conditions
Gamma radiation test for solid basematerial (FY2018 project)

Specimen

Irradiation conditions
Radiation source

Takasaki Advanced Radiation Research Institute, the National Institutes for Quantum andRadiological Science and Technology

OPC, M, MB20, MB40
(Base material only)
Curing time: 28 days

OPC, M(Contains simulated carbonate slurry
30wt%)
Curing time: 28 days

Fixed irradiation (Fix the irradiation container towards radiation source)
Co-60

Dose rate

Approx. 1.8 -5 kGy/ hIntegrated dose
(Plan)

3 kGy, 10 kGy, 30 kGy

OPC base materia
only
Simulated solid
OPC
M base material
only
Simulated solid M

G value

Test site

Gamma radiation test for solidifications
containing simulated slurry (This project)•

Test contents•Integrated dose

Fig. 12 G value of hydrogen gas with respect to integrated dose

When G value of hydrogen gas in solidifications
containing carbonate slurry was compared with
irradiation results for only the base material in the
FY2018 project, the G value in case of solid containing
slurry tended to be higher than that of the basematerial of both OPC and M.
After 3 kGy
After 10 kGy
After 30 kGy
irradiation
irradiation
irradiation
Fig. 11 Appearance before and after irradiation of solidifications
containing simulated slurry
There was almost no change in the appearance of OPC before and after
irradiation at any integrated dose. No change was noticed in M whenirradiated with integrated doses of 3 kG and 10 kGy, but the solid after
irradiation with 30 kGy was brittle and crumbled in the irradiation
container and did not maintain its original shape.
Before irradiation

・Irradiation test was performed in the same way forsolid containing carbonate slurry (MB20, MB40).
Cracks etc. did not occur after irradiation.
・Currently, analysis is being conducted on hydrogengas etc.
©International Research Institute for Nuclear Decommissioning102

[Study Results]-Data acquisition: Gamma radiation properties, dissolution test -Compressive strength after irradiation test

Dissolution test Being performed currently

M containing
carbonate slurry

Axial compressive
strength

OPC containing
carbonate slurry【Confidential】

Before irradiation

Fig. 13 Compressive strength before and after
irradiation
Compressive strength of the solidifications containing
carbonate slurry• No change in strength was noticed in OPC before and after
irradiation.
• In case of M, apart from brittleness and crumbling due toirradiation with 30 kGy, there was also slight deterioration instrength after irradiation. However, the strength wasmaintained above the strength standards (1.47 N/ mm 2).

(Items for which data is being collected)◼ Continuation of dissolution test (Verification of each type of property such as solubility of solidifications, etc.)◼ Confirmation of water content (TG/ DTA), mineral phase (XRD) after gamma irradiation◼ Confirmation of leachability of simulated nuclides©International Research Institute for Nuclear Decommissioning103

【Confidential】[Study Results]- NMR of solid containing carbonate slurry NMR

29Si-NMR and 27Al-NMR measurements were carried out for the initial samples ofsolidified cement and AAM solidifications (3 types)

29Si-NMR spectrum of solidified cement

27Al-NMR spectrum of solidified cement

containing carbonate slurry

containing carbonate slurry

29Si-NMR spectrum of AAM

solidifications* containing carbonate
slurry27Al-NMR spectrum of AAM

solidifications* containing
carbonate slurry*AAM solid in the graph is the solidified sample of metakaolin only without blast
furnace slag

⚫ In future, qualitative analysis of attributes will be carried out taking into account the results ofqualitative XRD analysis of minerals.
©International Research Institute for Nuclear Decommissioning104

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Low-temperature treatment technology
i. Acquisition and evaluation of data on low-temperature treatment technology that would
contribute to the identification of technology(ii) Acquisition of data on properties of cement and AAM solidifications in the slurry[Summary/ Plan for the following year]➢ Current Summary•

According to the studies conducted so far, the simulated carbonate slurry was prepared bysimulating the process of dehydration by means of filter press, etc.

•

The composition of cement and AAM was studied using the obtained carbonate slurry. The
composition that fulfilled the evaluation criteria had a waste filling rate of up to 45 wt% (weightof dry powdered particles) for cement and up to 40 wt% for AAM.

•

For a proper comparison, a composition having a waste filling rate of 30 wt% (weight of dry
powdered particles) was selected as the representative composition considering variations inproperties.

•

After preparing solidified samples from the determined composition, data on viscosity
measurement, irradiation testing, etc. was collected.

➢ Plan for the following year•

Simulated iron co-precipitation slurry will be made to collect data in a similar manner.

•

The obtained results will be summarized, which will be utilized to clarify the scope of application
for studying the approach.

©International Research Institute for Nuclear Decommissioning105

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Low-temperature treatment technology
i. Acquisition and evaluation of data on low-temperature treatment technology that would
contribute to the identification of technology(iii) Investigation of special cements[Implementation Details]
In order to study the applicability to secondary waste generated from contaminated water
treatment, which have components with an adverse effect on Ordinary Portland Cement(OPC), conduct an investigation of cements other than OPC having the essential special
properties.
(Overview of investigation)
After deciding the types of special cements to be investigated, investigate the properties of solidifications
focusing on the essential properties, as well as the usage results and research examples pertaining to theuses for solidification of radioactive waste, and in addition, select the special cement solidifications to betested and studied from the viewpoint of ease of availability of raw materials within Japan or feasibility of
conducting tests.
FY

Implementation Details

2019

Selection of special cements with high usage feasibility (Literature survey)
Viewpoint of investigation: Cements that reduce the impact of sodium carbonate ascompared to OPC (cements with a wider scope of application as compared to OPC)
Evaluation of base material
Performance evaluation of base material that uses special cement

2020

Study that will contribute to the selection of the applicable range of various lowtemperature solidifications (Addition of waste simulants, sodium carbonate andborate salt, etc. )©International Research Institute for Nuclear Decommissioning106

[Study Results]- Selection of special cements ➢➢➢【Confidential】

An investigation was conducted on the OPC solidification inhibitors until last year, and borate salt andsodium carbonate, etc. were identified as substances that were likely to have an impact even whenthese were present in small amounts in the secondary waste generated from contaminated water
treatment.
In conjunction with the AAM solidification applicability study, effective low-temperature treatment options
for these substances were searched.
Special cements having a high degree of technical maturity were selected comprehensively (Table 1),
and their applicability was studied.
Table 1 Selected special cements(1) JIS-standard types of cements
Portland Cement
・Ordinary Portland cement
・High-early-strength Portland cement
・Ultra high-early-strength Portland cement
・Moderate-heat Portland cement
・Low-heat Portland cement
・Sulfate-resistant Portland cement
Mixed cement
・Blast furnace slag cement (Types A, B, C)・Fly-ash cement (Types A, B, C)・Silica cement (Types A, B, C)
Other cements
・Eco-cement (Ordinary, Rapid-setting)(2) Types of special cements (Other than JIS-standard)
Types of cements with Portland cement as the base・Expansive cement
・Binary-system low-heat cement
・Ternary-system low-heat cement
Types of cements in which the elements or particle size composition of
the Portland cement has been changed
・White Portland cement
・Cement-based solidification material
・Ultra fine particle cement
・High belite cement
Types of cements having elements different from Portland cement
・Ultra rapid-setting cement
・Non-hydraulic cement
・Alumina cement
・Other cements (Calcium Sulphoaluminate (CSA) cements, Magnesiaphosphate cements (MFC) )©International Research Institute for Nuclear Decommissioning107

【Confidential】[Study Results]- Viewpoints of investigation and classification of the targeted special cements [Viewpoints of investigation]
 Usage results and research examples pertaining to the uses for solidification of radioactive waste
Investigation of presence of the usage results and research examples of said cements
 Impact of elements causing adverse effect on OPC
Inferences drawn from the information about the magnitude of influence of the components "Borate salt" and "Sodium carbonate(Na2CO3) " in the slurry, or the inferences drawn from the components of cement
 Characteristics that specialize in the required properties
Attention to "High fluidity" from the viewpoint of feasibility of the solidification process or the suppression of the amount ofhydrogen generated (decrease in the ratio of water and cement)
Attention to "High heat resistibility" from the viewpoint of waste nuclide inventory that can be handled or the possibility ofleading to an increase in the waste filling rateTypes of cements with Portland cement as the base
Expansive cement
Low-heat cement of multi-component systems(Investigation that includes admixture mixed cement
which exceeds the substitution rate of JIS)
 Types of cements in which the components orparticle size composition of the Portland cement
has been changed•
White Portland cement,•
Cement-based solidification material,•
Ultra fine particle cement,•
High belite cement



⚫

⚫

•
•

Classification of
CementSummarized
features⚫

As OPC is the base, the reactions or the properties etc.
are similar to those of OPC.
Properties that are different from OPC, arestrengthened as the quantity of admixture increases.

⚫

Types of cements having components
different from Portland cement•
Ultra rapid-setting cement•
Non-hydraulic cement•
Alumina cement (Includes cements withalumina cement as the base)•Calcium Sulfoaluminate cements (CSA)•
Magnesia-phosphate cements (MFC)

As the components contained are significantly
different from OPC, the reactions or propertiesare different.
The properties tend to be extremely different
since adjustments are made assuming a specificusage.
©International Research Institute for Nuclear Decommissioning108

[Study Results]- Applicability study based on the features of each special cement -【Confidential】

Example of investigation results of special cements
Expansive cement

•
•
•



Low-heat cement of multicomponent systems(Investigation that includes
admixture mixed cementthat exceeds the substitution
rate of JIS)••

Alumina cement (Includes
cementswithaluminacement as the base)••Magnesia-phosphate
cements (MFC)••◼

Blast furnace slag high substitution cement⇒ It is likely to have excellent applicability to the sodium carbonate mixed waste. (Carbonate slurry)Fly-ash mixed alumina cement⇒ It is likely to have excellent heat resistance and fluidity. Also, the impact of components may be different from that of ordinary
cement.

Selection

Selection◼

No usage results and research examples pertaining to the uses for solidification were found.
There are results indicating control of cracking (near surface pit) when used in disposal facilities.
Based on the composition of the cement, it is inferred from the observations of this investigation that there is no major difference from ordinary
cement.
⇒ It is inferred that the usage possibility is low
No usage results pertaining to the uses for solidification were found within Japan.
As regards research, usage possibility has been reported during the case of research of sodium carbonate solution solidification by means of
blast furnace slag high substitution cement (1:9)•
An applicability study of fly-ash and blast furnace slag high substitution cement ("Cast Stone") as the basic solidification material for low-levelliquid waste (Nitric acid Na) was conducted overseas at Hanford.
⇒ Since the proportion of ordinary cement is extremely low in the "Finely powdered blast furnace slag high substitution cement (1:9) ", the level ofimpact of sodium carbonate and boric acid are different and hence it was determined to be usable.
No usage results pertaining to the uses for solidification were found.
There is a research example in which the material design for preventing the reduction in strength caused due to the defects of a stand-alone
system of alumina cement such as the initial hydration heat or progression of aging of material, wherein fly-ash or blast furnace slag is mixed
with alumina cement, was researched. Since the composition is quite different from ordinary cement, there is a possibility that the impact of
sodium carbonate and borate salt will be different.
•
Heat resistance is high. As regards fluidity, it has been reported that high fluidity is obtained even with an ultra-low water cement ratio when an
efficient water reducing agent is used.
⇒ It is determined to be usable from the viewpoint of impact of harmful components and heat resistance with respect to ordinary cement.
⇒ "Alumina cement + Fly-ash" is selected since it is considered important to control hydration heat and to control the reduction in strength that iscaused by the progression of aging of material.
No usage results pertaining to the uses for solidification were found.
As for uses for solidification, there is an applicability study of "Ceramicrete" conducted at Hanford, which considers Magnesia-phosphate
cements (MFC) for solidification of Nitric acid (Na) liquid waste.
•
Since the composition is quite different from ordinary cement, there is a possibility that the impact of sodium carbonate and borate salt will bedifferent.
⇒ Inferred to be usable. ⇒ But testing is deemed difficult based on the manufacturing conditions (material and composition) and the presence ofreference literature.

Study the representative composition, and acquire the basic physical properties like compressive strength etc. for the obtained representative
composition.
©International Research Institute for Nuclear Decommissioning

109

[Study Results]- Relation between composition and fluidity/ condensation performance -Plan for the current fiscal year (Verification of basic required performance)
Evaluation of setting properties and fluidity with respect to the solidifications manufactured by changing the composition (water/fineparticle ratio)•
Selection of a representative composition and evaluation of compressive strength, pore size distribution, constituent phase (XRD) andmoisture conditions (free water and absorbed water)
A representative composition was decided by measuring the fluidity and setting properties after changing the composition(water/fine particle ratio).
•

Condensation (Vicat needle test)

Fluidity (J14 Funnel flow test)
As CB90 (water/fineparticle ratio is 40%)
is an excessively
thick paste, it does
not pass through thefunnel.

OPC
CB90: Fly-ash mixed Alumina cement(AF): Ordinary Portland Cement(OPC): Blast furnace slag highsubstitution cement (CB)

CB90

Final setting time

AF20
J14 Funnel flowtime➢

【Confidential】• Fluidity: Fly-ash mixed Alumina cement > Ordinary cement >Blast furnace slag high substitution cement• Alumina cement has high fluidity as found in the literature
survey.

Representative
composition: Fly-ash mixed Alumina
cement (AF)AF20

: Ordinary Portland Cement(OPC): Blast furnace slag highsubstitution cement (CB)OPC

Start time

Water/fine particle ratio• All the solidifications have a sufficient handling time (worklife) and the final setting takes place within 24 hours, so it isexpected that the manufacturing process can be established.

Ratio of water/ fineparticle [wt%]J14 Funnel flow time
[sec]

CB90

55

AF20
OPC (For comparison)

Types

Setting is slow forcompositions inwhich the water/fineparticle ratio is high.

SettingStart [min]After 24h

3. 6

790

1190

40

4. 7

310

450

45

4. 2

270

400

CB90: OPC/Blast furnace slag fine powder was mixed in the ratio of 10/90, AF20: Alumina cement/ Fly-ash was mixed in the ratio of 80/20.
©International Research Institute for Nuclear Decommissioning110

【Confidential】[Study Results]- Data on the properties of the representative composition and the representative composition solidifications -•Solidified specimens were manufactured for the representative composition and properties such ascompressive strength, amount of free water etc. were acquired.
Amount of free water (Material aging changes)

Compressive strength (Material aging changes)

The amount of freewater in the fly-ash
mixed Alumina cement
is minimal. (Since theratio of water/fineparticles is small.)Amount of freewaterCompressive
strength

In case of the blast
furnace slag highsubstitution cement,
although the initialstrength development
is low, with thepassage of time, theimprovement instrength is large.

Material age
Material age⚫

Strength at material age 28 days: Ordinary cement > Blast furnace
slag high substitution cement ≒ Fly-ash mixed Alumina cement⚫

Amount of free water at material age 28 days: Fly-ash mixed
Alumina cement < Ordinary cement < Blast furnace slag highsubstitution cement

In the case of CB90, OPC : Blast furnace slag fine powder was mixed in the ratio of 10:90
In the case of AF20, Alumina cement : Fly-ash was mixed in the ratio of 80:20•

Additionally, data on the analysis items (pore size distribution, constituent phase (XRD) ) is being
collected.
©International Research Institute for Nuclear Decommissioning111

【Confidential】(a) Establishment of techniques for selecting the preceding processing method
 Low-temperature treatment technology
i. Collection and evaluation of data on low-temperature treatment technology to contribute to theidentification of technology (iii) Investigation of special cements[Summary/ Plan for the following year]➢ Current Summary•••

A literature survey was carried out for the special cements, and blast furnace slag high substitution
cement and the fly-ash mixed Alumina cement were chosen for testing as alternative materials for OPC.
Basic data such as fluidity and compressive strength etc. were acquired by using the base material.
Data on some data items (pore size distribution, constituent phase (XRD) ) is being collected.

➢ Plan for the following year••

The impact on substances (sodium carbonate, borate salt etc. ) that have an adverse effect on the
OPC solidifications, simulated waste (various dried and powdered slurries) and of heating/ drying will
be verified on the two types of special cements selected this year.
When using a special cement, applicable range of waste properties for which cement solidification isavailable will be studied and examined. (Provided as feedback to ② Study on the approach forevaluating the applicability of treatment technologies)
Table 2 Study items of special cements
Applicability evaluation axis
for specific waste
Ordinary Portland Cement
properties, etc.

Blast furnace slag highsubstitution cement

Fly-ash mixed Alumina
cementSodium carbonate

Cement wt% 2% or less

FY2020 Study

FY2020 Study

Boric acid

Cement wt% 0. 1% or less

FY2020 Study

FY2020 Study

Slurry simulated waste
permissible filling rate

Carbonate: 30 to 40%(*Selected from i(ii) study)

FY2020 Study

FY2020 Study

Heating/ drying load impact

Selected from ii (i) study

FY2020 ii (i) Study

FY2020 ii (i) Study©International Research Institute for Nuclear Decommissioning112

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications[Goals](iii) Investigation of the influencing factors that contribute to longterm changes in properties and evaluation of the propertychange behaviors• To calculate the stability phase assuming the shift to the
crystallizing phase, rise in temperature (up to 60℃) and theleaching alteration. (Figure 2)

Amount of radioactivity (Bq/container)

Fig. 1 Evaluation of the relation between the Cs inventory and thetemperatures of solidifications (Example)

Volume(ii) Evaluation of the relation between the Cs inventory and thetemperature of solidifications (Figure 1)• To clarify the changes in the heating conditions of thesolidifications inside containers of different shapes (several typesof containers).
• To investigate the values of density and thermal conductivity for
mixed waste, reflect them in the analysis, and clarify themagnitude of the effect.

Maximum
temperatures(i) Investigation of change in properties of solidifications due toheating, etc.
• To conduct a study on consolidation of factors that lead tochange in properties of each solid material and the possibilities
therein, based on the study of approach implemented in FY2018
and the investigation results of the methods of measuring thephysical properties of the solidified materials. To clarify thechange in properties of solidifications during the heating anddrying states in low-temperature treatment, based on the resultsof TG acquired in FY2018.

Slurry content

Fig. 2 Study of composition of the mineral phase when slurry is
contained (Example)©International Research Institute for Nuclear Decommissioning113

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications(i) Investigation of change in properties of solidifications due to heating[Implementation Details]
The change in properties while in storage during the heating and drying states was investigated in orderto study the effects of storage.
The findings regarding the possibility of change in properties were consolidated based on the survey ofliterature and investigation of test data obtained from other items, and tests were conducted to collect dataon drying, etc., which was insufficient.
In FY2019, the findings regarding the possibility of change in properties were organized; and in FY2020,
the insufficient data will be collected.
FY
2019

Implementation Details
Investigation of changes in performance due to heating and drying
・Organization of findings regarding the possibility of change in properties based on the literature
surveyEvaluation testing of the base material
・Evaluation of the "Effect of drying" and the "Compound effect of heating and drying"
Evaluation of changes in the performance of the base material decided under "①-i-(ii) Collection ofdata on properties of cement and AAM solidifications for the slurry”, caused due to heating and drying(The wide temperature zone and the wide drying strength zone were set up and verified.)2020

Evaluation testing of the base material
・On receiving the FY2019 results, the range (temperature, drying strength) in which performancechanges are significant will be narrowed down and verified.

©International Research Institute for Nuclear Decommissioning

114

[Study Results]- Literature survey results for setting the test conditions -Findings regarding the possibility of change in properties have been organized on the basis of literature
survey, etc.
AAM solidifications<Heating> Structural changes due to drying and heating• 70 to 100C: Dissipation of free water inside gaps and frompartial hydrates (Ettringite) ⇒ Structural shrinkage, coarseningof gap size
• 100C to 450C: Dehydration of C-S-H ⇒ Further structural
shrinkage and coarsening of gap size
• 450C and more: Decomposition of Ca(OH) 2, CaCO3(Conversion into CaO) ⇒ Loses the solidified appearance andthere is a significant deterioration in strength (Figure 1)<Dryness (Humidity) > Mainly structural shrinkage due todryingRH (relative humidity) due to 40C drying and smaller shrinkage
distortion due to 11% drying (Figure 2)<Heating> Structural changes due to drying and heating• Up to 100C: Dissipation of free water inside gaps ⇒ Structural
shrinkage, coarsening of gap size (Inference)• 100C onwards: Continuous dehydration from gel formation ⇒Structural shrinkage, coarsening of gap size
The strength deterioration at high temperatures is less ascompared to cement solidifications. (Figure 3)
There are reports that high temperature (1000 C onwards) heat
resistance is low, when the Ca quantity is large.
<Dryness (Humidity) > Mainly structural shrinkage due todrying (Inference)
There areonly a few research
examples.
There is no unified opinion
about the trends.

Heating
temperature

Fig. 1 Relation between temperatureand strength in concrete (Up to 800C)
1)

Compressive
strength

Compressive
strength prior
to heating

Shrinkage distortiondue to drying

Cement solidifications (OPC)

Compressive strength survival rate➢

【Confidential】

CementFig. 2 Relation between concrete
drying conditions and shrinkage
distortion due to drying2)

Slag system AAM

Fig. 3 Relation between temperature andstrength in AAM (slag systems) 3)

It is necessary to collect data through testing for the heating/ drying effect of AAM (Compressive strength, shrinkage distortion)
Reference literature
1) Japan Society of Civil Engineers: Value Technology Research Subcommittee Report for Concrete Structures, Symposium thesis collection and Concrete technology series 63, pp. 32-38, FY2004
2) Itou et al: Study for tensile strengthening for cracking in concrete due to drying following temperature and humidity changes, Annual concrete engineering thesis collection, Vol. 36, pp. 382-387, FY2014
3) Ichinomiya et al: Composition of geo polymer with a fly-ash base and high temperature resistance, Annual concrete engineering thesis collection, Vol. 36, pp. 2230-2235, FY2014©International Research Institute for Nuclear Decommissioning

115

[Study Results]- Thermal load and drying load conditions set based on the investigation -【Confidential】➢ Based on the investigation results, the thermal and drying load conditions are set as follows for thetesting of solidifications.
(Thermal load conditions)••20C (RH60%) (Standard)
80C (Temperature often set during advance research
for the thermal load test)
105C (Dissipation of almost all the free water present
inside the gaps (Conditions similar to the conditions
for measuring quantity of free water))
200C (Temperature zone wherein the dissipation had
settled down during the TG analysis of the AAM
solidifications) (Figure 1)

Temperature

•
•
•
•

RH 100% (20C) (When there is no drying)
RH 80% (20C)
RH 60% (20C) (Standard) # Same as thestandard for thermal load
RH 11% (20C) (Cement solidifications:
Dissipation of almost all the free water present
inside the gaps)
Cement solidifications

Shrinkage distortiondue to drying••(Drying load conditions (Figure 2)

Temperature

Fig. 1 DTA curve (FY2018 data)

Fig. 2 Relation between drying
conditions and shrinkage distortion due
to drying

Presently, the thermal load test has been started by manufacturing solidifications of only the base material.
©International Research Institute for Nuclear Decommissioning116

【Confidential】[Study Results]- Evaluation testing of the base material: Test Overview -Table 1 Composition conditions of the test specimen for the thermal and drying load tests

Types

M/ Basematerial fineparticles[wt%]

BFS/ Basematerial fineparticles[wt%]

Si/ Al molar
ratio in basematerial[mol/mol]Wastefilling rate[wt%]

Water/solidificati
ons ratio[wt%]

Na/Si molar ratio
in base material[mol/mol]

OPC

-

-

-

0

45

-

M

100

0

1. 8

0

150

0. 84

MB20

80

20

1. 8

0

110

0. 50

MB40

60

40

1. 8

0

100

0. 63

Kneading and pouring

CuringDrying and thermal load

20℃
Sealed curing
over a 28 day
periodFor strength test Φ5 x 10 cm
A 4x4x16 cm specimen has been
made for measuring theshrinkage distortion

Drying(Constant temperature andconstant humidity desiccator)⚫ Changes inshrinkage
distortion withthe passage oftime⚫ Changes incompressive
strength with thepassage of time
⚫ Changes in theamount of freewater, amount ofbonding water
Heating
and XRD(Electrical drying furnace)analysis with thepassage of time

©International Research Institute for Nuclear Decommissioning

117

[Study Results]【Confidential】- Evaluation testing of the base material: Changes in compressive strength due to heating and drying -0 day Load7th day Load
28 day Load

Compressive strength

There was no
decrease instrength. Rather
strength tends toincrease in caseof a thermal load(Due toacceleration ofreaction).

Compressive strength

0 day Load7th day Load
28 day Load

Strengthdecreased evenunder weakdrying conditionsof RH80% andremained almost
constant from
RH60%.

0 day Load7th day Load
28 day Load

AAM: MB20
Decreased fromRH60%
but remained
almost constantfrom 80C.

AAM: M

Compressive strength

OPC

AAM: MB40

Compressive strength

0 day Load7th day Load
28 day Load

Trend similar to
M. Although
there wasstrengthdeteriorationeven at RH100%,it remained
almost constantfrom 80C.

Fig. 1 Compressive strength changes in the various solidifications due to thermal and drying load➢➢

OPC solidifications: There was no significant strength deterioration even when thermal and drying load was applied.
AAM solidifications: There was strength deterioration even under weak drying conditions although the extent varied (thechanges settled down after 7 days of load)©International Research Institute for Nuclear Decommissioning118

【Confidential】[Study Results]- Evaluation testing of the base material: Changes in mass due to heating and drying -Compressive strength survival rate

Relation between rate of mass changes and compressive strength survival rate under various
thermal and drying load conditions with 28 days of load

Temp ℃

Humidity %

20

100

20

80

20

60

20

11

80

(-No control)

105(-No control)

200(-No control)

The massdecrease ratetended to be large.

Mass change rate
Fig. 1 Relation between mass change rate and compressive strength survival rate (28 days of thermal and drying load)•

The strength survival rate decreases as the mass change rate becomes larger. (However, in case of MB40 at
RH100%, there was a deterioration in strength although there was no decrease in mass.)⇒ Strength deterioration due to drying⇒ Possibility of strength deterioration due to the generation of cracks following shrinkage distortions due todrying©International Research Institute for Nuclear Decommissioning

119

【Confidential】[Study Results]

Shrinkage distortion (μ)

Shrinkage distortion (μ)

Shrinkage distortion (μ)- Evaluation testing of the base material: Long-term changes due to heating and drying (shrinkage
distortion) -No. of days (d) with thermal and drying load

RH 100%The shrinkage distortion of MB40
is large.

No. of days (d) with thermal and drying load

RH 60%The shrinkage distortion of MB40
alone is significantly large.

No. of days (d) with thermal and drying load

80C
Shrinkage distortion:
M>MB20≒MB40>>OPC➢ On the whole, the shrinkage distortion is larger in case of the AAM solidifications as compared to the
OPC solidifications.
➢ The main cause of the deterioration in strength of AAM solidifications is inferred to be shrinkage
distortion, since there is a correlation between the changes in compressive strength and the magnitudeof the shrinkage distortion of the AAM solidifications owing to heating and drying.
➢ Since the extent of moisture dissipation and shrinkage distortion due to drying will change due to various
factors pertaining to the solidifications (amount of moisture, gap structure etc.), there is a possibility thatthe trend may be different in simulated waste solidifications.
©International Research Institute for Nuclear Decommissioning120

(a) Establishment of techniques for selecting the preceding processing method
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications(i) Investigation of change in properties of solidifications due to heating[Summary/ Plan for the following year]【Confidential】➢ Current Summary•

The findings about the possibility of changes in cement and AAM due to heating and drying
were consolidated based on the past literature surveys.

•

Specimens were prepared and data related to compressive strength and amount of distortion
under the heating and drying conditions was gathered, and it was verified that on the whole,shrinkage distortion was larger in case of AAM solidifications as compared to OPC
solidifications.

➢ Plan for the following year•

The effect of heat and drying load on the special cement solidification will be verified.

©International Research Institute for Nuclear Decommissioning121

(a) Establishment of techniques for selecting the preceding processing method【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications (ii) Evaluation of the relation between
the Cs inventory and the temperature of solidifications[Implementation Details]

The limiting values for the inventory that can be solidified,as studied from the decay heat, will be acquired as data byclarifying the changes due to shape and size, by conducting
an analysis of the maximum temperatures.
FY

Implementation Details

2019

Investigation of the thermal effects due to the shape andinventory of the OPC solidifications
・Selection of the system (cylindrical/rectangular etc. )・System model creation and setting the analytical conditions
・Cs thermal effect analysis and thermal effect analysis of
nuclides other than Cs

2020

Investigation of thermal effects of newly solidified
materials
・Thermal effect analysis of AAM solidifications

Temperature (℃)•

Upper limits were decided for each type of material,
since the performance of the solidifications is
influenced by temperatures under the low-temperaturesolidification treatment technology (cement and AAM).
It is expected that the limiting values for the waste filling
amount will change according to the shape and size ofcontainers, since heat transfer depends on the system.

Distance (cm)Fig. 1 Simulation result by PHITS and COMSOL (Example)

Maximum
temperatures•

Amount of radioactivity

Fig. 2 Nuclide inventory and maximum temperature ofsolidifications (Example: 200L drum)©International Research Institute for Nuclear Decommissioning122

【Confidential】[Study Results]- Selection of containers for analysis -⚫ The containers to store the solidified waste have not been determined.
⚫ The shapes and sizes of the typical waste containers that are in use at Fukushima Daiichi and othernuclear power facilities as well as the containers under study, etc. are being investigated.
⚫ There are broadly 2 types of containers namely, cylindrical and rectangular, hence 3 containers from eachtype have been selected for analysis (Hatching of Table 1, 2).
Table 1 Main cylindrical containers
CylindricalTankcontainer
タンクコンテナ

Pail
ペール缶

Canister
キャニスター

Drum
can
ドラム缶

内容量(m3)
capacity (m3)Internal

0.022

0.15

0.22

1.25

2.69

24.00

Filling capacity
充填量(m3)
(m3)

0.018

0.12

0.17

1.00

2.15

19.20

0.84

0.42

0.65

円柱容器
containers

MJ PIC

HIC(type2)

容器外観
Container
appearance

Internal
内W/H
W/H
Other
その他

JIS Z 1620

JAEA

JIS Z 1600

0.96
TAIHEIYOCEMENT, etc.
太平洋セメント他

0.88
TEPCO

Rectangular
角型容器
containers

2.46
Uranium system
ウラン系コンテナ
containers
JIS Z 1624

1m

3

Table 2 Main rectangular containers
3 (With inner lining)

1m （内張有り）

6m3

LLW-2
type
LLW-2型

20FT

Container

容器外観
appearance

From
NFT HP
NFT
HPよりInternal capacity
3
内容量(m
)
(m3)

1.20

1.35

0.77

3.02

5.20

32.91

Filling capacity

0.96

1.08

0.62

2.42

4.16

26.33

1.34

1.63

1.00

3.11

1.54

AssumedAssumedmodel (Shield
想定モデル
thickness 70 mm)（遮蔽厚さ70mm）

3
充填量(m
)
(m3)

Internal
内W/H
W/H
Other
その他

JAEA

想定モデル
model

NFT

Assumed
想定モデル
model

2.46
JIS Z 1610

©International Research Institute for Nuclear Decommissioning123

【Confidential】[Study Results]- System model and analytical conditions ➢ Creation of a system model••

A system model was created for use in the analysis ofradiation during transportation and in the heat analysis of
6 containers.
Cylindrical type: Pail, Drum, HIC
Rectangular type: 1m3, 6m3, 1m3 with shieldIn each of the containers, solidifications containing
radioactive nuclides were filled up to 80% of the volume.

Cylindrical container model➢ Analytical conditions•••

Rectangular container
model

Fig. 3 System model used for the analysis of radiation duringtransportation (Example)

Solidification material
Ordinary Portland Cement (OPC)
Density 2.69 g/cm3, thermal conductivity 0.5 W/(mk)Radioactive nuclides
Typical nuclides contained in the secondary waste generated
from contaminated water treatment were assumed.
Sr-90, Co-60, Cs-137
Radioactive concentration
The values where any temperature change was expected
were analyzed.
1.0E+5 to 1. 0E+10 Bq/cm3

Fig. 4 System model used for heat analysis (Example)©International Research Institute for Nuclear Decommissioning

124

[Study Results]- Relation between the container system and maximum temperature -【Confidential】•

Maximum temperatures were analyzed in 3 types of cylindricalcontainers and 3 types of rectangular containers.

•

The trends in maximum temperatures were understood with respect tothe volume of each shape (top image in Figure 6).

•

It was suggested that the maximum temperature can be expressedwith a single relational expression irrespective of shape by organizing
it with the number obtained by dividing the volume of the solidifications
by the distance between the center of the solidifications and the
midpoint (edge) of the upper surface of the solidifications. (Bottom
image in Figure 6)

Maximum
temperature➢ Changes in maximum temperature due todifferences in containers

Cylindricalcontainer
Rectangularcontainer
Volume ofsolidifications

Bq/cm3➢ Investigation concerning the influence of nuclides•

The temperature of the solidifications was investigated by changing the nuclides.

•

It was estimated that there was no rise in temperature in case of Cs-137 and Co60 based on a trial calculation carried out using the analytical values ofradioactive concentrations of the secondary waste generated from contaminatedwater treatment.

•

Temperature

Fig. 5 Distance from the center of the solid till the edge

Volume of solidifications/Distance till
the edge

Fig. 6 Relation between temperature and the container systems

Meanwhile, a temperature rise of about 3C was estimated in case of Sr-90©International Research Institute for Nuclear Decommissioning125

(a) Establishment of techniques for selecting the preceding processing method【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications (ii) Evaluation of the relation between
the Cs inventory and the temperature of solidifications[Summary/ Plan for the following year]➢ Current Summary•

The shapes and sizes of the typical waste containers that are in use at Fukushima Daiichi and atother nuclear power facilities as well as the containers under study etc. were investigated.

•

There are two types of containers, cylindrical and rectangular, so 3 containers from each typewere selected for analysis and their effect on the maximum temperature was analyzed. In asingle container, there is a high possibility that the temperature will not reach such high levels so
as to cause any change in properties, like the strength of the solidifications.

•

It was found that irrespective of the shape of the container, the maximum temperature can bederived by a relational expression of the distance between the center and the edge middle point.

➢ Plan for the following year•

The thermal conductivity of the AAM solidifications prepared using the acquired data will bemeasured. After reflecting it in the calculations, AAM will be calculated and the difference withthe cement solidifications will be surveyed.

•

A trial calculation of the coefficient that can estimate the temperature of the solidifications fromthe analytical values will be performed based on the calculations for the main nuclides. (Sr-90,
Co-60, Cs137).

©International Research Institute for Nuclear Decommissioning126

(a) Establishment of techniques for selecting the preceding processing method【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications (iii) Investigation and evaluation offactors influencing long-term change in properties[Implementation Details]
In order to investigate the long-term property change behaviors of solidifications, an investigation was conducted to
study the effects of passage of time and external factors. An investigation of the thermodynamics data necessary for
analyzing the cement solidifications and the AAM solidifications and the evaluation regarding the adequacy of the data
was conducted with respect to the methods devised by means of the thermodynamics equilibrium calculations, which are
based on the results of long-term property change behavior evaluation of cement.
From the information such as phase changes obtained as a result of evaluation, an attempt was made to selectspecific elements (waste) that affect the solidification process of OPC and AAM and which need to be excluded.
FY

Implementation Details

2019

Collection and organization of the thermodynamics data and the study of applicability of the thermodynamics equilibrium
calculation
Implementation of the stability phase calculation assuming the shift to the crystallizing phase, drying and rise intemperature (up to 60C)・Attempting of presumptions regarding the assessment of changes in the mineral phase that influences long-term
behavior, by means of equilibrium calculation
・Study on the adequacy of the thermodynamics data used for forecasting long-term behavior and the applicability ofthe equilibrium calculation for AAM.

2020

Applicability of the thermodynamics equilibrium calculation technique for evaluating the property change behavior ofsolidifications
Identification of the applicability range for various low-temperature solidifications. Simulation of property change
behavior of solidifications due to contact with underground water. Study of adequacy and applicability of thethermodynamics data by comparing with the experiment results.
・Presentation of the findings regarding the effect of the carbonate slurry and the iron co-precipitate slurry on the longterm property changes in cement solidifications and demonstration of applicability of the thermodynamics equilibrium
calculation technique for the evaluation of property change behavior of solidifications.

©International Research Institute for Nuclear Decommissioning

127

[Study Results]- Study items and implementation status (1/6) -【Confidential】➢ Collection and organization of the thermodynamics data (common for cement and AAM)•••

The database used is 'Thermoddem' created by the French Geological Survey.
The calculation code used is 'The Geochemist's Workbench'.
The thermodynamics data of the N-A-S-H phase is investigated and added to the database to calculate the amorphous
phase of AAM. It is confirmed that the minerals necessary for the calculations pertaining to cement have been included.

➢ Calculation of the AAM system material•

Fig. 1 pH of various natural water bodies and a list of
secondary minerals found in salt water and alkali water lakes(Inoue 2004)

Volume%Study of the thermodynamics stability phase
Based on the findings (Figure 1) on the changes in the properties ofnaturally found Zeolite or natural minerals such as the ones found in thevicinity of salt water or alkali water lakes which resemble AAM in termsof constituents, the mineral phase emerging after the change inproperties was estimated.
⇒ The mineral phase and the N-A-S-H phase presumed from the literature
survey were used for the AAM calculations.
•
Calculation of the mineral phase
Along with implementing the calculations of the AAM solid phase and thecoexisting liquid phase, the calculations were compared with theleaching test results of the metakaolin system AAM acquired in FY2018.
Figure 2 shows the comparison between the calculation result of thesolid phase, the calculation result of the pH of the liquid phase and theexperimental values.
⇒The solid phase was not consistent during the period of the equilibrium
solubility test. In the calculation of the liquid phase, the result
reproducing the tendency (The pH and Na concentrations were roughly
reproduced. The Si concentration differed.) of being close to theexperimental value was obtained.

Liquid-Solid ratio

Calculated
valuesExperimenta
l values

Fig. 2 Changes in volume and changes in pH of the mineralphase due to the changes in the Liquid-Solid ratio©International Research Institute for Nuclear Decommissioning128

【Confidential】[Study Results]- Study items and implementation status (2/6) •Investigation of equivalent mineral phases
It was difficult to reproduce AAM with the N-A-S-H single phase, since
AAM was amorphous in the thermodynamics equilibrium calculation, andhad thereby become a combination of minerals. However, even in thatstate, the tendency of changes in the pH or the amount of components inthe liquid phase could be reproduced partially.
If one can understand the tendency of changes in the liquid phasealone, then there is a possibility that it may contribute to the study ofdisposal, hence the combination of solid phase minerals (mineral phaseequivalent to amorphous AAM) was studied as it can well reproduce thepH, Si/Al ratio, Si quantity, Al quantity, Na quantity, etc. of the liquid
phase, which were obtained through the leaching test.
⇒A comparative evaluation of the calculation results and theexperimental values was carried out. (Table 1)

Table 1 Comparative evaluation of the calculation results and theexperimental values using mineral combinations.
Mineral combination

Solid
phase
Si/Al ratio

Liquid phase

PointsExperiment
al values

Findings concerning the long-term behavior of the AAM
material or the thermodynamics data are very few, but it issuggested that short-term leaching behavior can be
reproduced by setting multiple metastable phases.

Si/Al ratio

There are a few combinations by which the liquid phase composition canbe reproduced well, and also the solid phase Si/Al ratio can be
reproduced to a certain extent.
⇒The pH (Al and Na concentrations) calculated under various
conditions had values (Figure 4) closer to the experimental values in allof the mineral combinations. As for the Si/Al ratio of the solid phase, thecalculated value was not close to the experimental value under thecondition where the Liquid-Solid ratio was high.

Liquid-Solid ratio

Liquid-Solid ratio
ExperimentalvaluesExperimental
valuesExperimentalvaluesFig. 4 Comparison with the equilibrium solubility test results(pH and solid phase Si/Al ratio)©International Research Institute for Nuclear Decommissioning

129

[Study Results]- Study items and implementation status (3/6) -【Confidential】➢ Calculation of cement material• Study of the thermodynamics stability phase
There are results of evaluation of cement by means of the thermodynamics equilibrium calculation. However, basedon experimental findings, manipulations were made to intentionally select the mineral phase.

Not applicable for the analysis of solidifications containing waste without the experimental
findings.

[1]

Calculation results considering all minerals in the DB[1]

Result recalculated after excluding the minerals assumed
to be in the thermodynamics stability phase and the C-SH minerals of the crystals from [1][1]In addition to [1], calculation was carried out after
further excluding the minerals not assumed to begenerated due to the conditions such as temperature andpressure, etc.

[2]

The minerals assumed to be generated in the target
system were selected from the DB, and the mineral phasecomposition was calculated under the conditions which
would generate only the selected minerals.

Percentage ofvolume

The analysis was performed by changing the method of
excluding the mineral phase as follows.

Fig. 4 Volume percentage of generated mineral phase at eachlevel

Although the result (Figure 4) is significantly different due to the method of exclusion, it was found that the results
obtained were similar to the case where the calculations were carried out by selecting minerals assumed to be generated
based on findings, using the method of exclusion by following a certain rule based on the thermodynamics stability phasestudied during this research, ([1] and [2])©International Research Institute for Nuclear Decommissioning130

[Study Results]- Study items and implementation status (4/6) •【Confidential】

The calculations for changing the content rate of the carbonate slurry, calculations for temperature changes
and calculations for leaching due to contact with water, were carried out for the cement solidifications.

(# At the present stage, there are indications from the results obtained from the thermodynamics equilibrium
calculation but a comparative verification with the experimental findings needs to be conducted separately.)➢ Mineral phase composition when contained in the slurry (Figure 5)
Slurry: Carbonate slurry
Temperature: 20C
Content is 0 to 55 [mass%] in increments of 5%[Indications from the calculation results]# The calculation did not converge above 60 [mass%] or more.
◯ Changes in the cement-based minerals due to theaddition of carbonate slurry・ Decrease in portlandite, which is the main mineral in
cementPore water

Mineral phasecomposition

・ Changes in the Aluminate system minerals(KatoiteSi1, Ettringite → Monocarboaluminate,hydrotalcite)
・ Changes in the ferrite-based minerals(C4FH13→C3FH6)
MineralsfromcarbonateslurryCarbonate slurry content→ There are no major volume changes.
◯ Carbonate slurry 45 mass% or more
Increase in the solid phase volume due to a newly
generated phase originating from the carbonate slurry

Fig. 5 Volume changes in the mineral phase of the cement
solidifications following changes in the carbonate slurry content©International Research Institute for Nuclear Decommissioning

131

[Study Results]- Study items and implementation status (5/6) Changes in the mineral phase composition
during changes in temperature

・Waste content rate: Carbonate slurry is 30 [mass%]
・Temperature: 0 to 60 [℃], 100 [℃] (Crystallizing conditions)# The temperature at which calcium silicate hydrate (C-S-H) changes to the crystalline phase isunknown, hence the calculations were performed in the temperature range of 0 to 60 ℃ where
there was no change, and at the temperature of 100 ℃, where the changes are presumed tohave taken place.
[Matters indicated from the calculation results]◯ A temperature rise of up to 60℃
・ Changes in the Ferrite-based minerals (C4FH13→C3FH6)・ No changes in the minerals from the carbonate slurry.
→ No big changes were seen in the solidifications.
◯ Crystallizing conditions at 100℃
・ C-S-H and Portlandite disappear and the crystal mineral is generated.
→ Changes in the mineral phase, decrease in volume in the solid phase.

Mineral phasecomposition➢

Temperature

Pore water

Minerals from thecarbonate slurry

Fig. 6 Volume changes in the mineral phase of the carbonate slurry cement
solidifications following temperature changes

Mineral phase changes in case of leaching
following contact with water
Leached part

Mineral phase
Volume %➢Liquid-Solid ratio(Pure water/Solid phase mass weight)【Confidential】

Minerals fromthecarbonateslurry・ Amount of waste filling: Carbonate slurry is 30 [mass%]
・ Immersion conditions: Pure water・ Solid-Liquid ratio: 1:10, 1:100, 1:500, 1:1000, 1:2000
・During the calculations of the various Liquid-Solid ratios, the decreased portion of
the mass and volume of the solid phase was assumed to be the leached out
portion.
[Matters indicated from the calculation results]○ Solubility alteration in the cement-based minerals
・ Portlandite disappears.
・ The C/S ratio of C-S-H decreases.
・ M-S-H is generated.
○ Changes in the minerals from the carbonate slurry
・ At the Liquid-Solid ratio of 2000, Brucite disappears andCalcite decreases.
・ M-S-H is generated.
→ Decrease in volume in the solid phase due to dissolution.

Fig. 7 Change in mass ratio in the mineral phase followingchanges in the Liquid-Solid ratio

The calculation results for cement were obtained with respect to the essential filling rate, temperaturechanges and the leaching properties required for studying the waste solidifications.
©International Research Institute for Nuclear Decommissioning132

(a) Establishment of techniques for selecting the preceding processing method【Confidential】
 Low-temperature treatment technology
ii. Investigation of change in properties of solidifications (iii) Investigation andevaluation of factors influencing long-term changes in properties[Summary/ Plan for the following year]➢ Current Summary•

The thermodynamics data was collected and maintained, and the thermodynamics stability
phase and the metastable phase were selected from the literature survey.
(AAM)• Calculations were implemented for the AAM solid phase and the coexisting liquid phase, werecompared with the leaching test results of the metakaolin system AAM acquired in FY2018, andthe possibility to reproduce was obtained by setting multiple metastable phases.
(Cement)• As regards the solidifications containing carbonate slurry, analytical values were obtained for themineral phase, mineral phase changes due to changes in temperature, and mineral phasechanges in case of leaching following contact with water.

➢ Plan for the following year
(AAM)
• Additional calculations will be performed when waste elements are added to clarify the applicability ofthe thermodynamics equilibrium calculation at the present point and examine the issues.
(Cement)• The iron coprecipitation slurry will be calculated. Additionally, analysis results of solidifications
containing slurry and the calculated results for elution, heating and drying test which are being
collected under other implementation items will be compared to clarify the applicability of thethermodynamics equilibrium calculation at the present and to examine the issues.

©International Research Institute for Nuclear Decommissioning133

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies[Goals]i. Investigation and study regarding the influence of waste composition on the performance of solidifications
To study the effect of the composition and chemical form of solid waste and the impact of the composition ofsolidifications on the performance of the solidifications, in order to evaluate the range of solid wastes to which
various treatment technologies may be applicable.
ii. Data acquisition for equipment configurations for various treatment technologies
To study the concept of solidification treatment equipment, collect data on economic efficiency such as equipment
configuration and its treatment efficiency, maintenance details, types of consumables and replacement frequency
and types and weight of secondary waste etc. and thereby make additions and updates in the Technology
Investigation Table, in order to evaluate the waste specifications of various treatment technologies.
iii. Investigation regarding the volatility of Cs during high temperature treatment and its control
To study the volatility of Cs and its control mechanism and evaluate the volatile properties of Cs by means oftreatment methods and operating conditions, since the volatility of Cs is a problem in the high temperature treatment
technology.
Technical Information on the melting technology A
MeltingAAMsolidifica
tion
Cementsolidification

Evaluation axis A

Evaluation axis A

Evaluation axis B•Property value of waste A

Melting technology

Fig. 1 Image showing the applicability evaluation method to be developed (Evaluation approach)©International Research Institute for Nuclear Decommissioning134

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies
i. Investigation and study of the effect of waste composition on the performance of solidifications[Implementation Details]
In order to evaluate the range of solid wastes to which the various treatment technologies may be applicable,
the effect of the composition and chemical form of the solid waste and the composition of solidifications on
the performance of the solidifications was studied. The scope of the high-temperature treatments was studied
using a database and the scope of the low-temperature treatments was investigated through the results oftests conducted under other items and through literature surveys.
In FY2019, the scope of applicability was studied using databases or literature, and in FY2020, theinsufficient data will be collected experimentally and the scope of applicability will also be studied.

FY

Implementation Details

2019

Evaluation of the filling concentration (composition range) with respect to the various secondary waste generated fromcontaminated water treatment
・Collection of data on vitrification testing (including the supplementary activities) of the secondary waste generated fromcontaminated water treatment through literature survey, and organization of glass properties for various filling concentrations
or compositions
・Incorporation of data in the glass properties model, and evaluation of the maximum filling concentration of the main
secondary waste generated from contaminated water treatment in the range within which the glass properties such as themelting point, etc. are ensured

2020

Evaluation of filling concentration (composition range) when multiple wastes are mixed
・Reflection of the results obtained from the low-temperature system experiments and the results of the literature survey, andorganization and evaluation of the limiting values for the range of waste properties and the filling concentration (possible
mixing range) for the slurry
・Use of the glass properties model to evaluate maximum filling concentration of the main secondary waste generated fromcontaminated water treatment in the scope within which glass properties such as the melting point etc. are ensured. And
verification of the properties by means of the glass melting test, if required (FY2019-2020)©International Research Institute for Nuclear Decommissioning135

【Confidential】[Study Results]- Study methods➢

Collection of glass data from public literature together with the results of glass solidification tests of thesimulated waste from the Fukushima Daiichi NPS, which was measured under the Government-led
R&D Program on Decommissioning and Contaminated Water Management➢

Setting up of restricting conditions (such as the melting point, viscosity and presence of precipitation
phase etc.), and evaluation of the highest concentrations of the secondary waste generated fromcontaminated water treatment under these conditions by extrapolating the glass database• Analysis based on INTERGLAD ver. 8 of the New Glass Forum (NGF) and some proposers’ data

➢ Evaluation of the secondary waste generated from contaminated water treatment as consisting mainlyof Carbonate slurries (Ca, Mg, Na), the Zeolites (Si, Al, Na) and the Titanic acids (Si, Ti), which arepresent in large quantities
Carbonates

Recommended composition range
Experimental values (literature etc.)of the glass properties

Titanic acids

Zeolites (KUR-H, IE-96 etc.)

Fig. 2 Illustration of the Evaluation Results©International Research Institute for Nuclear Decommissioning136

【Confidential】[Study Results]- Analysis with the Glass Properties Model -➢ A total of 18,209 units of physical property data on glass containing more than 50% Zeolite could
be extracted from INTERGLAD (Fig. 3), and it was possible to derive an estimation formula for
liquid phase temperature, viscosity etc. by means of multiple regression analysis (Fig. 4).
Meanwhile, since there was insufficient data about the glass containing carbonate slurry or titanicacid, it was difficult to derive an estimation formula for the same.
➢ On comparison with the PNNL estimation formula (which is in the public domain), the viscosities
were almost the same below 103 dPa/s, but higher above that value (Fig. 5). The concentration
range will be evaluated next year when the accuracy of the estimation formula will be improvedby outsourcing PNNL.

Measurement values

Fig. 3 Liquid phase temperature data
(The black spots confirm the melting point
at 1250℃)

Viscosity

Viscosity/ dPa・s (PNNL)

Liquid phase
temperature

Liquid phase temperature LT/℃
(InterGlad)

Liquid phase
temperature

Estimation
formulaFig. 4 Multiple regression results and estimation formula
examples of the liquid phase temperature data (‘x’ shows
the quantity of the contained element i while ‘a’ shows thecoefficient for the element i.)

Viscosity/dPa/s (InterGlad)

Fig. 5 Comparison of evaluation values with different
viscosity estimation formulas (Estimation formula by
INTERGLAD and publicly available formula of PNNL)©International Research Institute for Nuclear Decommissioning137

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies
i. Investigation and study of the effect of waste composition on the performance of solidifications[Summary/ Plan for the following year]➢ Current Summary•

Based on public data of INTERGLAD ver. 8 of the New Glass Forum (NGF) and the data
proposed by related parties, analysis was carried out. The analysis showed that it waspossible to derive an estimation formula through multiple regression analysis for the liquid
phase temperature and viscosity of glass with Zeolite content.

➢ Plan for the following year•

Public data will be added to PNNL-specific database as well as the data obtained fromcompanies involved in the project, which enables estimation of a glass physical properties.
By using the estimation formula, the concentration range of the secondary waste generated
from contaminated water treatment including Zeolites (chabazite), Silica and Titanic acids(CST) and Carbonate slurry, which can melt in glass, will be evaluated.

©International Research Institute for Nuclear Decommissioning138

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies
ii. Collection of data on equipment configuration for the various treatment technologies[Implementation Details] (1/2)
Economic data such as equipment configuration and its treatment efficiency, maintenance details, typesof consumables and replacement frequency and types and weight of secondary waste etc. wasinvestigated and organized in order to evaluate the waste specifications of various treatment
technologies.
In case of some of the technologies, the lacking economic efficiency data was acquired by studying theconcept of solidification treatment equipment, and this data was organized as technical information forvarious technologies.
The lacking economic efficiency data was collected in FY2019, and in FY2020, the investigation results
including all the data obtained till date will be organized.

FY

Implementation Details

2019

Data acquisition on equipment configuration for each of the treatment technologies
・Investigation of equipment configuration and its treatment efficiency, maintenance details, types ofconsumables and replacement frequency, secondary wastes etc., as part of the applicability evaluationof various treatment technologies, and study of the target technologies.
・Further study in case of some of the technologies such as the study of the process flow or material
handling etc., and organization of its results as technical information for the various technologies.

2020

Consolidation of the investigation results for each technology
・Consolidation of investigation results that will include all the data that has been collected till date.

©International Research Institute for Nuclear Decommissioning139

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies
ii. Collection of data on equipment configuration for the various treatment technologies- Technologies to be investigated Types of investigations (Major classification)
Classification

Solidification treatment technology methods
Technical Investigation (Investigation about theProcess flow, Equipment configuration etc.)

Investigation based on economic efficiency data
(Study of equipment scale etc. depending on theradiation dose)

Joule Heating (LFCM)

FY2019 (There are no examples for a low radiation dose
and therefore, a study was conducted in FY2019 tosee whether an investigation is possible.)

Induction Heating (AVM, AVH)

(It is determined that an alternative evaluation
between Joule Heating and Induction Heating ispossible)(It is determined that an alternative evaluation
between Joule Heating and Induction Heating ispossible)

Induction Heating (CCIM)[Partial project][Partial project]

Joule Heating (ICV)FY2018[Partial project]

External Heating (Dem & Melt)[Partial project][Partial project]

Induction Heating(It is determined that an alternative evaluation
between Incan and Induction Heating is possible)(It is determined that an alternative evaluation
between Incan and Induction Heating is possible)

Plasma Heating

FY2019 (There are no examples for a high radiation dose
and therefore, a study in FY2019 to see whether or
not an investigation is possible.)

Induction Heating

FY2018 (There are no examples for a high radiation dose
and therefore, a study in FY2019 to see whether or
not an investigation is possible.)

Out drum type

FY2018FY2019In drum type

FY2018FY2019Flow type

Vitrification

Hightemperature
treatment

Incan type

Flow type
Meltingsolidification

Incan type

Cementsolidification
Lowtemperature
treatment

Out drum type
AAM
solidification
In drum type

(Not studied from point of view of engineering. An
alternative evaluation is deemed possible as there are(An alternative evaluation of the adaptability of thecompositions similar to cement.)
equipment to the differences in radiation dose isconsidered possible with cement solidification)
FY2018➢ Technical investigation of methods (Joule Heating method and Plasma Heating method) from amongst thesolidification treatment technology methods that were not investigated➢ Study of the cement solidification technology concerning equipment addition in line with the radiation dose
(confinement or shielding) or the size of the equipment, as an investigation based on the data on
economic efficiency©International Research Institute for Nuclear Decommissioning140

【Confidential】[Study Results]- Investigation of applicability to the secondary waste generated from contaminated water treatment•
The properties of the secondary waste generated from contaminated water treatment were enumerated, andapplicability of the high-temperature solidification treatment technology was investigated.

Waste

Limiting values

Problematic
waste
properties
Melting point
Viscosity

All typesof wasteCs concentrations

LFCM
Less than 1200C

Less than 1650C

10Pa/s order or less at the treatment temperature #1

Approx. 1 Pa/s order or less at the treatment temperature #1

Must be less than equivalent viscosity even for hot water.

Must be less than equivalent viscosity even for hot water.

There is migration to off gas regardless of the concentrations(Control measures with a cold cap can be executed).

Ferrocyan
ide sludge

Titanate

There is a shift to off gas regardless of the concentrations.

Moisture content

Can be supplied in liquid or slurry state

5 wt% or less is the standard

BaSO4

Since the leaching tolerance of the solidifications decreases due
to formation of the sulfate layer, it is necessary to curb thequantity of the waste mixture.

Due to the drop in melting point, the quantity of waste mixture is curbed. The
quantity of waste input is about 50wt% for SiO2. There are no restrictions fromthe viewpoint of the formation of the sulfate layer, since it is expected to bevolatile as SOx.

CyanideCarbonate slurry

Plasma Heating

CN decomposes or oxidizes and does not remain inside thesolidifications.

CN decomposes or oxidizes and does not remain inside the solidifications.

CaCO3

It is necessary to add the melting point lowering agent .

Mg(OH) 2

It is necessary to add the melting point lowering agent.

Na2CO3

Although there is no issue since Na lowers the melting point, depending on the chemical form of Na, there is a rise in the volatility of Na or Cs.

Boric acid

Although there is no issue since B lowers the melting point, depending on the chemical form, there is a rise in the volatility of B or Cs.

Titanate

When the melting point exceeds the treatment temperature, the melting point lowering agent is necessary.

It is necessary to add the melting point lowering agent. (For instance, waste atweight ratio : SiO2 = 66:34)#1 It is based on the investigation results. However, although it differs by an order of magnitude, it does not mean that a lower viscosity is required in principle in the case ofplasma heating.

©International Research Institute for Nuclear Decommissioning141

【Confidential】[Study Results]- Collection of data on economic efficiency (Study results concerning layout) -•Items such as confinement, shielding etc. or the scale of facility was investigated considering theradiation dose of the secondary waste generated from contaminated water treatment.

Waste storage
container route

Waste carrying in area

Shielding or Remote operations
area

Curing area

Controlled area

Kneading area

Non-controlled area

Transfer tothe upper
level with EV

Raw material carrying-in area such as for
cement etc.

Fig. 1 Equipment layout plan for remote in-drum type cement solidification treatment©International Research Institute for Nuclear Decommissioning142

【Confidential】[Study Results]- Collection of data on economic efficiency and comparison of practical engineering application -Comparison and study of practical engineering usage of the various technologies following from FY2018
Technology investigated this year

LFCM

Cement solidification (In drum type)(High dose #1-Remote operations)Plasma heating

Target secondary waste generated fromcontaminated water treatment → Solid

Zeolite＋Carbonate slurry

Carbonate slurry

Carbonate slurry

products→ Glass solidifications→ Molten solidifications (slag)→ Cement solidifications

Stage of practical application on an actual scale within Japan andoverseas

Stage of practical application on an actual scale within Japan andTechnical achievements

Many waste (LLW) treatment results available within Japanand overseas

overseas(HLW, LAW)
(LLW)

Process

Treatment Temperature

Operability

Room temperature1100 to 1200℃

Up to 1600℃

Approx. Up to 0.3 t/h #2

Approx. 0.2 to 0.3 t/h #3

Cs: Unknown #4

Cs: As an example it is 50%

Cold cap formation, Treatment temperature, Hot water output

Composition adjustment, Treatment temperature, Hot water output

control

controlFormation of non-homogeneous solidifications, blockage of hot water,
container damage, uncontrollable cold cap

Leakage of cooling water, container damage, inadequate hot water

0.24 t/h (12 batches/ series) - 2 systems
Treatment Speed
Cs Volatilization rate

Main items of Evaluation

None

High influence parameters

Process Risk

Main equipment configuration

Moisture content of waste, Amount of sodium carbonate

Power supply equipment, off-gas system, shielding equipment

Back-up (water addition) difficult when fluidity is poor

supplyDrying equipment, power supply equipment, torches, off-gas
systems, shielding equipment

Material supply equipment, agitator equipment, dustproof/
exhaust equipment, shielding equipment (remote control
system), forklift with shielding

HEPA filter, Filter residue, Refractory re-covering,

HEPA filter, Filter residue, Refractory re-covering,Electrodes

Electrodes

Difficulties in remote operation

Refractory replacement in the furnace

Refractory replacement in the furnace

Process in the shielded area

Waste supply/mixing, Vitrification/hot water

Secondary waste generation

Economic

•

None

None

Unknown due to low dose performance only, estimated to be similar Waste crushing, Weighing, Kneading, Transfer of

Shielding/Controlled Area/Total
Building Area (m2)

Estimated maturity for on-site
application

Unknown

Demonstration test at a full-scale level with simulated waste isrequired.

to LFCMsolidifications, Lid closure, Curing

0/Unknown/1450 (2F underground, 2F above ground, 4800 in total)

700/1400/2100

Demonstration test and active test at the full- scale level withsimulated waste are required after optimizing the solidification
treatment conditions.

Demonstration tests and active tests at a full-scale levelwith simulated waste are required.

#1 Approximately 1/100 of the maximum inventory measurement value of each nuclide in the currently obtained carbonate slurry was set as the threshold for high dose and low dose.
# 2 The glass production rate was 1 t / h, the waste filling rate in the solidifications was estimated to be 0.2 and the moisture content of the pretreatment waste was estimated to be 50 wt%.
# 3 As per the investigation results. Approximately 0.2 to 0.5 t / h when estimated with PACT-8 of Japan Atomic Power Company or with the Plasma melting furnace of JAEA. The composition ratio for treatment is Carbonate slurry : Glass material = 5: 5.
Density of solidifications 2500 kg/m3, pretreatment waste moisture content 50wt% and the volatile matter in the solidifications in the carbonate slurry are considered.
# 4 The plan is to investigate the actual values in the United States next year.

©International Research Institute for Nuclear Decommissioning143

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologies
ii. Collection of data on equipment configuration for the various treatment technologies[Summary/ Plan for the following year]➢ Current Summary•

A technological investigation was carried out with respect to the high-temperature treatment
techniques namely, the LFCM and Plasma melting.

•Items related to economic efficiency such as confinement, shielding etc. or the scale of facilities
was investigated considering the dose of the secondary waste generated from contaminatedwater treatment with respect to the cement solidification (In-drum type), which is treated under
low-temperature.

➢ Plan for the following year•

The data investigated so far will be organized and evaluation items that can be used as
evaluation axis to compare technologies will be compiled. Additionally, the scope of
application of each technology will be shown in figures.

•

Any data required additionally during summarizing the data will be collected.

©International Research Institute for Nuclear Decommissioning144

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologiesiii. Investigation of Cs volatilization and control during high-temperature treatment[Implementation Details]
The volatilizing of Cs is a problem in the high-temperature treatment technology and affects the model orthe equipment configuration strongly.
Therefore, the phenomenon of Cs volatilization and its control mechanism has to be evaluated on alaboratory scale, and the volatile properties of Cs have to be evaluated based on the treatment methods
or the operating conditions. In FY2019, information regarding the Cs volatility was collected and a study
was conducted regarding the control of Cs volatility, and in FY2020, data will be acquired byexperimenting on small amount of samples.
The mechanism for implementing measures to control the volatilization rate of Cs by means oftemperature control or additional loading of waste used in various facilities, was investigated andconfirmed to evaluate the Cs volatilization rate and present effective control measures.
FY

Implementation Details

2019

Organization of findings related to volatility control of Cs during high temperature treatment of waste
・Organization of the concerned literature surveys and the engineering test results for each candidate
technology.

2020

Measurement of the efficacy of Cs volatilization control during high temperature treatment of thesecondary waste generated from contaminated water treatment
・ Evaluation of the Cs volatilization control effect with KURION-H, IE-96, silicate titanium acid etc.
(FY2019-2020)
・Evaluation of the Cs volatilization control effect in a sample containing a glass component thatsimulates vitrification (FY2019-2020)・Comparison of the engineering test results of each organized model with the volatilization control
effect under the obtained ideal conditions. (FY2019-2020)©International Research Institute for Nuclear Decommissioning145

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】
 Study on the approach for evaluating the applicability of treatment technologiesiii. Investigation of Cs volatilization and control during high-temperature treatment- Outline of the investigative tests (1/2) •Use TG-DTA, to measure the extent to which the control measures are able to bring down the volatility of CS from thesecondary waste generated from contaminated water treatment.
i) Cold Cap (low temperature)
700℃
simulation
900℃
weight
1100℃
1200℃

Top Off Frit
EffectDried secondary waste generated fromcontaminated waterii) Simulation of Top Off Frit

weightΔW
Cold Cap Effect

Time
700℃
900℃
1100℃
1200℃

1/T

Top Off FritTime
Top Off Frit Effect: The effect of controlling volatilization by lowering the melting temperature by adding a low melting point vitrifying agent to the molten surface
Cold Cap Effect: The effect of decreasing the evaporation area to control volatilization by floating a solid cold cap at low temperature on the upper part of themolten substance➢

Test flow
Assessing volatile behavior
with a waste sample only

Assessing volatile behavior
when a vitrifying agent has
been added to the waste

Verifying the volatility
control effect in the case ofwaste + vitrifying agent +Cold Cap©International Research Institute for Nuclear Decommissioning146

【Confidential】[Study Results]- Consolidation of findings concerning Cs volatility control during high-temperature treatment ofwaste -➢ Literature and actual results investigation••

It has been reported that the Cs volatility at the time of vitrification of Zeolite changed significantly depending on themelting temperature or the glass additives (B2O3 etc.). There are almost no reports on other adsorbents for Cs
volatilization.
The volatility rate of Cs differs depending on the model, but it is difficult to judge, since it depends not just on the device
conditions such as air tightness or temperature control inside the reactor, stirring method (bubbling) etc. but also on theoperating conditions such as the melting temperature or the cold cap formation, mixing ratios of waste, selection of
vitrifying material etc., which continue to vary.

Model

Operatingtemperature/
heating method

Liquid Fed Ceramic Melter(LFCM)

1100 to 1200℃/Direct energized joule heating

Rotary kiln Induction heating(AVM, AVH)
1150℃

1000 to 1200℃/High frequency induction heating/High frequency induction heating(metallic container)(glass)・Real-time results with adsorbent
Volatility data of Cs
infused high-reslurry at West Valley,

・Real-time operation with ORANO-UP3

SRS and Hanford. Real-time results

・There is a report that Cs volatility is a

etc. during
engineering scalewith the high-level nitric acid liquid

problem but no clarity about any

waste at WAK, JNFL.

specific amount of volatility.

testing/ Full-scaletesting

Cold Crucible Induction Melting(CCIM)

・Results for Cs volatility are uncertain.

Joule heating(GeoMelt)

External heating type
(Dem&Melt)

1250℃

900 to 1100℃/Direct energized joule heating/External heater

・The off gas migration rate is・The off gas migration rate is・The off gas migration rate is

Cs = several % or more during theCs = 1 to 9 % during the engineering

Cs = Less than 1% during theengineering scale testing at IHI (Therescale testing at KURION (There are no engineering scale testing at ORANO

are no details for actual testing by other details for actual testing by other(There are no details for actual testing

subsidy projects.)

by other subsidy projects.)

subsidy projects.)
・Due to melting from the lower side and・Formation of Cold Cap
Method of・Formation of Cold Cap・A low temperature scorch layer is・Decrease in the melting temperature

formed on the molten surface due tobecause of optimization of the glass

controlling volatility

by maintaining low temperature on theupper side, volatility is controlled.
・Until the final stage where the waste

・Comparatively low temperature・Temperature control easily controls

has melted till the upper-most side,
the supply of liquid waste.

composition.
・Bubbling is minimized.

the Cs volatility.
volatility is controlled by infusing a lowmelting point glass material called Top
Off Frit.

➡ Data on change in volatility properties depending on operating conditions is necessary in order to interpret the differences in
the Cs volatility rate.
©International Research Institute for Nuclear Decommissioning

147

[Study Results]- Verification of the effect of volatility control ➢【Confidential】

Measurement of the amount of Cs volatility during high-temperature treatment○Clarification of the amount of volatility in different
temperature zones
・Zeolite that has adsorbed Cs is heated with the vitrifying
agent and the weight decrease in each temperature zone ismeasured.

○Changes in the amount of volatility due to the addition of avitrifying agent
・The weight decrease is measured following heating after changing
the vitrifying agent added to Zeolite that has adsorbed Cs.

Forecasted decreased
values due todehydration

Forecasted
decreasedvalues due todecarburization

Assumed vitrified
body weight
Weight decrease due tothe volatility of Cs etc.

(Reagent base)(Without reagent base B)(Glass base)Initial sample weightFig. 2 TG analysis results for each method of addition
Fig. 1 TG/DTA result of the sample added with reagent base
By measuring the amount of weight decrease due to volatility in the different
temperature zones, an ideal volatility decreasing effect resulting from differences in
the melting temperature could be clarified (Cold Cap).

Due to differences in composition of the vitrifying agent (presence of B) and the formof addition (reagent powder, vitrified powder), there are changes in the amount ofvolatility, and data suggesting that the variations are significant at the temperature of1000C or more, is being collected.

⇒Continue collecting data in order to understand the tendency of changes in the rate of volatility due to variations in theoperating conditions. Study whether the variations in the results of the engineering scale testing can be explained by thevariations in the operating conditions, and reflect the results in the comparison of the rate of volatility of Cs based on the model.
©International Research Institute for Nuclear Decommissioning148

【Confidential】(a) Establishment of techniques for selecting the preceding processing method
 Study on the approach for evaluating the applicability of treatment technologiesiii. Investigation of Cs volatilization and control during high-temperature treatment[Summary/ Plan for the following year]➢ Current Summary•

Findings concerning the control of volatility of Cs during high-temperature treatment ofwaste were organized.

•

Measuring test of phenomenal Cs volatility on a laboratory scale was started.

➢ Plan for the following year•

Measuring test of phenomenal Cs volatility on a laboratory scale will continue to collect data.

•

Differences in engineering test results will be studied whether the operating conditions affect theengineering test results. The study results will be reflected in the comparison of Cs volatility ratethat can be an evaluation axis.

©International Research Institute for Nuclear Decommissioning149

(a) Establishment of techniques for selecting the preceding processing method 【Confidential】- Summary of the Progress Status Implementation Items① Lowtemperature
treatment
technology② Study on theapproach forevaluating
applicability oftreatment
technologies

Outlinei. Collectionand evaluationof data on lowtemperature
treatment
technology tocontribute totheidentification oftechnology(i) Study on the verification method for lowtemperature treatment solidification⚫ Selection of simulated waste and study of evaluation methods has ended.
⚫ The screening flow concept has been created.

(ii) Collection of data on properties of cement andAAM solidifications for the slurry⚫ Simulated carbonate slurry has been manufactured.
⚫ Composition conditions for the simulated solidifications have been selected.
⚫ Simulated solidifications have been manufactured and the various analyses
continue on an ongoing basis.

(iii) Investigation of special cements⚫ Literature survey and selection of cement for testing has ended.
⚫ Performance testing of the target special cements base material has ended

ii. Investigationof change inproperties ofsolidifications(i) Investigation of change in properties ofsolidifications due to heating⚫ Literature survey regarding the change in properties of the solidifications dueto heating etc., has ended.
⚫ The heat addition test has been executed (part of it is still underway).

(ii) Evaluation of the relation between the Cs
inventory and the temperature of solidifications⚫ Modeling and simulation calculation (OPC) of the container has ended.
⚫ Trial pertaining to the relation between inventory and temperature ofsolidifications(iii) Investigation and evaluation of factors
influencing long-term changes in properties⚫ Comparison between the thermodynamics equilibrium calculation for the AAM
solidifications with the experimental results⚫ Trial run of the thermodynamics equilibrium calculation pertaining to theproperty change behavior of the carbonate slurry cement solidifications isconducted.

i. Investigation and study of the effect of waste composition on theperformance of solidifications⚫ Collection and organization of the vitrification test data has ended.
⚫ The calculation to evaluate the physical properties using the glass
characteristic model (InterGlad) has ended.

ii. Collection of data on equipment configuration for the various
treatment technologies⚫ Investigation concerning plasma melting, LFCM and cement solidification has
ended.
⚫ Addition into and update of the Technology Investigation Table is carried out.

⚫ Organization of literature survey and engineering test results concerning Cs
volatility has ended.
⚫ Testing to measure the amount of Cs volatility during high-temperature i (ii) Although there has been some delay in the collection of data pertaining to thetreatment
propertiesof thetocement
AAM
solidifications
withcontinues
be carriedandout onan ongoingbasis.

iii. Investigation of Cs volatilization and control during hightemperature treatment➢

respect to slurry, it is mostly progressing as scheduled.
©International Research Institute for Nuclear Decommissioning150

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques◼ Results
 Important case examples were identified by conducting anoverseas disposal site survey. In addition, the applicability to thewaste disposal at 1F was evaluated.
 A process for studying the disposal concepts was established inorder to study waste management strategies.
◼ Goals
 To create the image of waste packages and present disposal
methods based on the classification of solid waste.
◼ Implementation details
 In FY2019, several disposal methods were studied in accordancewith the classification of solid wastes, based on the survey details
on domestic and overseas disposal concepts and safetyassessment techniques and based on the information on solid
waste properties. The items that need to be studied for theestablishment of safety assessment techniques for each disposal
method were identified. The concepts of the studied disposal
methods were organized to clarify the image of waste packages
resulting from each of the treatment technologies applicable inaccordance with the characteristics of solid waste.
 In FY2020, studies will be continued on multiple disposal methods
in accordance with the classification of solid wastes, based on thesurvey details on domestic and overseas disposal concepts andsafety assessment techniques and based on the information on
solid waste properties. Necessary information regarding theidentified items that need to be studied will be collected andconsolidated. The image of waste packages resulting from each of
the treatment technologies applicable in accordance with thecharacteristics of solid waste, for the studied disposal methods will
be studied.
◼ Indicators to determine goal achievement
 (FY2019) The items that need to be studied for the establishment ofsafety assessment techniques for each disposal method should bepresented. The concept for clarifying the image of waste packages
should be presented.
 (FY2020) Multiple disposal methods based on the classification ofsolid waste should be presented. Necessary information regarding
the identified items that need to be studied should be presented.
The image of waste packages should be presented.

Processing,recycling, disposal
at VLLW disposal
facility, etc.

Disposal at
LLW disposal
site

Processing,recycling,
disposal at VLLW
disposal facility,etc.

Disposal at
LLW disposal
site

Fig. 1 Waste disposal strategy in the UK

Understanding theproperties of wastes[Characteristics]•Large amount of waste•Various types of wastes•High uncertainty

Study on disposal
concepts•Depth of disposal•Structure of the facility•Environmental
conditions, etc.

Plan for manufacturing
waste packages•Measurement / sorting•Solidification treatment•Disposal container, etc.

Performance evaluation of
disposal site
•Evaluation scenarios•Evaluation model•Evaluation parameters, etc.

Fig. 2 Process for studying the disposal concepts©International Research Institute for Nuclear Decommissioning151

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques① Study on disposal methods in accordance with classification of solid wastes andcollection & consolidation of information for establishing safety assessment techniques

2019

2020

1. Creation and review of research plan
2. Study of specifications and contracts
pertaining to outsourced task

3. Organization and classification of waste
characteristics

Targeted waste in FY2019

Targeted waste in FY2020

4. Study of disposal concepts
Targeted waste in FY2019

Targeted waste in FY2020

5. Study on safety assessment techniques
6. Study on the requirements for waste
packages
7. Study on the image of waste packages◼ Implementation details (proposed)
1. The research plan will be created and reviewed.
2. Specifications and contracts of the outsourced task will be studied.
3. Waste characteristics will be organized and classified.
4. Disposal concepts suitable for the characteristics of solid wastes will be studied.
5. Performance of disposal site will be evaluated for the studied disposal concepts.
6. The requirements of waste packages from the perspective of disposal will be studied.
7. The image of waste packages based on the applicable treatment technology will be studied.
©International Research Institute for Nuclear Decommissioning152

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

Results and issues in FY2017 – FY2018, and study policy for FY2019 – FY2020<Major results of FY2017 – FY2018>
1. Survey of case examples of overseas disposal sites● Case examples from overseas disposal sites were identified and their applicability to 1F waste disposal was
evaluated.
2. Establishment of a process for studying the disposal concepts● The process for studying the disposal concepts, in which the plan for waste conditioning (measurement /sorting, solidification treatment, containers, etc.), the disposal concepts, and performance evaluation of disposalsites are studied serially, was established.
3. Case study on multiple disposal concepts● A case study on improved confinement function of the disposal facility, disposal container, and solidifications
respectively was conducted, and it was confirmed that the effect could be verified and feedback could be given.

<Issues and study policies for FY2019 – FY2020>
1. Study on disposal concepts considering the characteristics of each type of waste▸ In the study to be conducted during FY2019 - FY2020, the characteristics of each type of waste will be studied
and consolidated.
▸ The process for deriving the disposal concepts based on the characteristics of each type of waste will be
studied.
2. Identification of issues in establishing safety assessment techniques and items that need to be studied▸ Safety assessment of hypothetical waste packages and hypothetical disposal sites will be conducted inconsideration of latest regulations, and the issues in safety assessment and necessary parameters will be studied.
3. Study on the concept of waste package requirements from the perspective of disposal safety▸ The concept of waste conditioning and its requirements will be studied based on the characteristics of eachtype of waste.
©International Research Institute for Nuclear Decommissioning153

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

Course of action for FY2019-FY2020 (1/2)

Table 1 Examples of 27 typesof accident waste
Types of waste<Items discussed by the concerned parties while formulating the plan>

Kurion● 27 types of accident wastes from the assumed inventory studied by means of characterization, will be
studied, based on the fact that these cover a wide range of wastes and that the results of studies
conducted so far can be used.
● At present, since the inventory is uncertain and there is uncertainty in dose evaluation parameters as
the disposal sites are yet to be determined, it is estimated that the result of dose evaluation will include
significant uncertainty.

Waste
generated fromwater
treatment

AREVA sludge
SARRY
Carbonate precipitation slurry
Used adsorbent from ALPS
(Titanate)<0.005 mSv/hTherefore, in this study, the focus will be on establishing a process for determining the disposal methods
and on identifying related issues, using the results of dose evaluation as reference values.

0.005-0.1 mSv/h
Debris● First, as a starting point, preliminary dose evaluation will be performed. The dose, when 27 types ofaccident waste is considered as raw waste and hypothetically disposed of in trenches, pits or at medium
depths, will be evaluated, and compared with the standard value including the one under study by the
Nuclear Regulation Authority. As a base for the disposal concepts, a preliminary study will be carried out
on which disposal classification is expected for each type of waste.
Based on this disposal classification, the aim is to study the treatment and disposal measures based on
the characteristics of each type of waste and to present a disposal concept based on the waste
classification.
ProposedDistribution of
27 types ofdisposal concept
raw waste
waste
Characteristics
Kurion
AREVA
Carbonate precipitation
slurryDebris 1
Debris 2
RPV waste (L1 before theaccident)
Metal inside PCV( L2 before
the accident)・・・<Trenchdisposal>
Waste A, B・・・
Preliminary dose
evaluation<Pit disposal>
Waste C, D・・・

Figure 1 Flow of Study・・・

of each waste Waste A
Study of disposal
concepts
Dose evaluation

TreatmentaDisposal x

Treatmentb

Disposal y0.1-1 mSv/h
1-30 mSv/h>30 mSv/h
RPV waste (L1 before theaccident)

Dismantling
waste

Metal inside PCV (L2 before theaccident)
Metal inside the building (L3before the accident)
Concrete inside the building(Not released before theaccident)・・・

Waste B--

Treatment Disposal zc・・ ©International Research Institute for Nuclear Decommissioning・154

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

Course of action for FY2019-FY2020 (2/2)
I. Preliminary dose evaluation foreach type of waste

II. Organization of waste information required
for the study of disposal concepts

III． Evaluation of characteristics foreach type of waste
IV. Selection of waste to be studied in
FY2019The 27 types ofwastes are
selected astargets(Comprehensive
study)

Narrowing down

▸ Information and characteristics of the waste (II, III)
▸ Domestic and overseas disposal case examples(Studies in FY2017 - FY2018)
Process for studying the disposal concepts(Studies in FY2017 - 2018)
IX. Process until waste conditioning (treatment,
segregation, measurement, etc.)#

Process for studying the disposal concepts (Studies in FY2017- 2018)

V. Study on the process until wasteconditioning (treatment, sorting, measurement,etc.)#

Feedback ofevaluation results<Scope of implementation in FY2020 (planned)>・Selection of waste to be studied based on the study in FY2019
・Study of disposal concepts for the selected waste and its dose
evaluation

VIII． Selection of waste to be studied
in FY2020▸ Information and characteristics of the waste (II, III)
▸ Domestic and overseas disposal case examples(Studies in FY2017 - FY2018)

VI. Study on disposal
methods#<Scope of implementation in FY2019＞・Preliminary dose evaluation of waste
・Organization of waste information and evaluation of characteristics
・Study of disposal concepts for the selected waste and its dose
evaluation

Targeting theselected
waste(Detailedstudy)

X. Study on disposal
methods#

Ⅺ． Dose evaluation

Feedback ofevaluation results

VII. Dose evaluation#：In this study, the process up to waste conditioning and thedisposal methods are collectively referred to as the "disposal
concept".

Output Image●Proposed disposal concept based on the type of waste●Issues with the dose evaluation techniques and items
that need to be studied●Requirements of waste packages from the perspective ofdisposal safety©International Research Institute for Nuclear Decommissioning155

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidation of
information for establishing safety assessment techniques

I. Preliminary dose evaluation for each type of waste

PurposeOverview of
evaluation
methods
Inventory(27 types ofwaste)The 27 types of accident waste (secondary waste generated from contaminated water
treatment, debris, felled trees, dismantling waste) will be allotted for trench disposal, pitdisposal, medium depth disposal, and geological disposal.
▸ In I, raw waste is assumed. However, the sensitivity to exposure dose during the elution
period in the Groundwater Scenario is evaluated by a parameter study.
▸ In I, overseas case examples (artificial barriers, etc.) surveyed during FY2017 - FY2018 are
not considered.

Groundwater Scenario and Nearing Scenario are used.
The assumed inventory for the 27 types of accident waste (secondary waste
generated from contaminated water treatment, debris, felled trees, dismantling waste),which is the result of the FY2017 - FY2018 IRID study, was used.
Placed by humans

Dose evaluationin Nearing
Scenarios

Dose evaluationin Groundwater
Scenarios
Allotment withrespect topreliminary disposal
classification

The following parameter surveys were conducted.
▸ Inventory (50 %ile, 95 %ile)▸ Generation time (trench: 50 to 600 years, pit: 300 to 600 years)

Exposure route•Person staying at site
•Excavation worker•Geotechnical engineer•Lab analyst*In the preliminary dose evaluation, a personstaying at the site is evaluated as a representative
example.
Use of well
Personwater
Riverinvolved infishery

The following parameter surveys were conducted.
▸ Inventory (50 %ile, 95 %ile)▸ Impact on disposal depending on the presence of affecting
substances
Ground water# In the preliminary dose evaluation,▸ Elution period (1y to 100,000y), amount of penetrated water
the river route is evaluated as arepresentative example.
(reference value x 1 to x 0.001)
The 27 types of accident waste (secondary waste generated from contaminated water treatment，
debris，felled trees，dismantling waste) were allotted for trench disposal, pit disposal, medium
depth disposal, and geological disposal.
©International Research Institute for Nuclear Decommissioning156

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

I. Preliminary dose evaluation for each type of waste
Summary of
preliminary dose
evaluation results

The 27 types of waste were allotted disposal classes assuming that these consist mainly of raw
waste.
In the Nearing Scenario, the dose when the scenario occurs largely depends on the disposal method (stationary
density).
Following could be the countermeasures for this scenario.
・Reducing the maintainability of the inventory
・ Extension of management period
In the Groundwater Scenario, most of the waste needs to be disposed in deep depths if no measures are taken.
Following could be the countermeasures for this scenario.
・Reducing the maintainability of the inventory
・The elimination of impacts of substances with impact on disposal does not reduce the distribution and sorption
performance of the natural barriers and of areas where distribution equilibrium is achieved.
・ Nuclide migration measures (Solidification treatment, installation of artificial barriers)
Sensitivity evaluation of the effects (changes in dose and disposal classification) upon implementing the abovemeasures was performed and the results were organized based on the type of the waste.

©International Research Institute for Nuclear Decommissioning157

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

II. <Organization of waste information>
The waste information necessary for studying the disposal concepts was organized, based on
the results of the preliminary dose evaluation and the IRID studies conducted so far.

Organized items (excerpts)

Targeted waste
27 types of waste linked to the inventory
Secondary waste generated from contaminatedwater treatment▸ KURION▸ AREVA▸ SARRY▸ Slurry 1 Iron co-precipitation▸ Slurry-2 Carbonate precipitation▸ ALPS-4 Ag impregnated activated carbon▸ ALPS-2 Titanate
etc.
Debris, felled tress▸ Debris 1 to 5
▸ Felled trees
Dismantling waste▸ RPV waste (L1, L2 before the accident)▸ PCV waste (metal) (L2, L3 before theaccident)▸ PCV waste (concrete) (L2, L3 before theaccident)▸ Waste inside the building (metal) (L3 before
the accident, not released)▸ Waste inside the building (concrete) (L3before the accident, not released)<Source of information>
Basic Information▸ Quantity
Inventory estimatedRadioactivity properties
during the FY2018▸ The amount of radioactivity per nuclide
study▸ Uncertainty in the amount ofradioactivity▸ Form of disposal
I. Preliminary dose
Nearing Scenario, Groundwater
evaluation
Scenario

Physical properties▸ Physical properties
Shape, size, shape stability, dispersibility,
Solubility, confinement property▸ Chemical properties
Material/substance, moisture content,
inclusions (toxic substances,
impacting substances, organic substances)

Storage and treatment▸ Storage status

Study results ofWaste project,public information,
etc.

Status of occurrence (existing occurrence, future
occurrence), storage form

▸ Treatment options©International Research Institute for Nuclear Decommissioning158

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

Example of waste in large “quantity”

III． Evaluation of waste characteristics▸ Type of debris - Debris 1 (0 to 0.005 mSv/h), etc.

The characteristics of 27 types ofwaste were evaluated focusing on
quantity, diversity, and uncertainty.

The entire quantity of debris is to the order of a few hundred thousand m 3, which is about the same as the maximum planned amount at
Rokkasho.

▸ Concrete inside the building (not released before the accident)
The quantity is to the order of a few hundred thousand m 3 as the material has become radioactive waste due to contamination at the time of theaccident. The quantity changes depending on the state of contamination.

▸ KURION, carbonate slurry
The quantity of secondary waste generated from contaminated water treatment is relatively small, however, as an adsorbent KURION is greatest in number andas slurry sludge the quantity of carbonate slurry is largest.

Examples of waste with great “diversity”▸ General secondary waste generated from contaminated water treatment
Waste specific to 1F. The amount of radioactive substances and chemical substances mixed in the same type of waste differs depending on the operation
history of the facility.

▸ Types of debris
Although the main waste comprises metals and concrete, there are variations in contamination (amount, form of contamination) and contaminants (type,amount).

▸ Dismantling waste

Similar to the dismantling waste generated during general reactor decommissioning, the waste itself is radioactivated, and the extent of contamination byradioactive gas, fine particles, and contaminated water is large.
Contamination is likely to occur due to other types of waste and chemicals during fuel debris retrieval and dismantling.

▸ Concrete inside the building (not released before the accident)
It is not classified as radioactive waste in general decommissioning and is a contamination unique to 1F.

Examples of waste with high “uncertainty”▸ General secondary waste generated from contaminated water
treatment

The amount generated may fluctuate depending on the operating conditions of the equipment in the future. There is also great uncertainty in the radioactivity
properties due to variations in the amount of nuclides adsorbed.

▸ Dismantling waste
Uncertainty is high as the radioactivity properties vary depending on the contamination status.

▸ Concrete inside the building (not released before the accident)
Uncertainty is high as the amount of waste generated and the radioactivity properties vary depending on the contamination status.
©International Research Institute for Nuclear Decommissioning159

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

IV. Selection of waste to be studied each year
8 types of waste were selected for detailed study (V to VII) based on the results of the studies conducted so far.
Table 1
No.

Selected waste

Waste

Category

Overview of waste

1

ＫＵＲＩＯＮ

Secondary waste generated from contaminatedwater treatment

Adsorbents such as zeolite that adsorbed radionuclides such as C-137, and metal containers containing them

2

ＡＲＥＶＡ

Secondary waste generated from contaminatedwater treatment

Precipitates containing radionuclides such as Sr-90 (including ferrocyanide, barium sulfate, etc.)

3

Slurry-2 Carbonate precipitation

Secondary waste generated from contaminatedwater treatment

Sr-90 is a typical nuclide of carbonate slurry (Main components are CaCO3, Mg(OH)2) generated in water
treatment facilities.

4

Debris １ (0 – 0.005 mSv/h)

Various debris

The main materials are metal and concrete contaminated by the crushing and dismantling of structures such asbuildings inside the premises.

5

Debris 2 (0.005 – 0.1 mSv/h)

Various debris

Similar to Debris 1 except that it has a higher amount of radioactivity

6

Waste from pressure vessel (L1 before
the accident)

Dismantling waste

Fuel debris, cesium, etc. from the accident that are adhered to the activated metal in the pressure vessel

7

Waste inside PCV (metal) (L2 before theaccident)

Dismantling waste

Fuel debris, cesium, etc. from the accident that are adhered to the radioactivated metal (PCV body, equipment)in the containment vessel.

8

Waste inside the building (concrete) (notreleased before the accident)

Dismantling waste

Highly concentrated contaminated water and / or contaminated air that are adhered to the building and the nonradioactive concrete in the building<Representativeness of selected waste>
Multiple wastes are selected from the categories of secondary waste
generated from contaminated water treatment, various debris,and dismantling waste.

Radioactivity
concentration

Covers waste from extremely high-level radioactive waste(KURION, waste from pressure vessel) to low-level
radioactive waste (Debris 1).

Expected disposal
concepts

Covers waste from L3 to L1 and waste that is expected tobe geologically disposed.

Important nuclides

Includes short half-life nuclides and long half-life nuclides caused byradioactivation and contamination.

Physical form
Types of substances

Includes slurries, building materials (building concrete),and metal-based reactor components.

Uneven radioactivity
distribution

Includes relatively uniformly contaminated secondary waste generated
from contaminated water treatment and solid waste with unevensurface contamination.

Affectingsubstances

Includes disposal of various debris that may containorganic matter, and substances affecting theenvironment (AREVA).

Water content andmobility

Includes wet waste that may require solidification to reduce the riskof hazards.

Quantity of
raw waste

Covers waste from the two most abundant raw wastes(Debris 1 and concrete inside the building (not released
before the accident)) to the least abundant raw waste.

Time of occurrence

Includes existing waste currently stored at the site (debris, etc.) anddismantling waste that will be generated decades later.

Category©International Research Institute for Nuclear Decommissioning160

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

IV. Selection of waste to be studied each year
Study policy for “V. Study on the process until waste conditioning”, “VI. Study on disposal methods” and “VII.
Dose evaluation”① In “V. Study on the process until waste conditioning”, multiple treatment options shall be studied based on
the characteristics of each waste.
② In “VI. Study on disposal methods”, the concept of disposal site with artificial barriers to reduce the amountof penetrated water shall be studied for the three categories, namely, surface disposal concept (L3), nearsurface disposal concept (L2), and medium depth disposal concept (L1).
③ In “VII. Dose evaluation”, the treatment options studied in V and VI and the model parameters forincorporating the artificial barrier into dose evaluation, shall be studied. Here, with long-term institutional
control as the pre-requisite, parameters such as nuclide migration and biosphere shall be reviewed withreference to cases from other countries.
Preliminary dose evaluation andcharacteristics evaluation(I to III)
Target: 27 types of waste(comprehensive)

Sensitive evaluation of time ofoccurrence, Kd, elution rate,
amount of penetrated water, etc.

Study on the direction of
measures

Study of disposal concepts anddose evaluation(V to VII)
Target: 8 types of selected waste● Setting of disposal options and disposal methods● Provision of models and parameters for dose
evaluation

Study on specificcountermeasures andevaluation methods©International Research Institute for Nuclear Decommissioning161

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

V. Study on the process until waste conditioning
Multiple processes up to waste conditioning from the concrete inside thebuilding (not released before the accident) are studied and the amount ofradioactivity and quantity for each waste package that gets generated is set.
Table 1 Flow of each treatment option(Illustration of waste inside the building (concrete) (not released before the accident))

Option A

Cutting withdiamond wire

Breaking, smashingand separation ofmaterials

Chipping ofcontaminatedsurface
Contaminateddebris
Decontaminatedsubstance

Substanceswith surface
contamination

Breaking, smashingand separation ofmaterials

Filling of grout
inside ISO
container

Concreteblock

Concrete waste, steel
reinforcement(CL level)

Packages withsurfacecontamination

Concrete waste(CL level)

Concrete waste
packageSteelreinforcement(CL level)

100%・100%

0%・98%

100%・2%

0%・89%

100%・17%0%・1%Treatment

Change inradioactivity
and volumeNote)

Option C

Concrete inside the building (not released before the accident)

Raw waste

Disposal /ReuseOption B

Fig.1 Sample image of building concrete

Separatedbased onradioactivity

Radioactive
debrisPackaging andfilling of grout©International
Research Institutefor Nuclear
DecommissioningNote) Set with reference to NEA / RWM / R (2011) and NDA / RWM / 120 (2015). %is the percentage
againstraw waste.

162

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

V. Study on the process until waste conditioning
Multiple processes up to KURION waste conditioning are studied and the amount of radioactivity and quantity for each waste package that gets generated
is set.

Table 1 Flow of each treatment option (Illustrating KURION)
Option A

Option B

Raw waste

Option C

Option D

KURION

Extraction andseparation of adsorbents

Adsorbent

Temporarystorage container

Treatment

Extraction andseparation of adsorbents

Adsorbent

Temporarystorage container

Adsorbent

Temporarystorage container

Melting

Melting

Slag
Filling of grout
inside SS container

DryingMelting

Dehydration
compression
Filling of grout
inside SS
container

Extraction andseparation of adsorbents

Vitrification

Slag

Filling of grout
inside highly
durablecontainer
provided withshielding

Filling of grout
inside SS
container

Slag
Storage in
SS container
Filling of grout
inside SS
container

Disposal /Reuse

Adsorbent
packageSlagpackage

Metal(CL level)

Adsorbent
packageSlagpackage

Metal(CL level)

Adsorbent
packageSlagpackage

Metal(CL level)

Dried
raw
waste

Change inradioactivity andvolume Note )

99%・
198%

1%・
3%

0%・
96%

99%・
60%

1%・
3%

0%・
96%

99%・
36%

1%・
3%

0%・
96%

100%・
100%

©International
Research
Institutefor Nuclear
DecommissioningNote) Set with reference to IAEA. TRS No 401 (2001), NDA / RWM / 120 (2015), IHI Engineering Review/ Pp 25 Vol
5, Issue
1 (2018).
% is thepercentage
against raw waste.

163

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

V. Study on the process until waste conditioning
Table 1 Treatment options for secondary waste generated from contaminated water treatment
Waste

Treatment
optionExplanation

Waste package

Change in radioactivity
and volume Note )A

Grout filling of sludge in stainless steel (SS) container

Sludge package (grout filling in SS container)

100%・330%

B

Compression of sludge by dehydration and storage in a highly durable container provided with
Sludge package (high durability container)
shielding

100%・100%

C

Vitrification of sludge and storage in SS container

Vitrified solid package (SS container)

100%・60%

Raw waste

Disposal of raw waste (treatment / conditioning is not studied)Raw waste

100%・100%99%・198%

A

Adsorbent package (grout filling inside SS container)
Segregation of temporary storage containers and adsorbents
Grout filling of primary waste (adsorbents) in SS container
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS Slag package (grout filling inside SS container)
container
Metal (CL level)

99%・60%B

Segregation of temporary storage containers and adsorbents
Adsorbent package (grout filling in highly durable container)
Compression by dehydration of primary waste (adsorbents) and grout filling in a highly durable
container provided with shielding
Slag package (grout filling inside SS container)
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS
Metal (CL level)
container
Vitrified solid package (SS container)
Segregation of temporary storage containers and adsorbents
Vitrification of primary waste (adsorbents) and storage in SS container
Slag package (grout filling inside SS container)
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS
Metal (CL level)
container

99%・36%C

D

Dried inside the temporary storage container and disposed directly

Dried raw waste

100%・100%

Raw waste

Disposal of raw waste (treatment / conditioning is not studied)Raw waste

100%・100%Slurry package (grout filling inside SS container)

Segregation of temporary storage container and slurry.
Grout filling of primary waste (slurry) in SS container
Slag package (grout filling inside SS container)
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS
container
Metal (CL level)

99%・195%

Slurry package (high durability container)
Separation of slurry and high-performance containers.
Compression by dehydration of primary waste (slurry) and storage in a highly durable container
provided with shielding
Slag package (grout filling inside SS container)
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS
container
Metal (CL level)

99%・53%Vitrified solid package (SS container)
Segregation of slurry and high-performance containers.
Vitrification of primary wastes (slurry) and conditioning in SS container
Slag package (grout filling inside SS container)
Melting of secondary wastes (temporary storage container) and grout filling of slag inside SS
container
Metal (CL level)

99%・32%

D

Dried inside the temporary storage container and disposed directly

Dried raw waste

100%・100%

Raw waste

Disposal of raw waste (treatment / conditioning is not studied)Raw waste

100%・100%

AREVA

KurionA

Slurry-2
Carbonate
precipitation

B

C

1%・3%
0%・96%

1%・3%
0%・38%

1%・3%
0%・38%

1%・3%
0%・45%

1%・3%
0%・45%

1%・3%
0%・45%©International
Research
Institutefor Nuclear
DecommissioningNote) Set with reference to IAEA. TRS No 401 (2001), NDA / RWM / 120 (2015), IHI Engineering Review/ Pp 25 Vol
5, Issue
1 (2018).
% is thepercentage
against raw waste.

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V. Study on the process until waste conditioning
Table 1 Treatment options for various types of debris and dismantling waste
Waste

Concrete inside
the building(Not released
before theaccident)

Treatment
OptionExplanation

Waste package

A / Raw waste

Cut with diamond wire, made into concrete block and disposed directly

Concrete block

B

C

Concrete waste (CL level)
Chipping of parts of the contaminated surface
Steel reinforcement (CL level)
Grout filling of contaminated surface inside ISO container
Package with surface contamination (grout filling inside ISO
container)
Broken and sorted into radioactive concrete waste, non-radioactive concrete Concrete waste (CL level)
Steel reinforcement (CL level)and steel reinforcement
Grout filling of radioactive concrete waste into ISO container

Metal inside
PCV(L2 before theaccident)

RPV waste(L1 before theaccident)
Debris 1(0 – 0.005
mSv/h)

Package of dismantled solidifications (grout filling inside SS
container)
Package with surface contamination (grout filling inside SS
container)
Radioactivated metal package (grout filling inside SS
container)
Raw waste
Package of dismantled solidifications (grout filling inside SS
container)

100%・100%
0%・98%
0%・1%
100%・2%
0%・89%
0%・1%

100%・17%A

Disassembled, shredded and grout filled in stainless steel (SS) container

B

Dismantling and chipping of contaminated surfaces
Grout filling of primary waste (surface contamination) in SS container
Grout filling of secondary waste (radioactivated metal) in SS container

Raw waste

Disposal of raw waste (treatment / conditioning is not studied)A

Disassembled, shredded and grout filled in (SS) container

Raw waste

Disposal of raw waste (treatment / conditioning is not studied)Raw waste

100%・100%A

Volume reduced by crushing and grout filling inside ISO container
Sorted into radioactive debris and non-radioactive debris throughmeasurement
Grout filling of radioactive debris inside ISO container
Disposal of raw waste (treatment / conditioning is not studied)
Volume reduced by crushing and grout filling inside ISO container
Segregated into radioactive debris and non-radioactive debris throughmeasurement
Grout filling of radioactive debris inside ISO container
Disposal of raw waste (treatment / conditioning is not studied)

Debris package (grout filling inside ISO container)
Debris (CL level)

100%・170%
0%・65%

Debris package (grout filling inside ISO container)

100%・60%

Raw waste
Debris package (grout filling inside ISO container)
Debris (CL level)

100%・100%
100%・170%
0%・65%

Debris package (grout filling inside ISO container)

100%・60%

Raw waste

100%・100%B
Raw wasteA

Debris 2(0.005 – 0.1
mSv/h)

Concrete waste package (grout filling in ISO container)

Change inradioactivity andvolume Note 1 )B
Raw waste

100%・255%

Set for each nuclide
Note 2) , 17%
Set for each nuclide
Note 2) , 153%
100%・100%
100%・255%

Note 1) Set with reference to NEA / RWM / R (2011), NDA / RWM / 120 (2015), NWP-REP-120-Issue 1 (2016), ED26.RP / 340737 / PROJ / 00033 Version 2 (2015). % is
the percentage against raw waste.

Note 2) Set for each nuclide as it is accompanied by activation products (Be-10, C-14, Cl-36, Ca-41, Ni-59, Ni-63, Co-60, Mo-93, Nb-94).

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VI. Study on disposal methods

An image of the surface disposal concept (L3)
Backfilling of excavated soil

Soil cover to limit penetration

Wastes filled in containers and loosely
packedIt is assumed that the floor and side walls have low
permeability, and the upper part has a low
permeability layer that reduces the amount ofpenetrated water and is covered with soil. The
disposal facility is backfilled with soil.
Illustration
barrier

Quaternary
depositofartificial

Soil cover
Drainage layer
Lowpermeability
layer

Low permeability watersealing wall

Low permeability watersealing wall
Base rock<Assumptions pertaining to near-field during dose evaluation>
✓
✓
✓
✓
✓
✓
✓
✓

Example of artificial barrier at VLLW disposal site
(Kings Cliffe) in the UK(Based on the 2017-2018 overseas disposal facility
survey results)

The soil clay liner after the soil cover continuously maintains a permeability coefficient of 1E-09 m/s.
The container is sound until the initial damage period, after which the container gets damaged at a unit rate.
The nuclide release begins immediately after the container is damaged. In addition, the contaminants are instantly transported from the waste package to thebackfill.
The saturation periods of grout waste and debris shall be conservatively ignored. Complete saturation occurs immediately after the damage of container.
The backfill is modeled as a mixed compartment with no cracks. Advection is assumed as the dominant transport process to the outside through backfilling.
The effect of the organic complex is considered in terms of the solubility increasing coefficient and the sorption decreasing coefficient.
It is assumed that the soil cover always has a smaller effective permeability coefficient than the bottom and side walls of the facility. Therefore, the flow of
seepage water through the near field is dominated by the permeability of the soil cover.
The transport through the unsaturated layer is ignored. That is, the nuclide transport from backfilling to the outside goes directly to the saturated layer.
©International Research Institute for Nuclear Decommissioning

166

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

VI. Study on disposal methods

An image of the concept of near-surface disposal (L2)
Similar to surface disposal facilities. The possible differencesare:・ Large amount of cement is used as grout / backfillingmaterial
・ Highly durable barrier that suppresses water flow
・ Low permeability transportation route

Waste package
Cement grout / backfilling
Soil cover that limits
penetration

Quaternary
deposit

It consists of a reinforced concrete vault in which
low-permeability clay is laid on the floor and theupper part is covered with multiple soils. The
disposal facility is backfilled with cement grout#.
The bentonite wall surface controls seepage water.
#: When burying vitrified solidifications, bentonite is used forbackfilling.

Diagram of artificial barrier
Surface soil / rooting
layer
Moistureretention
layer
Filter layer (sand / coarse
gravel)Bio-intrusion
barrierDrainage layer (course gravel) Round gravel smaller than 150 mm

Foundation slab
Low permeability watersealing wall

Base rock

Geotextile protection layer
Geomembrane

Low permeability
water sealing wall

Low
permeability
barrier
Geotextile separation layer
Waste: Flow of water

Example of artificial barrier at LLW disposalsite (Drigg) in the UK(Based on the 2017-2018 overseas disposal
facility survey results)<Assumptions pertaining to near-field during dose evaluation>
Same as surface disposal except for the following items.
✓ The initial pH of backfilling is maintained for a long period of time, ignoring the evolution of backfilling pH over time.

©International Research Institute for Nuclear Decommissioning

167

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

VI. Study on disposal methods

An image of medium depth disposal facility (L1)

Depth about 50m
Cement grout / backfilling

Concrete

Waste package

It is installed at about 50m below theground. It consists of a tunnel withconcrete walls. It is plugged in andclosed. The voids are backfilled withcement grout.

Low permeability
liner

Bentonite
Lining concreteRock bolt

Base rock<Assumptions pertaining to near-field during dose evaluation>
Same as surface disposal except for the following items.
✓ The permeability coefficient of the barrier of the medium depth burial facility shall be similar to the structural concrete typically used in nuclear
facilities. As a result, the initial permeability coefficient is set to 1E-10 m/s, and after 5000 years, the function of the permeability coefficient of themedium depth burial facility will be completely lost, and the permeability coefficient will become the same as that of the base rock.
©International Research Institute for Nuclear Decommissioning168

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniquesVII. Dose evaluation – Overview of evaluation
Setting of evaluation conditions for dose
evaluation
scenarios,parameters -▸ The scenarios and reference dose used here are the same as those used for I. Preliminary dose evaluation (some scenarios that were notconsidered earlier, are also evaluated).
▸ For nuclide migration / biosphere parameters, the parameters that have a proven track record overseas are used.
▸ Study on the evaluation methods for the amount of penetrated water, nuclide release, and time of container damage (an example is shown on thenext page).

Table 1

Overview of the dose evaluation conditions and comparison with the preliminary dose evaluation

Major items

Items

VII Dose evaluation<Comparison> I. Preliminary dose evaluation

Scenario

Nearing scenario

Land users (L3, L2), excavators (L3, L2),
Concentration limit (L1)Land users (L3, L2), concentration limit (L1)300 μSv/y (L3), 1 mSv/y (L2), 20 mSv/y (L1)

Nearing scenario reference dose
Groundwater scenario

River users, sea users, well users
10 μSv/y per type of waste

Groundwater scenario reference
dose
Parameters

Institutional control period (y)Nearing scenario inventory
Release of nuclides
Container damage timeAmount of penetrated water
Migration distance to the well
Migration distance to the river
River flow
Nuclide migration parameters in
cement-based materials and soil
Biosphere transition coefficient
Changes in distribution - sorption
coefficient due to substances having
an impact on disposal

River users, sea users

About 400y (until 2400)

50 to 600y (L3), 300 to 600y (L2)

Takes groundwater migration into consideration
Does not take groundwater migration into consideration
Set for each waste package
Instant release (Raw waste)
Set for each container
Not used (Raw waste)
Controlling the amount of penetrated water by means of
Typical setting in Japan (L3:1, L2:0.1, L3:0.16 m/y)
artificial barrier
50 m (L3, L2)，150m (L1)
100 m (L3, L2)，300m (L1)
107 m3/yData from LLWR in the UK (LLWR 2011 ESC)

Typical setting in Japan

Data from LLWR in the UK (LLWR 2011 ESC)

Typical setting in Japan

Set based on the data from LLWR in the UK (LLWR 2011
ESC)

Two sets of distribution - sorption coefficient are set and are
evaluated individually.

<Assumptions pertaining to the geosphere and biosphere in dose evaluation>✓✓✓

The geosphere shall be a porous medium.
The biosphere model of LLWR2011ESC is applied. These are time-dependent models. The effective dose of PEG is evaluated by multiplying the nuclide
flux in ocean and geosphere by the biosphere coefficient.
The impact of organic matter in the geosphere is considered by evaluating the concentration of organic matter in the geosphere and the sorption
inhibition coefficient. This method is the same as the method for considering the impact on backfilling.
©International Research Institute for Nuclear Decommissioning169

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniquesVII. Dose evaluation - Example of model and parameter settings<Setting the amount of water penetrated from artificial barriers in L3 and L2
facilities>

Burial disposal
facility

Artificial barrier

Amount of penetrated
water (initial)

Degradation behavior

L3

Clay liner

3.16e-2 m/y―Clay liner

3.16e-2 m/y―Geomembrane(Asphalt sheet)

3.16e-5 m/yL2Damage starts from 150y
Loss of function at 1000y

Amount of penetratedwater

The L3 facility is assumed to have a clay liner, and the L2 facility is assumed
to have an artificial barrier composed of a clay liner and geomembrane.

Elapsed time after burial<Reference> SERCO/E.003796/010 issue Feb 2011

<Setting of container confinement function>

Fig. 1Changes in the amount of penetrated water overtime at the L3 and L2 facilities
ISO container

The container shall have the function of trapping radionuclides from the waste
package for a certain period of time. In addition, empirically, the settings wereconfigured such that once the damage starts the damage of the container isdistributed over the same period as the damage time, and the nuclides are
released in stages over a period of the same length as the damage period.

Type of container
ISO Container
Stainless steel container
Highly durable container
with shielding

Damage time (yearafter closing)
10
300
1000

Damage period(year)
10300
1000<Reference> SA/ENC-0658 issue Feb, 2007

Flexible container
bag
Stainless steel container

Highly durable container withshielding

Cost, durability

Fig. 2 Illustration of disposal containers©International Research Institute for Nuclear Decommissioning

170

(b) Presentation of disposal methods, and development of safety assessment techniques【Confidential】① Study on disposal methods in accordance with classification of solid wastes and collection & consolidationof information for establishing safety assessment techniques

Summary and Challenges
Summary✓

The accident waste was comprehensively classified into 27 types and preliminary dose evaluation was performed to determine their anticipateddisposal class, indicating their disposal in trench, pit, medium depth, or geological disposal. A parameter survey was conducted on the percentile of theinventory, the sorption distribution coefficient, the time of occurrence of the Nearing Scenario, the elution rate of the waste package, and the amount ofpenetrated water, and the sensitivity was evaluated with respect to the changes in the disposal classification, when respective measures were
implemented. (I)✓The information necessary for studying the disposal concepts pertaining to the 27 types of waste was consolidated and the characteristics were
evaluated. From these results, 8 types of waste were selected, for which disposal concepts were studied in detail. (II to IV)✓

For each of the selected wastes, multiple possible treatment options were presented. In addition, the disposal images of L3, L2, and L1 with theartificial barrier installed were presented. (V, VI)✓

The models and parameters for the dose evaluation of disposal concepts were provided, dose evaluation for multiple treatment options was performed,
and the expected disposal class was presented. (VII)

Challenges
The evaluation of this study presumes that there is a large amount of uncertainty, and the challenge for the future is to alleviate theuncertainty of the techniques developed here. Following are the challenges for each type of uncertainty. Challenges pertaining toindividual waste are described in VII.
✓

Uncertainty of inventory: First of all, it is important to provide feedback on characterization. Then, it is necessary to consider the
methodology for alleviating the uncertainty of inventory. For example, C-14 and I-129 are the dominant nuclides in many wastes. The
uncertainty can be expected to be reduced by increasing the accuracy of these inventories on a priority basis.

✓

Uncertainty in parameters: Overseas literature is referred to for some of the safety evaluation parameters. It is necessary to consider the
regulations in Japan and the characteristics of waste at 1F. For example, with respect to the performance of solidifications, more realistic
parameters need to be provided, such as by incorporating the study results of latest treatment methods.

✓

Uncertainty of disposal site: The challenge is that there is significant uncertainty depending on the disposal site. Although thisuncertainty cannot be eliminated until the disposal site is determined, countermeasures must be studied in consideration of thepossibility that the disposal site may not be determined early. One approach is to set parameter sets for multiple hypothetical disposalsites and study the site uncertainty through mutual comparison.
©International Research Institute for Nuclear Decommissioning171

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques Development of techniques for assessing impact of affecting substances, etc. on disposal- Goals and implementation details -◼
Results
 Nuclide migration parameters and barrier materials that have a highpriority in assessing the impact on nuclide migration during disposal wereselected based on the investigation of substances having an impact,which are likely to be contained in solid waste, and the assumed conceptof disposal.
 The approach towards the method for assessing the understanding of theaffecting processes based on the substances having an impact, and theimpact on nuclide sorption parameters based on the data volume, wasestablished. (Fig.1)◼
Goals
 To develop methods to quantitatively evaluate the impact of various
affecting materials on nuclide migration during disposal along with thedata required for the evaluation, and to incorporate them in the studies on
the feasibility of disposal and as the basic information for safety
evaluation.
◼
Implementation details (Fig.2)
 In FY2019, the findings that had been insufficient to evaluate the impactof the main substances affecting nuclide migration were investigated andthe data was expanded together with improving the methods forquantitatively evaluating the impact (Sorption Reduction Factor: SRF) on
nuclide sorption during disposal.
 In FY2020, the investigation of insufficient findings and expansion of data
will continue, impact assessment methods for nuclide sorption will beestablished, and trial settings will be specified for the nuclide sorption
parameters (distribution coefficient: Kd) and their uncertainty range withrespect to the current safety evaluation assumption conditions and issues
pertaining to the settings, will be identified.
◼
Indicators to determine goal achievement
 Data required for assessing the impact of the combination of important
affecting substances and nuclides should be expanded and the impact
assessment techniques should improve. (FY2019) Trial settings should be specified for the nuclide sorption parameters andtheir uncertainty range with respect to the safety assessment assumption
conditions and issues pertaining to the same should be identified.
(FY2020)

Types and amounts of affecting material such aswaste components and coexisting materials(Organic material, boric acid, salt, cement, … etc.)

Reference conditions
Barrier materials, pore water chemistry) Direct impact assessment technique: Indirect impact assessment technique:
Direct impact such as reduction in sorption due tocomplex formation between affecting material
components and nuclides

Threshold for impact + Sorption
Reduction Factor
Method : Evaluated from data on rise insolubility
Method : Evaluated from thermodynamic
calculations
Method : Evaluated based on data on impact
of sorption

Kd + Uncertainty

Changes in pH and salt concentration ofpore water in barrier materials due to
THMC changes (especially chemistry)caused by affecting materials

Impact assessment technique forenvironmental conditions such as
pH and salt concentration[Conditional conversion method,
sorption model, etc.]

Kd + Uncertainty

Reference Kd+ Uncertainty

 Direct + Indirect impact:
Consideration of both impacts by combining evaluations  and 

Fig.1 Overview of the method of assessing the impact ofaffecting materials on nuclide sorption

Perspective of data expansion in FY2019-FY2020:
・Expansion of data on impact of boric acid andferrocyanide on actinides
・Expansion of data on insufficient nuclides such asorganic materials and sulfatesFig.2 Sample chart of organization of the Sorption Reduction Factors forthe barrier materials with respect to the affecting factors©International Research Institute for Nuclear Decommissioning172

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Overall flow of the study -Table List of affecting substances (Example ofconsolidation of the relation of impact)
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

陽イオン

NaOH
Ca(OH)2

Na+,Ca2+

Trivalent actinide
Tetravalent actinide
Pentavalent actinide

Hexavalent actinide
HalogenAnionic species

Sorption Reduction Factors (SRF)
Material Material Material Material
B
C
D
E
2
2
1
2

Cs
Sr

1

7

3

10

8

31

Ni (II)

1

10

36

1

12

1

Zr (IV) Presentation
1
10 of impact
10
103

12

109

12

10

12

1

12

1

12

1

12

1

–

–

assessment
data10for 1
1
10
substanceshavingan 102Am (III)1
1010
impacton
disposal,Th (IV)11010106contributes
Np (V) which
1
100
10 to 102
safetyevaluation
U (VI)
1100
10102
Nb (V)

Se

–
–

–
–

➢

二次鉱物生成固相表面収着エッジ量変化

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. Example of organization of relationshipbetween the information on the affecting
process and nuclide interaction③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④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
核種-OH, -CO3錯体
濃度変化

液相pH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

Investigation and organization of information
on target materials, and organization ofrelationships among processes, mechanisms,
and interactions affecting the sorption
behavior
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

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

中性分子

液相中性イオン
組成変化

核種-影響物質の錯体濃度変化e-

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

–
–

1
–

–

Material
F
46–

Sample chart of organization of Sorption Reduction Factors forbarrier materials with respect to affecting factors➢

陽イオン交換
容量変化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

Material
A
1

I

固相溶解

Sorption affecting data that is unavailable/insufficient is being collected, mainly for ferrocyanides,
sulfates, and carbonates.
Evaluation methods appropriate to the volume of data are being studied and assessed.

1.E+066⑧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

10

1.E+05
10 5

Study of evaluation methods formaterials and nuclides thatcannot be evaluated using data
alone
10 4

Collection of data on materials
and nuclides, which is
SRF insufficient, and study ofevaluation
methods based on
Dagradation
II
it

SRF

1.E+0410 4
1.E+03
10 3

Method手法①

10 3

Method手法②

1.E+02
10 2

Kd (m3/kg)

Alkali earth metal
Divalent transition
metal
Tetravalent transition
metal
Pentavalent transition
metal

核種-OH, -CO3錯体
濃度変化

液相pH変化OH-

SRF
Factors)
SRF (Sorption Reduction(Organic substances, seawater component,
boric acid, ferrocyanides, sulfates, carbonates)

Alkali metal

核種スペシエーション
変化➢ Completion of comprehensive investigation of
information and consolidation of generalrelationships with respect to the organized affecting
substances➢ Target substances: 6
substances identified

Representativeelement

陰イオン

固相表面収着サイト状態変化

液相陽イオン
の増加

間隙水へ
の溶解

Identification ofmaterials for study

Element group

核種の収着への影響メカニズム

液相・固相の状態変化

影響物質の溶解

1.E+01
10 1

10 2

Method手法③
1

1.E+0010 0

Impact threshold
影響発現の閾値

10

Sorption
収着低減
Reduction
係数 Factor-110
1.E-01
1.E-0610 -6

1.E-0510 -5

1.E-0410 -4

1.E-03100-310ISAconcentration
concentration [M] (M)
ISA
Example of impact assessment by
10 -1
means of multiple methods
10 -6
appropriate to the volume of
information
Sorption Reduction Factor of Am to cement in an
organic material (ISA) coexisting system-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 material effect of Am on Kd©International Research Institute for Nuclear Decommissioning

173

(b) Presentation of disposal methods, and development of safety assessment techniques Development of techniques for assessing impact of affecting substances, etc. on disposal【Confidential】- Study results up to FY2018 “Selection of affecting materials and barrier materials for evaluation”◼



Secondary waste
generated from contaminated water
汚染水処理二次廃棄物
treatment

）

）

燃
料
デ
ブ
リ

）

）

Dismantling waste (other)

（
そ
の
他

waste

使
用
済
保
護
衣

Dismantling waste (metal)

（
金
属

土
壌

Fuel debris

（
コ
ン
ク
リ
ー
ト

伐
採
木

Dismantling waste (concrete)

瓦礫

Used protective clothing

瓦
礫

Soil

瓦
礫

Felled trees

（
カ
ラ
ム

Dismantling
解体廃棄物

瓦礫／伐採木等
Debris/felled trees

Debris (other)

フ
ィ
ル
タ
ー

Debris (concrete)

樹
脂
系
吸
着
材

Filters

キ
レ
ー
ト
樹
脂

Resin-based adsorbents (column)

フ
ェ
ロ
シ
ア
ン
化
合
物

Chelating resin

酸
化
チ
タ
ン

Ferrocyanides

チ
タ
ン
酸
塩

Titanium oxide

A
g
添
着
活
性
炭

Titanate

炭
酸
塩
ス
ラ
リ
ー

Ag-loaded activated carbon

鉄
共
沈
ス
ラ
リ
ー

Carbonate slurry



除
染
装
置
ス
ラ
ッ
ジ

Iron coprecipitation slurry



○

解
体
廃
棄
物

解
体
廃
棄
物

解
体
廃
棄
物

（
コ
ン
ク
リ
ー
ト

（
金
属

（
そ
の
他

）

）

影
響
物
質

○

シリカ系物質Silica-basedmaterials ○○Iron水酸化鉄
hydroxide
炭酸塩
Carbonate

○

○
○

水酸化
Magnesium hydroxide
マグネシウム硫酸塩
Sulfate原Activated
廃 活性炭carbon
棄
Organic
material
有機物
物
Titanium-basedmaterial
チタン系物質成
フェロシアン分
Ferrocyanides
化合物
金属○
Metals

○
○
○
○

○

○

○

○

○

○

○

○

○

○

○
○

○

○

○

○

○

○

○

○

○

○

○

○
○

○

○

○

○

コンクリートConcrete
デブリ
DebrisMCCIデブリ
MCCI
debris

○

○

○

○

○

○
○
○

ホウ化物・

○

Borides/
炭化物carbides
Coexisting substances◼

Target nuclides were selected from the nuclides to be evaluated for the
existing disposal concepts and from the approach of grouping based on
chemical similarity.

Decontamination system sludge

Barrier materials and nuclide sorption parameters that are common to
all disposal concepts and are expected to be significantly impacted byaffecting substances in areas close to waste, were selected (Fig.1).

第
二
セ
シ
ウ
ム
吸
着
塔

）
ゼオライト
ZeoliteRaw waste components
Affecting substances

Barrier materials, parameters, and nuclides to be evaluatedNo.2 cesium adsorption tower

原廃棄物
Raw
waste

Ferrocyanides, sulfates, and carbonates were selected as raw waste
components that may have an impact among secondary waste
generated from contaminated water treatment that are not very common
but have a relatively high priority.

⇒ Selection was done with respect to these 6 types of affecting substances(Table 1)
◼

セ
シ
ウ
ム
吸
着
塔

Debris (metal)

Organic substances, seawater component, and boric acid solution wereselected as coexisting substances commonly contained in most of theaccident waste.

Cesium adsorption towerTable 1 Major accident waste and affecting substances
contained therein

Affecting substances to be evaluated

海水成分
Seawatercomponent

○

○

○

○

○

○

○

○

○

○

○

○

ホウ酸水
Boric
acid solution
共存
Oil 油分content
物
有機物
Organic substances
質
シリカ系物質Silica-based
substances

○

○

○

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Fuel燃料付着物
deposits

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○： 原廃棄物に含まれる影響物質
O:
Affecting substances contained in raw waste

Combination of affecting substances and barrier materials to beevaluated

The existing findings and methods can be applied to indirect impact due
to chemical changes in pore water. Hence the focus of this study is on
the direct impact of affecting substances for which the existing findings
are insufficient.
For the direct impact, evaluation targets were selected in considerationof the level of sufficiency of existing knowledge (limiting targets forbuffer material).

廃棄体層
Waste
layer
Waste
Waste
material
充填材（モルタル）(mortar)
Distribution
metal
Release
放出率
分配係数
coefficient
rate

廃棄物
activation
放射化金属等

その他廃棄体
Other
waste
Instant
瞬時放出
release・イオン強度変化
・Changein ionic strength
・錯形成 formation
・Complex
・pH、Eh変化
・Changein pH and Eh
・セメント変質・Cement
alteration
・金属腐食
・Metal corrosion

高High

Concrete pit
コンクリートピット（コンクリート）(concrete)

低拡散層
Low
diffusion（モルタル）
layer(mortar)

Distribution
分配係数
分配係数
coefficient(Diffusion（拡散係数）
拡散係数
coefficient)(Permeabilit（透水係数）
透水係数ycoefficient)

Distribution
分配係数
coefficient
Diffusion
拡散係数
coefficient(Permeabilit（透水係数）
ycoefficient)

・Changein ionic・イオン強度変化
strength・錯形成・Complex
formation
・pH、Eh変化
・Changein pH and・セメント変質Eh・Cement alteration

・Changein ionic・イオン強度変化
strength・錯形成・Complex
formation
・pH、Eh変化
・Changein pH and・セメント変質Eh・Cement alteration

Low低透水層
permeability（ベントナイト）layer(bentonite)
Distribution
分配係数
coefficient
Permeability
透水係数
coefficient(Diffusion（拡散係数）
coefficient)

Distribution
分配係数
coefficient
Diffusion
拡散係数
coefficient
Permeability
透水係数
coefficient

・イオン強度変化
・Changein ionic・錯形成
strength
・pH、Eh変化
・Complex
formation
・ベントナイト変質
・Changein pH and Eh・Bentonite alteration

・イオン強度変化
・Changein ionic・錯形成
strength
・pH、Eh変化
・Complex
formation
・鉱物変質・Changein pH and Eh・Mineral alteration

間隙水中影響物質濃度
Concentrationof affecting substances in pore water

岩盤

Bedrock

低Low：評価で考慮する主な核種移行パラメータ

Fig.1 Impact on nuclide sorption in each barrier layer©International Research Institute for Nuclear Decommissioning174

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques Development of techniques for assessing impact of affecting substances, etc. on disposal- Status of data collection and parameter settings Element group
元素群

Cement based

Alkaliアルカリ金属
metal

代表
Representat
ive元素
element

Anionic
species
陰イオン種

Cs
Sr
Ni(Ⅱ)
Sn(Ⅳ)
Nb(Ⅴ)
Am(Ⅲ)
Th(Ⅳ)
Np(Ⅴ)
U(Ⅵ)
I
Se

Element group
元素群

Representat
代表
ive element

Alkaliearth metal
アルカリ土類金属
Divalent
transition metal
Ⅱ価遷移金属Tetravalenttransition metal
Ⅳ価遷移金属Pentavalenttransition metal
Ⅴ価遷移金属Trivalent
actinide
Ⅲ価アクチニド
Tetravalentactinide
Ⅳ価アクチニド
Pentavalentactinide
Ⅴ価アクチニド

Hexavalentactinide
Ⅵ価アクチニド
Halogenハロゲン

Bentonite based

Alkaliアルカリ金属
metal
Alkaliearth metal
アルカリ土類金属
Divalent
transition metal
Ⅱ価遷移金属Tetravalenttransition metal
Ⅳ価遷移金属Pentavalenttransition metal
Ⅴ価遷移金属Trivalent
actinide
Ⅲ価アクチニド
Tetravalentactinide
Ⅳ価アクチニド
Pentavalentactinide
Ⅴ価アクチニド
Hexavalentactinide
Ⅵ価アクチニド

HalogenハロゲンAnionic
species
陰イオン種

有害物質
Hazardous
substances
Organic※1
有機物
substances*1

海水 *1
Seawater

※2

ホウ酸
Boric
acid

〇
〇
〇
〇
〇
〇
〇
〇
〇
〇※3
〇※3

〇
〇
〇
〇
〇
〇
〇
〇
〇
〇※3
〇※3

●
●
●
■
■
●
×
●
●
〇※3
〇※3

フェロシアン
Ferrocyanides
化合物
●
〇
〇
■
■
●
×
〇
〇
〇※3
〇※3

硫酸塩
Sulfates

炭酸塩
Carbonates

〇
〇
●
■
■
●
×
●
〇
〇※3
〇※3

〇
〇
〇
〇
■
〇
〇
〇
〇
〇※3
〇※3

Hazardous
substances
有害物質

元素

Boric
acid
ホウ酸

Cs
Sr
Ni(Ⅱ)
Sn(Ⅳ)
Nb(Ⅴ)
Am(Ⅲ)
Th(Ⅳ)
Np(Ⅴ)
U(Ⅵ)
I
Se

●
●
●
■
■
●
×
●
●
〇※3
〇※3

フェロシアン
化合物
●
〇
〇
■
■
●
×
〇
〇
〇※3
〇※3

Ferrocyanides*1 Isosaccharinic acid (ISA)*2※1：イソサッカリン酸（ISA）
Ionic strength 0.68 M*3※2：イオン強度0.68M
Set at Kd = 0= 0で設定O:※3：Kd
Data hasbeen set up〇：データ設定済み
Newdata
to be collected or set up in this fiscal (to be continued●:●：本年度新規取得・設定予定（データ不足の場合次年度継続）
next
fiscal if data is insufficient)■：次年度新規取得・設定予定
New data to be collected or set up in the next fiscal■:
x:×：データ取得困難
Data acquisition is difficult：本年度追加調査・設定予定: Additional investigation and settings planned in this fiscal©International Research Institute for Nuclear Decommissioning

175

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Study of sorption impact assessment techniques for each affecting material: Trial parameter setting HCP, pH 13.2 [4]1.0E+03

Example of study on assessment of impact oforganic substances ➢

It was confirmed that evaluation technique 
based on thermodynamic data for all nuclides
leads to more conservative evaluation results thantechnique  based on ideal actual measurement
sorption data.

➢

Meanwhile, it was also confirmed that technique largely depends on thermodynamic data, and that
selection of thermodynamic data and verification of
its reliability are important.

Threshold
しきい値
value

1.0E+01

In addition to the past evaluation cases of trivalentactinide Am targeting the sorption effect ofisosaccharinic acid (ISA), which is said to have thegreatest impact，application and evaluation
targeting tetravalent actinide Th (IV) andhexavalent actinide U (VI) were studied.

➢

(a-1)

1.0E+02

In FY2019, the comparison of the three techniques#
proposed as the techniques for evaluating theSorption Reduction Factor and the evaluation casesof their effectiveness were improved.

Kd (m3/kg)◼

Technique : Evaluated from data on rise in solubility
Technique : Evaluated from thermodynamic
calculations
Technique : Evaluated based on data on impact ofsorption

しきい値

Threshold
しきい値
value
収着低減係数

Sorption
Reduction
収着低減係数
Factor

1.0E-01(b-1)

HCP, pH 13.3 [4]CSH 0.97, pH 13.3 [2]1.0E-02
1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

ISA concentration (M)1.0E-01

1.0E+00

Technique ③: Evaluation based on
actual measurement data

(a-2)

(b-2)

Technique ②: Evaluation based onthermodynamic data

method 2
1.E+06
SRF (Th) - Andra-TDB

SRF [Sorption Reduction Factor]

Reduction Factor

HCP, pH 11.9 [4]Sorption
Reduction
収着低減係数
Factor

1.0E+00

1.E+04(b-3)

SRF (Th) - JAEA-TDB*

1.E+05#Three evaluation techniques for the Sorption

HCP, pH 11.9 [4]HCP, pH 13.2 [4]Th(IV) - ISA(a-3)

1.E+03

Method手法①

1.E+02
1.E+01

Method手法②

Method手法②
Method手法③

Method手法③3
method

1.E+00
1.E-01
1.E-06

1.E-05

1.E-04
ISA concentration [M]1.E-03

1.E-02Comparison of techniquesFig.1 Evaluation results based on each technique and example study of comparison (Left: Th (IV), right: U (VI))©International Research Institute for Nuclear Decommissioning176

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Study of sorption impact assessment techniques for each affecting material: Trial parameter setting Example of study on assessment of
impact of organic substances ◼

➢➢Literature
SKB TR14-10

The typical domestic and overseas
evaluation cases pertaining to the impact ofsorption on various organic substanceswere investigated and organized.
The target organic substances and theirevaluation techniques are organized foreach evaluation case considered in theevaluations in Sweden SKB, Swiss Nagra,
and Japan's Second TRU report.

Nagra NTB 14-08

Each evaluation case mainly studies theevaluation of isosaccharinic acid (ISA).
For other organic substances, if there is
an impact, the ISA value is assigned
conservatively or the actual measurement
data is used to evaluate whether the
impact will become apparent.

⇒ In light of the above, the Sorption
Reduction Factor is set for ISA based on
the actually measured sorption data.
Second TRU
compilation

Technique for assessment of impact
concentration (threshold value)

Technique for assessment of impact of organic substances(Sorption Reduction Factor)

ISA• Set from experimental results, chemical
analogs, etc.
• Consideration to Ca concentration

The Sorption Reduction Factor is set from the experimental
results and chemical analogs. Divide the reference settingsof distribution coefficient by that value to set the distribution
coefficient considering the impact of organic substances.

Organic substances other than
ISA, such as EDTA and NTA

Roughly set from the ISA chemical analog

Concrete admixture and its
deteriorated material

Assuming no potential effect as it does not strongly affect metal ions

Nitrate/ nitrite

Low content, hence not evaluated.

EDTA

None

Sorption Reduction Factor of Ni, Co, Pb is set

CyanideNone

Sorption Reduction Factor of Ni, Co is set

Ammonia, amine

None

Cement deterioration stage and high NH3 concentration
affects Ni, Co

ISA

Set from experimental results, chemical
analogs, etc.

• The Sorption Reduction Factor is set from theexperimental results and chemical analogs. Divide thereference settings of distribution coefficient by that value
to set the distribution coefficient considering the impact oforganic substances.
• A constant Sorption Reduction Factor is set in the range
from a concentration that has no impact to the maximum
concentration expected at the disposal site.

GLU

Same as above

Same as above

Concrete admixture

The concentration in pore water is low and has no impact.

Organic polymer

The impact is negligible.

Bitumen, degradation productsof acidic ion-exchange resin

The concentration in pore water is low and does not affect nuclides other than Ni.

Organic substances such aswaste cloth• Cases evaluated based on experimental
results were referenced.
• The concentration in the filler pore water was
evaluated assuming that all organic
substances contained in Group 2 aredecomposed into ISA by cellulose.

Solvent waste• Thermodynamic data confirmed that DBP does not affect Ni solubility.
• Experimental results confirmed that TBP and DBP do not affect U solubility.

Asphalt solidification

The results of Pu solubility experiments using asphalt leachate indicated absence of impact.

Cement admixture

The distribution coefficient data for Pu in the presence of a water-reducing agent indicated a low impact.

Target organic substances• The Sorption Reduction Factor is set at the ASA
concentration from the setting example of the solubility
increase coefficient.
• The Sorption Reduction Factor is assumed to be equal to
the solubility increase coefficient.

©International Research Institute for Nuclear Decommissioning

177

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Study of sorption impact assessment techniques for each affecting material: Trial parameter setting -Example of study on assessment ofseawater impact
Cs (I)◼

➢➢The information on the assessment of the
impact of seawater in setting the nuclide
sorption parameters for cement-basedmaterials in other countries was investigated.
In the latest case in Switzerland, detailed
studies and parameter settings werecarried out, including the mechanism of thesorption effect of seawater components oneach nuclide and the effects of seawatercomponents on cement alteration andsecondary mineral formation.
A comparison of the trends of increase anddecrease in sorption due to the impact of
seawater in the table confirmed that thetrends of increase and decrease of many
nuclides are different in each report,
reflecting the complexity of the sorption
mechanism of nuclides on cement.

⇒ Here, the Sorption Reduction Factor isset based on the Nagra case (Wieland,2014), which studied the impact of
seawater, including its mechanism,
most systematically.

#The findings that reduce the distribution coefficientare hatched red, and those that increase thecoefficient are hatched blue.

Sr (II)

JAEA-Review 2006-011(Mihara, 2006)

Nagra NTB 14-08(Wieland，2014)

Ochs et al，2016

No impact on measured
valuesSorption Reduction Factor 2 was set.

Decrease in distribution
coefficient
Conflicting sorption with Na,
K, Ca

Sorption Reduction Factor 6 was set.
Conflicting sorption associated withcement alteration-

No impact

No impact

Increase in distribution
coefficient under seawater
system conditions

Ni (II)
Zr (IV)No impact on measured
valuesNb (V)

Increase in distribution
coefficient or no impact
under seawater system
conditions

Am (III)

Increase in distribution
coefficient under seawater
system conditions

A value that is smaller by an order ofmagnitude is set for a salt water
environment
See trivalent actinides
A value that is smaller by an order ofmagnitude is specified in Kd settings(Assuming that the sorption phase in the
cement paste is only C-S-H, thedecrease in sorption due to cement
alteration is taken into consideration)
Very little impact

Th (IV)Np (V)-

A value that is smaller by an order ofmagnitude is set for a salt water
environment
See trivalent actinides

U (VI)

Np is sensitive to salt
concentration
Increase in distribution
coefficient in salt water-I

Decrease in distribution
coefficient under seawater
system conditions
Conflicting sorption withanionsThe distribution coefficient 0 is set.
Conflicting sorption with Cl, no sorption
into C-S-H

Decrease in distribution
coefficient
Conflicting sorption withstable Cl

Se

Decrease in distribution
coefficient under seawater
system conditions

The distribution coefficient 10-4m3/kg is
set considering that a lot of ettringite in
the sorption phase remains even after
alteration.

-©International Research Institute for Nuclear Decommissioning178

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Setting of data collection targets and conditions for specifying the Sorption Reduction Factor ◼ In FY2019, data was collected for the affecting substances, barrier materials, and nuclides to be evaluated that have been
specified so far, and for which no data had been collected or for which data needed to be collected again.
◼ In order to collect the data of sorption in the presence of boric acid, ferrocyanides, and sulfates, a sorption test was
conducted with the concentrations of these affecting substances and liquid conditions (pH, etc.) as parameters.
◼ Considering the actual environment, the nuclides that form the carbonic acid complex were tested under conditions
wherein some carbonic acid was added. (Table 1)

Table 1 Data acquisition conditions
Affecting materials

Boric acid

Solid phase
Test solution

SulfateBentonite
NaCl

pH

Cement

NaCl + Carbonicacid
7, 9.3

Cement

NaCl + Carbonic
acid

NaCl

NaCl

NaCl + Carbonic acid

Cement equilibrium moisture (Approx.
pH12.5)

Cement equilibrium moisture (Approx. pH12.5)

Test parameters

Boric acidconcentration

Boric acidconcentration
Carbonic acidconcentration

Boric acidconcentration
NaCl
concentration

Boric acidconcentration
Carbonic acidconcentration

Sulfate concentration

Sulfate concentration
Carbonic acidconcentration

Target nuclides

Np (V)Am, ThCs, Sr, Ni,

Eu, Np (V), U (VI)Ni

Eu, Np (V)Affecting materials

Ferrocyanides

Solid phase

Bentonite

Cement

Test solution

NaCl

NaCl + Carbonic acid

NaCl

pH

4-12

8.8

Cement equilibrium moisture (Approx. pH12.5)

Test parameters

Ferrocyanide concentration

Ferrocyanide concentration
Carbonic acid concentration

Ferrocyanide concentration

Target nuclides

Cs

Th

Cs, EuKunipia F manufactured by Kunimine Industries is used for bentonite, and OPC Cement paste is used for the cement sample..
©International Research Institute for Nuclear Decommissioning179

【Confidential】(b) Presentation of disposal methods, and development of safety assessment techniques Development of techniques for assessing impact of affecting substances, etc. on disposal- Results of data collection for setting the Sorption Reduction Factor (boric acid) -1E+3

NaHCO31E+1

0 mol/L
ref._0 mol/L

1E+0

0.05 mol/L◼ The effect of reduction in theimpact of boric acid in thepresence of carbonic acid andthe impact assessment
considering the concentration ofcarbonic acid contained in
cement pore water and bentonite
pore water was studied.

y = 0.0004x-2.451

1E-1

1E-2
1E-4

1E-3

1E-2

ref._0.05 mol/L

1E-1

1E+0

Borate concentration [mol/L]

Boric acid concentration

Test results for sorption of U (VI)
into Bentonite in the presence ofboric acid

Test results for sorption of Eu (III), Np(V) into Cement (OPC) in the presenceof boric acid

1E+2

Bentonite*-U(VI)-Borate(pH9.3, 28days)

1E+1

NaHCO3y = 0.4754x-0.557

5E-4 mol/L

1E+0
ref._5E-4 mol/L
0 mol/L
1E-1

ref._0 mol/L*Kunipia F (without calcite)
1E-2
1E-4

1E-3

1E-2

1E-1

Borate concentration [mol/L]

1E+0

Sorption distribution
coefficient

Kd [m3/kg]◼ For the nuclides collected up to
this year, based on their data ofsorption in the presence of boric
acid and carbonic acid, it wasevaluated that the boric acideffect was not significant for
nuclides other than U (VI) (Cs, Sr,
Eu (III), Np (V)).

OPC-U(VI)-Borate(28days)y = 0.0014x-2.783

1E+2

Kd [m3/kg]◼ Since the sorption data from theboric acid coexisting system has
hardly been reported in theprevious studies, the Sorption
Reduction Factor is set based on
the sorption data collected in thisproject so far.

Test results for sorption of Cs, Sr
into Cement (OPC) in the presenceof boric acid

Test results for sorption of U (VI) into
Cement (OPC) in the presence ofboric acid

Sorption distribution
coefficient

Study on assessment of boric acid impact

Boric acid concentration©International Research Institute for Nuclear Decommissioning

180

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Results of data collection for setting the Sorption Reduction Factor (sulfate)Test results for sorption of U (VI) into
Cement (OPC) in the presence offerrocyanides

1E+1

y = 11.796x-0.133

OPC-U(VI)-Ferrocyanide(28days)
1E+0
1E-4

1E-3

1E-2

1E-1

1E+0

Na4[Fe(CN)6] concentration [mol/L]

Ferrocyanide concentration

Test results for sorption of Np (V) into Bentonite in the presence offerrocyanides
1E+0

Kd-Na4[Fe(CN)6] (28day) -C0-

1E+0

Kd-Na4[Fe(CN)6] (28day) -C0-

1E-1

Bentonite*-Np(V)-Ferrocyanide(28days)
pH4*Kunipia F(without calcite)

y = 0.088x-0.297

Kd [m3/kg]◼ Based on the data of sorption in
the presence of ferrocyanides
collected until this year, SRF was
set up in U (VI) and it wasevaluated that the impact offerrocyanide was not significant
for Cs, Sr, Ni.

OPC：Kd-Na4[Fe(CN)6]濃度 (28day)Kd [m3/kg]◼ The sorption data was collectedfocusing on the direct impact offerrocyanide complex formationand the pH dependence of the
impact (wide range of impact
including Bentonite pore water
pH).

1E+2

Kd [m3/kg]◼ Since the sorption data from theferrocyanide coexisting system
has hardly been reported in theprevious studies, the Sorption
Reduction Factor is set based on
the sorption data collected in thisproject so far.

Test results for sorption of Cs, Sr, Ni into
Cement (OPC) in the presence offerrocyanides

Sorption distribution
coefficient

Study on assessment of ferrocyanide
impact

pH8
pH4
pH10
pH8
ref._pH4
pH10
ref._pH8
ref._pH4

y = 0.088x-0.297

1E-1

ref._pH8

1E-2
1E-4
1E-2
1E-4

1E-3

1E-2

1E-1

Na4[Fe(CN)6] concentration [mol/L]
1E-3

1E-2

1E-1

1E+0
1E+0

©International
Research Institute for Nuclear Decommissioning
Na4[Fe(CN)6] concentration[mol/L]181

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Results of data collection for setting the Sorption Reduction Factor (sulfate)Study on assessment of sulfate impact

Test results for sorption of U (VI) into Cement (OPC) in thepresence of sulfates

1E+0

0

10

Kd [m3/kg]

ref.

0.25M
0.025M
0.0025M

0.25M
0.025M
1E+1
0.0025M
0.0M

0.0M

20

30

10
20
30
◼ The data of sorption in the presence 0
0.0M[d]TimeTime [d]of sulfate ions shows a clear time
1E+0
dependence, and a 28-day
test
0
10
20
30
Test
results for
sorption of Np
confirmed that sufficient sorption
Time [d](V) into Bentonite in theequilibrium may not have beenpresence of sulfates: Timereached.
dependence of sorption

Sulfate concentration
dependence (as of 28th
day)

1E+0
1E-4

1E-3

1E-2

1E-1

1E+0

Na2SO4 concentration [mol/L]

Sorption distribution
coefficient◼ From the data at the stage where
sufficient equilibrium was notreached, it was confirmed that U (VI)
was affected, while the impact was
not clear in Np (V).

OPC：Kd-Na2SO4 (28day)

1E+2

Kd [m3/kg]

Kd [m3/kg]

Kd [m3/kg]◼ Previous investigations have
[OPC] Na2SO4
1E+2
suggested that the direct impact of[OPC] Na2SO4
Time dependencecomplex formation with sulfates may
1E+2
not be apparent in many nuclides in
the impact assessment basedon Na2SO4
[OPC]
1E+2
1E+1
thermodynamic data, but sincealmost no data of sorption in the 1E+1
0.25M
presence of sulfates has beenreported, data has been1E+1
collected in
0.025M
this project.
1E+00.0025M

No sulfates

SulfateSulfateSulfateTest duration©International Research Institute for Nuclear Decommissioning

182

(b) Presentation of disposal methods, and development of safety assessment techniques Development of techniques for assessing impact of affecting substances, etc. on disposal【Confidential】- Summary of results of trial parameter setting (Impact threshold and Sorption Reduction Factor) Impact concentration (threshold value) and Sorption Reduction Factor (SRF)
Element group

Representat
ive element

Organic substances (ISA)

SeawaterThreshold value

Boric acidThreshold value

Ferrocyanides
Threshold value

Alkali metal

Cement based

Alkali earth metal
Divalent transition metal

Issues present Issues present

Tetravalent transition metal
Pentavalent transition metal

Trivalent actinide
Tetravalent actinide

Issues present Issues present

Issues present Issues present

Pentavalent actinide
Hexavalent actinide
HalogenAnionic species
Impact concentration (threshold value) and Sorption Reduction Factor (SRF)
Element group

Representative
element

Boric acid

Bentonite based

Threshold value•Set M: mol/L, *2 Kd = 0 (m3 kg-1)•SRF between precipitation conditions => seawaterconditions: Set based on the past information
investigation results: Set based on data collected through sorption
tests (values that are not specified, will be set
hereafter, or will be set after improving thedata.)

Ferrocyanides
Threshold value

Alkali metal
Alkali earth metal
Divalent transition metal
Tetravalent transition metal
Pentavalent transition
metal

Trivalent actinide
Tetravalent actinide
Pentavalent actinide
Hexavalent actinide
HalogenAnionic species

Issues present

Issues present

Issues present

Issues present: Investigation of past information and datacollection not implemented (includes items forwhich absence of investigation results data
has been confirmed)©International Research Institute for Nuclear Decommissioning

183

(b) Presentation of disposal methods, and development of safety assessment techniques 【Confidential】
 Development of techniques for assessing impact of affecting substances, etc. on disposal- Summary and future plans ◼ The data investigation and collection scheduled for this year was completed, the threshold value and SRF were set
according to the existing data, and it was confirmed that the values based on the thermodynamic data are evaluated moreconservatively than the experimental method.
◼ In the next fiscal year, the plan will be studied again based on the evaluation results of this fiscal year and the disposal
exposure dose evaluation results, and the data that is considered to be highly important will be collected, and attempts will
be made to set the SRF.
Table Current evaluations and future challenges with regard to affecting substances
Affectingsubstances
Organic substances

Current evaluation of impact of sorption during disposal

Future challenges

・Regarding the impact of organic substances, the existing knowledge is relatively rich, and the sorption impact data of

・It is essential to study the applicability and validity of evaluation techniques, including

isosaccharinic acid (ISA), which has the greatest impact, is extensive, so at present, it is appropriate to study the evaluation oftechniques based on thermodynamic data, for various combinations of organic

settings based on actually measured data of sorption in the presence of ISA.

substances and nuclides, assuming organic substances and their degradation

・Although data on the impact of other organic substances is inadequate, evaluations based on thermodynamic data may be

products that may be included in accident waste.

applicable.
Seawatercomponent

・As for the effect of seawater components on nuclide sorption, the existing knowledge is relatively extensive, and it is possible to・ Further review of the settings based on the latest knowledge on the sorption

evaluate sorption reduction to some extent quantitatively based on the understanding of sorption mechanism such as sorption

mechanism of each nuclide and the impact assessment of the seawater component is

conflicting with seawater component.
Boric acid solution

Ferrocyanides

required.

・Regarding the impact of boric acid, sorption impact data was collected, and based on the results, it can be evaluated that the・ The challenge is to obtain test data with sufficient accuracy for high sorption nuclides

impact of boric acid is small for alkali metals and alkali earth metals, while boric acid has some impact on actinides. However,such as actinide nuclides, and it is necessary to continue to study the evaluation of

considering the coexistence of carbonic acid, the impact is not significant.

the Sorption Reduction Factor with due consideration given to the relationship・When the boric acid content of waste has a high concentration, it is necessary to properly consider the impact of boric acidbetween the impact of boric acid complex formation and sorption reduction, theconcentration evaluation and nuclide migration around the waste, in addition to the fact that boron is an environmental

evaluation of conflict with other ligands such as carbonic acid complexes, and thestandard material.

consistency of the impact between nuclides.

・For ferrocyanides, data of sorption in the presence of ferrocyanides was collected, and it was confirmed that its impact on alkali

・It is presumed that their impact is complicated depending on conditions such as

metals and alkali earth metals is small, and that there is a possibility that its impact on sorption will become apparent in somenuclides and pH, and it is necessary to expand the data on the relationship between

actinides.

the impact on complex formation due to ferrocyanides and the reduction of sorption,

・In addition to the fact that all cyanides are environmental standard materials, it is necessary to evaluate the behavior of
ferrocyanide itself in the waste and in the disposal environment, and to conduct studies including measures during

to confirm the consistency of the impact between nuclides, and to study theevaluation of the Sorption Reduction Factor.

pretreatment.
Sulfates

・From the evaluation based on thermodynamic data, it was estimated that the sulfate complex had almost no impact on nuclide

・A clear time dependence has been confirmed in the sorption test in the presence ofsorption. On the other hand, in the sorption test of U (VI) to cement-based materials in a sulfate coexisting system, thesulfates, and further data collection and evaluation are required, including the data on

dependence of the distribution coefficient on the sulfate concentration was confirmed at a relatively high sulfate concentration,

the possibility of the impact of solid phase transition.

and clear time dependence of sorption was confirmed.
Carbonates

・Regarding the impact of carbonates, it is a component originally contained in the natural rock and groundwater environment,

・It is necessary to confirm the validity of the evaluation described in the left column.

and its impact assessment technique, which has been examined in conventional studies, can be utilized.
・Even in the absence of carbonates from accident waste, the impact of carbonic acid ion concentration in the pore water is taken
into consideration, and it is believed that it is not necessary to further consider the impact of carbonates contained in accident
waste.

©International Research Institute for Nuclear Decommissioning184

【Confidential】

2. Project detailsc. Characterization efficiency improvement(a) Characterization efficiency improvement① Establishment of a characterization method that combines
evaluation data based on analytical data and migration
models
i.
ii.
iii.

iv.
v.
vi.

Understanding the contamination mechanism
Evaluating the representativeness of analytical data
Investigation and consolidation of the characteristics of analytical
data
Statistical inventory estimation method
Acquisition of analytical data
Evaluation and management of analytical data

② Simplification and speeding-up of analytical methods
i.
ii.
iii.
iv.

Development of sampling technology
Study on streamlining the separation process
Development of automation technology
Establishment of standard analytical techniques(b) Development of sampling technology©International Research Institute for Nuclear Decommissioning

185

【Confidential】(a) Characterization efficiency improvement
Overview of analysis, and R&D using that data
Improving the analysis method

AnalyticaltechniqueAnalysis plan

Analyzing

Development will be
continued targeting
the Okuma facility.

Planning the analysis

Analysis will be continued, data will
be accumulated, the database will
be steadily updated and thesetasks will be continued.

Data

Updating the database

Will be started from thisyear, studies will be
continued and the results
will be reflected in theestimation techniquesequentially.

Waste information

Study of contamination mechanism

Study of representativeness
Studying of the characteristics of waste

Contamination mechanism model

The techniqueincorporating
Bayesian statistics isroughly established.
Further challenges willbe identified andimplementation of
measures will start.

It is difficult to support
the representativenessof existing data. Thisis expected fromfuture plans.

Legend: Status of initiatives
This year’s course of action

Since the standard
analysis planning
method and Bayesianstatistics are expected
to be introduced, themethodology will be
studied.

Will be started from the firstyear,
incorporation into theestimation technique will beconsidered.

Waste model

Establishment of inventory
estimation method
Analysis and evaluation method

Estimated inventory
Study of disposal concepts

Proposed disposal strategies

The inventory will be calculated again
using the improved technique and theresults are planned to be provided as
input for studies to be conducted nextyear.

The nuclide believed to benecessary as part ofanalytical data will beidentified through safety
analysis of disposal..

Information required in the future©International Research Institute for Nuclear Decommissioning186

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization- Flow from generation of radioactive waste to its disposal, and implementation details Contamination

Source term

Storage

ContaminantsMigration of nuclides

Storage

Waste

Generation of waste

Treatment

DisposalTreatment(Stabilization)

Treatment(Waste conditioning)

Burial disposal, etc.

Volumereduction

CementMovement of waste

Reactor Building

Clearance

L3
Damaged fuelTurbine Building

Incineration

Geopolymer

Radwaste Building

L2

Centralized Waste
Treatment Building
Activation product
Structural material
Cooling water, etc.

Structures associated withdecommissioning activities
such as contaminatedwater treatment, etc.

Stored waste

Existing structures

Vitrification(ii)
Representativeness
Data traceability
Dehydration

Other

Mid-level

Geological strata(iii) Characteristics
Properties ofcontamination distribution(i) Understanding the contamination
mechanism
Study of model illustrating the migrationof radioactive nuclides from the source
term to the contaminant.

(v) Acquisition of analytical data
Collection, transportation and analysis of samples(iv) Statistical inventory estimation
Estimation of amount of radioactivity
present in waste(iv) Evaluation and management of analytical data
Recording in database and planning of analysis©International Research Institute for Nuclear Decommissioning187

【Confidential】(ａ) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism◼ Goal To understand the contamination mechanism in order to create amodel that expresses migration behavior of radionuclides.

Legend10-8

Unit 1 R/B

concentration

10-9
235 U Radioactivity◼ Results
 The analytical data of a variety of samples was collected and thecontamination behavior of the nuclides was organized. For example,
U in the contaminated water has a different ratio to Cs dependingon whether it is present in PCV or in the downstream side of PCV(Fig. 1). Also, Np shows a different behavior than U and Pu (Fig. 2).
 Similarly, the concentrations of radionuclides in solidifications andcontaminated water were compared relative to each other with Cs
as reference, and the group of elements with similar behavior werestudied.

Unit 2 R/B

10-10

Unit 3 R/B

10-11

Unit 1 T/B
Unit 2 T/B
Unit 3 T/B
Unit 2 PCVUnit 3 PCVCentralized RW

137 Cs Radioactivity concentration

1.E-01

1.E-02

Np-237/U-236
Pu-238/U-236
Cm-244/U-236

Sampling location

SARRY outlet

KURION outlet

採取場所

HTI/B
basement

1.E-03
Basement ofReactor
Building
Basement of
TurbineBuilding
Centralized RW
basement◼ Indicators to determine goal achievement
 A model that represents the contamination phenomenon should becreated. (FY2019) The model should be reviewed. (FY2020)

Transport ratio (with
輸送比（U-236基準）
as reference)
U-236

237Np分析◼ Implementation details
Fig.
1 Ratioof radioactivity concentration of 235U and 137Cs
３．滞留
水の Study models containing multiple source terms and multiple
236
in a variety of contaminated water samples#1
Power the Fu
mediums with respect to migration of nuclides. Radionuclides have 【 U基準のα核種の輸送比】
1.E+03things that are considered important in waste management. Hence
if data cannot be obtained through analysis, estimate the behavior.
1.E+02
 (FY2019) The elemental processes of contamination was examined,
considering the fact that the phenomenon of contamination is
1.E+01
generally approximated by means of log-normal distribution.
 (FY2020) The study results will be reviewed based on the newly
1.E+00obtained data.

Fig. 2 Ratio of radioactivity concentration of 237Np and 236U
PCV内（2号機及び3号機）：U、Pu及びCmはほぼ同等の輸送比、一方Npは、他のα核種と比較しin a 約一桁高い輸送比を示し、Npは他のα核種に比べvariety of contaminated water samples
溶 解性が高と推察#1 * The 56th Session of Team for Countermeasures for Decommissioning and Contaminated
Water Treatment / Secretariat Meeting, Jul 26, 2018.

タービン建屋以降の滞留(Normalized
水：Puの輸送比は約1/100に減少。原子炉建屋やタービン建屋において
transport ratio in fuel composition)
Puは沈降又は吸着により除去されていると推察©International Research Institute for Nuclear Decommissioning

日本核燃料開発株式会社

188

13

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism
As the contamination status of waste varies widely depending on the material and the place of generation, a “case study” with limited
targets was conducted (Fig. 1).
◼ In the case study, the classification of target waste was studied and 21 types of waste were set up (Table 1). The method for evaluating the"importance" of waste was studied, selection was conducted in 3 stages using this index and the target wastes for this case study wereselected.
 In the first stage, it was narrowed down to 7 types including debris (3 types), secondary waste generated from contaminated water
treatment (2 types), and dismantling waste (2 types).
 Finally, the Reactor Building dismantling waste was selected. Of the Reactor Buildings, Unit 2 with a large amount of analytical data wastargeted. The two possible mechanisms for contamination of the reactor building were considered to be via air and contaminated water,which were suitable for the study. (Fig. 2)◼

Study on the waste
classification methods
Study on the methods forevaluating the importanceof wastes
Setting the evaluation
policySetting of evaluation items
Setting of evaluation
standards
Selection of wastes for thecase study
Evaluation and selection
at Stage 1
Evaluation and selection
at Stage 2

Table 1 Waste classification for the case study
Category 1

DebrisMaterial
Concrete，Metal☑

0.005 - 0.1 [mSv/h]Concrete，Metal☑

0.1 – 1 [mSv/h]Concrete，Metal

1 - 30 [mSv/h]Concrete，Metal> 30 [mSv/h]Concrete，Metal
Burned ash

Felled trees

Burned ash

Contaminated soil

Secondary waste
generated fromcontaminatedwater treatment

Dismantling waste

Stage 2(4) Containment vesselboundary☑

(3) Reactor pressure
vessel boundary☑

(For each adsorbent)

SARRY(For each adsorbent)

Decontaminating device(AREVA)

SludgeALPS slurry

Iron co-precipitation, Carbonate
precipitation

ALPS adsorbent(For each adsorbent)

Mobile type treatment
device(For each adsorbent)

Concentrated liquid waste

SlurryInside the pressure vessel

Concrete，Metal

From well plug torefuelling floor

Contamination
via air
From blowout
panel toenvironment(2) Fuel cladding tube

Contaminated soil
KURION

Inside the containment
vesselInside the Reactor
Building

Case study

Stage 1

< 0.005 [mSv/h]Protective clothing,etc.

Evaluation at Stage 3

Fig. 1 Case study flow

Category 2(5) Reactor Building
boundary(1) Fuel matrix☑

Physical barrier
After narrowing down in☑
☑
stages, finally thecontamination inside theReactor Building wasselected.

Concrete，Metal
Concrete，Metal☑

☑

Other buildings

Concrete，Metal☑

☑

Waste fromdecommissioning work
(including debris retrieval)

Various devices (Metal, etc.)
Debris is likely to have adhered.

Deposits

Deposits

From the suppression
chamber to the basement
Contamination via
water
To water treatment
via the Turbine
Building

Fig. 2 Case study on contamination model(Dismantling waste from inside the Unit 2 Reactor Building)©International Research Institute for Nuclear Decommissioning189

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism (Continued …)◼
◼

For contamination via air inside the Reactor Building of Unit 2, the data on analytical samples with precise collection location was studied (Table 1).
For the transport ratio of major nuclides (relative value to 137Cs, normalized in fuel composition) , similar values are shown for elements with similar
chemical properties (Fig. 1). The change in nuclide composition seemed to be small in the process of migrating through the medium (air). The analysisbased inventory estimation was carried out by reflecting these findings.
 Meanwhile, it is suggested that nuclides (3H, 14C, 60Co, etc.) with a different tendency to correlate with Cs may have a different source term andmigration mechanism than other nuclides.
◼ Regarding local nuclide migration in contaminated areas, measurement data was collected using an imaging plate (Fig. 2). It was found that thecontamination remained on the surface layer.
◼ Radiochemical analytical data provides information on the location of contamination, but does not include direct information on the source terms or nuclide
migration processes. It is necessary to study an inductive approach based on analytical data.

Table 1 Analysis sample from Unit 2 Reactor Building(Radiochemical analysis)
Properties of thesampleNumbe
r ofsamples

Ceiling

Deck plate coating

5

Floor - shield plug east
side

Boring core surface
coating

1

Floor - well plug upper
part (center)

Boring core - lower
layer coating,Concrete - upper andlower layers

3

Floor - northeast side

Same as above

3

3
floor

Floor - north side
(Near RCW pump)

Boring core strippable
paint

1

2nd
floor

Floor - north sidepassage (Near RCW
Hx)

Boring core strippable
paint

1

th

5
floor

rd

Surface (untreated)

After grinding 5 times

After grinding 15 times
Nuclides1
floor

Floor- northwest corner(Personal airlock room
entrance)

Boring core

1

PCV

TIP piping

Pipe blocking material

1stTransport ratio (relative to 137Cs)

Sampling location

1st floor Floor - northwest corner
2nd floor Floor - north side passage (near RCW Hx)
3rd floor Floor - north side (near RCW pump)
5th floor Floor - well plug east side
5th floor Floor - well plug upper side
5th floor Floor - northeast side
5th floor Ceiling
5th floor Ceiling
5th floor Ceiling
5th floor Ceiling
5th floor Ceiling
TIP piping blocking material
RadWaste Building The lowest basement floor (sludge)
Turbine Building The middle basement floor (sludge)
Turbine Building The lowest basement floor (sludge)

Fig. 1 Transport ratio of important nuclides toeach floor of Unit 2 Reactor Building(concentration ratio relative to normalized 137Cs)

Fig. 2 Example ofcontamination distribution databy means of an imaging plate #1(Sample number 3RB-OP-C1)©International Research Institute for Nuclear Decommissioning#1 The 56th Session of Team for Countermeasures for Decommissioning and Contaminated Water Treatment / Secretariat Meeting, Jul 26, 2018.

190

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism (continued...)◼

Regarding the contamination of the Reactor Building via contaminated water, the analytical data is limited to contaminated water. Studies
are under progress with the analytical data of the stagnant water and the solid contents (sludge) contained in it as a clue.
◼ The behavior of actinide nuclides in contaminated water appears to be different for U, Np, and Pu-Am-Cm respectively (Fig. 1).
 Compared to U, migration of Np to contaminated water is likely to be at a higher rate. There are cases where U may be contributed bynatural Uranium suggesting elution from structural materials. In the process of migrating from PCV to the Reactor Building and Turbine
Building, Pu, Am and Cm were lost from water suggesting that these got changed to another form.
◼ U, Pu and Am were detected in the analytical data on sludge contained in stagnant water, and their movement to the sludge was proved
from the mass balance with water (Fig. 2).
◼ Contaminated water is circulated, because of which a certain percentage of radionuclides are continuously flowing in. It is important to study
the process of migration from source terms to water. Moreover, it is necessary to estimate the sedimentation of insoluble actinide.
 Unlike in the case of air, as it is not possible to take an inductive approach, it is necessary to expand and enrich the analytical data(including the acquisition of various types of samples) and refer to findings from other projects as well.

1〜4号機滞留水・スラッジ － 放射能収支（1/2）

1.E+03

Stagnant water (after
滞
留水（ろ過後）filtration)

90%
80%

1.E+0170%60%

1.E+0050%40%

1.E-0110%

Cm-244/U-236

0%
SARRY exit

Pu-238/U-236

KURION exit

20%

HTI/B
basement

1.E-03

Np-237/U-236

Centralized
RW basement

1.E-02

30%

Basement ofReactor
Building
Basement of
TurbineBuilding(based on U-236)
Transport ratio
輸送比（U-236基準）

スラッジ
Sludge100%

1.E+02

Amount ofsludge
スラッジ
contained
含有量(g)(g)0.08750.09080.07680.08750.09080.07680.08750.09080.07680.08750.09080.0768

1号廃棄
2号廃棄
1号廃棄
2号廃棄
1号廃棄
2号廃棄
1号廃棄
物 Unit
物 Unit 33号
物 Unit
物 Unit 33号
物 Unit
物 Unit 33号
物 Unit2号廃棄物 Unit 33号
Unit
1 RadWaste
2 RadWaste
Turbine Unit
1 RadWaste
2 RadWaste
Turbine Unit
1 RadWaste
2 RadWaste
Turbine Unit
1 RadWaste
2 RadWaste
TurbineBuilding
The
Building
The
Building The
Building
The
Building
The
Building The
Building
The
Building
The
Building The
Building
The
Building
The
Building The
処理建屋
処理建屋
タービン建屋
処理建屋
処理建屋
タービン建屋
処理建屋
処理建屋
タービン建屋
処理建屋
処理建屋
タービン建屋lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement lowest basement
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
最地下階
floor
floor
floor
floor
floor
floor
floor
floor
floor
floor
floor
floor
[LI-1WB8-1][LI-2WB8-1]

Cs-137[LI-3TB8-2][LI-1WB8-1] [LI-2WB8-1]

Am-241[LI-3TB8-2][LI-1WB8-1][LI-2WB8-1][LI-3TB8-2]

Pu-238[LI-1WB8-1][LI-2WB8-1][LI-3TB8-2]

U-238

注）代表核種のみ50cm3あたりの放射能量の比を示した。白抜きプロットは検出下限値未満を用 いている。

Sampling
location
採取場所

Fig. 1 Transport ratio of important nuclides to contaminated water(Concentration ratio relative to normalized 236U) #1

図 滞留水（ろ過後）及びスラッジに含まれる核種の収支

Fig. 2 Distribution of Cs and actinide nuclides in➢ α核種の多くはスラッジに含まれている。
137Csの収支は、号機、建屋により異 なる。
➢ 滞留水（ろ過後）及びスラッジの主要な汚染核種である
stagnant water containing
solidifications (sludge)#1➢ スラッジの核種組成は場所によって異 なるので、スラッジと接触する建屋構造
物 等の汚染の核種組成は、一様ではないと見られる。

©International Research Institute for Nuclear Decommissioning#1 The 67th Session of Team for Countermeasures for Decommissioning and Contaminated Water Treatment / Secretariat Meeting, Jun 27, 2019.

191©International Research Institute for Nuclear Decommissioning

(a) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism (continued...)【Confidential】◼

It is desirable to determine the concentration of radionuclides contained in the waste by the simplest method possible. The scaling factor method is used
for difficult-to-measure nuclides contained in waste generated in association with the operation of nuclear reactors. Here, the concentration ratio is
approximated by means of log-normal distribution.
◼ We have assumed that the transport ratio, which is the normalized concentration ratio, can be approximated by means of log-normal distribution, and haveused it for inventory estimation. Cases of log-normal distribution were examined and its applicability to the waste at Fukushima Daiichi Nuclear Power
Station was studied.
◼ The log-normal distribution is used for the distribution with respect to events in a variety of fields. In the field of atmospheric science, the particle size of theaerosol and the concentration of the contaminating substances are expressed by means of log-normal distribution with respect to the contaminating
substances. (Table 1). Moreover, considering the elemental processes of contamination, log-normal distribution is applicable to the formation of particles(solid fine particles and water droplets), the flow of air which is the medium (wind velocity), and the concentration distribution which is the result ofcontamination. (Fig. 1).
◼ Normal distribution is reproducible, and due to its nature, log-normal distribution also has similar reproducibility. It is considered reasonable that particle
size, wind velocity and contamination distribution, all have log-normal distributions. We will continue to study and assume log-normal distribution for
inventory estimation.
Table 1 Various events to which log-normal distribution is applicable #1
Field

Examples where log-normal distribution
appliesCompeting distributions

Event time

Incubation period of infectious disease, illness
recovery time, survival period, years of service,
resale time, talk time, problem solving time, etc.

3 parameter log-normal,
Inverse Gaussian, Weibull,
Gamma, Beta Prime
distribution, etc.

Economics

Stock price, stock earning rate, insurance claim
amount3 parameter log-normal,
Pearson type, inverse
Gaussian, etc.

Industrial

Survival time distribution related to theapplications of thermal ion electron tube system,
particle size suppression, metal fatigue, etc.

Biology

Concentration of blood components, lowest
lethal dose

General gamma, Dirichlet,
multivariate gamma

Ecology

Population

Negative binomial distribution,
logarithmic series distribution

Atmosphericscience

Aerosolparticle
size,
concentration
contaminating substances, flood frequency

Geology

Particle size

Particle massofWind velocity

Contamination concentration

Extreme value, Johnson,
Weibull, Beta
Multivariate normal distribution(for spatial correlation)

Fig. 1 Log-normal distribution associated with radionuclide contamination #2,3

#1 Kunio Shimizu, Applied Statistics, 17 (1), 55–59 (1988). #2 H. Neustadter et al., EPA-650/4-74-038 (1974) . #3 Jun Hiroo et al., Journal of the Air Pollution Control Association, 14 (8), 311 (1979).
©International Research Institute for Nuclear Decommissioning192

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationi. Understanding the contamination mechanism (continued...)◼ The study of contamination mechanism focuses mainly on inductive approach based on radiochemical analytical data. Based on
the gradual accumulation of analytical data, the study will be continued by referring to the findings from analytical andexperimental approaches and by combining them. (Fig. 1)
 Along with the targets of contamination, limited nuclides (as a standard 137Cs, which is easy to handle and 14C and 129I
which are important actinides in disposal safety assessment) will be studied .
The validity of lognormal distribution willcontinue to be studied.

Findings and accident analysis
techniques from existing safety research
will be used for studying formation ofparticles and elution into water.

Source term

Migration process

Fixing of contamination
Waste A

Particle formation

B
Damaged fuelRadionuclide

Radionuclide

C

Transport in medium
Air, water

D
Activation productChemical reaction
Decontamination etc.

E

Stored wastes

Effect with waste:
Waste Z

Fallout / natural products

In addition to damaged fuel(fuel and cladding), the effectsof activation products are also
considered.

The diffusion in the medium
brings in the findings of heatflow analysis.

The state of contamination is studied
by considering waste as a group.
(Refer to the section “Characteristics
of waste”)

Fig. 1 Direction of the study on contamination mechanism©International Research Institute for Nuclear Decommissioning193

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
ii. Evaluating the representativeness of analytical data

Quantity (m3)◼ Results
 We have started digitization (creation of database) of information onradioactive waste that is being stored and managed, and havealready accumulated about 60,000 units of data. (Fig. 1)
 Meanwhile, analytical data on various samples has been
accumulated, and it was found that the concentration can be
expressed by means of log-normal distribution. (Fig. 2)◼ Goal To find a way to evaluate the representativeness of the dataobtained through analysis.

◼ Indicators to determine goal achievement
 The waste information should be organized, and the waste
characteristics should be evaluated. (FY2019)
 A method to indicate the position of analytical data with respect towaste information should be presented. (FY2020)

Fig. 1 An example of identification of information on stored waste

Transport ratio []◼ Implementation details
 Group the forms of waste generation into several categories andstudy the methods for estimating the inventory for each of them.
 (FY2019) Regarding contamination distribution of waste, theprocesses from generation to accumulation and storage wereassumed and the changes in contamination distribution werestudied based on the analytical data. The usefulness of Bayes
estimation to statistically study the subjects for which the parent
population has not been determined was investigated.
 (FY2020) The study results will be reviewed based on the newly
obtained data. A statistical method for inventory estimation will be
studied based on the results of the previous year. In addition, theapplicability of the method will be evaluated using analytical data.

Arrangementof frequency
distribution

Fig. 2 Study on statistical handling methods of analytical data
(Example of rate of release into the air inside the building)©International Research Institute for Nuclear Decommissioning

194

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
ii. Evaluating the representativeness of analytical data
◼ To estimate the relationship between the analytical
Table 1 Storage area information and waste traceability (as of August 30, 2019)
samples and waste, the storage and management
Surface dose rate
Area
status at the site is collated with the analytical database.
Area
Storage methods
Stored item
classification
occupancy
Traceability(mSv/h)rate(Fig. 1)
B
Open storage
Various debris
0.1 or less
95% No◼ It was found that the storage areas with traceability are
COpen storage
Various debris
0.1 or less
99% NoF2Open storage
Various debris
0.1 or less
85% Nolimited to "Area Q" and "solid waste storage vault”.
J
Open storage
Various debris
0.1 or less
78% No(Table. 1)
N
Open storage
Various debris
0.1 or less
96% NoO
Open storage
Various debris
0.1 or less
83% No The debris that was present in "Area Q" has already
P1
Open storage
Various debris
0.1 or less
80% Nobeen transferred to the solid wastes storage vault, andU
Open storage
Various debris
0.1 or less
100% No
V
Open storage
Various debris
0.1 or less
80% Noas a result, the only traceable storage area at present
AAOpen storage
Various debris
0.1 or less
26% Nois the "solid wastes storage vault”.
D
Covering with sheet
Various debris
0.1 - 1
58% No◼ The representativeness of the analytical data must be
E1
Covering with sheet
Various debris
0.1 - 1
88% Nostudied in accordance with the traceability of the waste.
P2
Covering with sheet
Various debris
0.1 - 1
62% NoEstimating the distribution position of analysis samples in theparent population.

Waste information database

Organizing
information on
stored wastes

analytical database

Estimating thecontrast relationshipbetween analyticalsamples and waste

Storage and management status at the site
Fig. 1W
X

Covering with sheet
Covering with sheet

Various debris
Various debris

0.1 - 1
0.1 - 1

27% No65% NoL

Soil cover type
temporary storage
facility

Various debris

1 - 30100% NoA

Temporary storage
facility

Various debris

1 - 3014% NoE2Storage in containers Various debris

1 - 3026% No1F

Storage in containers Various debris

1 - 3099% NoQ

Storage in containers Various debris

1 - 300%solid wastes
storage vault

Storage in containers Various debris―

36% Yes

G

Open storage

felled trees, trunk, roots,―
branches and leaves

63% NoH

Open storage

felled trees, trunk, roots,―
branches and leaves

74% NoOpen storage

felled trees, trunk, roots,―
branches and leaves

88% NoV

Open storage

felled trees, trunk, roots,―
branches and leaves

2% No

G

Temporarytankstorage felled trees, branches andleaves

―88% NoT

Temporarytankstorage felled trees, branches and―leaves

94% NoOpen storage

Used protective clothing―

81% No

M

Small
areas

characterYes, depending on
the item

References of waste data regarding the representativeness of analytical data
©International Research Institute for Nuclear Decommissioning

195

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
ii. Evaluating the representativeness of analytical data
◼ Frequency of dose rates in the waste storage data and analytical data was compared focusing on the areas in Units 1 to 4 andthe indoor environment. (Fig. 1)
 Waste storage data contained a lot of combustible / flame-retardant materials, metal waste, and felled trees with a low dose,and concrete with ≥30 mSv/h. The analytical data contained a lot of concrete with its dose exceeding 5 μSV/h.
◼ The analytical samples thus illustrate characteristic contamination rather than being just a simple representation of distribution.
◼ Based on these facts, the study on the representativeness and the method for evaluating it will be continued.

25

1000

600

伐採木
Felledtrees
土砂
Sediments

400

その他
Other
Mixed
wastes
混合廃棄物

200

Number of cases (number)Concreteor asphalt
コンクリート・waste
アスファルトガラ
Equipment/ Control
機器類・制御盤類
panelsCombustible / flame可燃・難燃物
retardant
substances

800

Number of cases (number)

金属
Metal
コンクリートConcrete
Other
その他

金属ガラ
Metal
waste
20

15

10

5

Unknown
不明

0

0
<0.1

0.1～1

1～5

5～30

不明
Unknown30≦<1

1～5 5～30 30≦ Unknown
不明

Dose (μSv/h)

Dose (μSv/h)
Waste storage data

Analytical data

Fig. 1 Indoor dose of debris in areas of Units 1 to 4
©International Research Institute for Nuclear Decommissioning196

【Confidential】◼ Goal To establish a method for statistically estimating the amount ofradioactivity (inventory) contained in waste.
◼ Implementation details
 Group the forms of waste generation into several categories andstudy the methods for estimating the inventory for each of them.
 (FY2019) The usefulness of Bayes estimation to statistically study
the subjects for which the parent population has not been
determined, was investigated. The processes from generation toaccumulation and storage were assumed and the changes incontamination distribution were studied based on the analytical data.
 (FY2020) The study results will be reviewed based on the newly
obtained data. A statistical method for inventory estimation will be
studied based on the results of the previous year. In addition, theapplicability of the method will be evaluated using analytical data.

◼ Indicators to determine goal achievement
 The characteristics of the analytical data should be organized andshown. The statistical methods suitable for application should be
identified (FY2019)
 Study results should be reviewed. A statistical method for inventory
estimation should be proposed. The applicability should be
evaluated and shown. (FY2020)

Probability mass

◼ Results
 It was found that the migration rate of nuclides can be estimatedfrom the analytical data by expressing it as probability distribution
based on Bayesian statistics. (Fig. 1)
 Although the distribution parameters (mean and variance) changewith the accumulation of analytical data, it was found that
improvements may not be seen depending on the type of nuclide
and waste, and hence it is necessary to improve the model andtreatment methods. (Fig. 2).

Band representation Statistical representation(a) Characterization efficiency improvement① Establishment of method for characterizationiii. Investigation and consolidation of the characteristics of analytical data, iv. Statistical inventory
estimation method

Ratio of release within the building
Alkali metal nuclides

Uncertainty case(minimum)
Reference case(minimum)
Reference case
Reference case(maximum)Uncertainty case(maximum)

Ratio of release within the building
Lanthanide and actinide nuclide

Fig. 1 Nuclide migration ratio obtained with the help of Monte Carlocalculations based on Bayesian statistics (expressed in terms ofprobability density distribution)

Average (μ)

Variance (σ)

Fig. 2 Changes in log-normal distribution parameters with increase in data
(Transport ratio of alkaline earth metal elements to the Reactor Building)©International Research Institute for Nuclear Decommissioning

197

(a) Characterization efficiency improvement① Establishment of method for characterization【Confidential】

iii. Investigation and consolidation of the characteristics of analytical data
◼ Characteristics such as dose rate are being investigated based on waste information in order to clearly specify thecharacteristics of waste that is currently stored.
◼ As there are relatively high number of records of stored waste generated in and around Units 1 to 4, it is expectedthat by understanding the contamination of the Reactor Building, the characteristics of the waste that is currently
stored can be roughly understood.
◼ It appears that waste unevenly exists in the container, and this was investigated from the dose rate data of thecontainer. (Fig. 1) Variations in data measured from multiple directions suggested that there had to be specificdistribution.
◼ The relationship between the dose and quantity of waste stored in the facility was investigated (Fig. 2). Although there
were variations in the blocks in which waste was stored, as a whole it was found that there was specific distribution.
◼ Therefore, when the variation in the amount of waste and the dose rate is considered as a group, it was found that the
distribution of each could be discussed.
◼ Therefore, the state of waste accumulation will be modeled in three stages (individual waste, stored in a container,and stored in the facility), and an attempt will be made to study and quantify the characteristics of contamination
distribution for each of them. (Fig. 3)

QuantityFrequency

1 block
5 blocks
6 blocks
7 blocks
All blocks

Dose

Generally, thesurface iscontaminated, andthe distribution is
uneven.

Solid wastes

Various solidifications
with different levels ofcontamination are
aggregated, so
distribution is uneven.

Storage container

Various containers withdifferent levels ofcontamination are
aggregated, so
distribution is uneven.

Storage facility

Standard deviation

Fig. 1 Uneven distribution of dose rates inhigh-dose debris containers (variation inmeasurement values from multiple directions)Fig. 2 Amount of debris storedin solid wastes storage vault
No.8 (B2F)

Fig. 3 Hierarchical structure related towaste accumulation and variations incontamination distribution©International Research Institute for Nuclear Decommissioning198

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
iv. Statistical inventory estimation method

50%ileCredible interval(Bayesian confidence
interval)(5 - 95%ile)

Data density

LN transport
rate

Analysis of transport ratio using theBayesian statistical technique(2.5 - 97.5%ile)

Release rate into the air inside the building◼

Probability mass◼

A method to set the parameters of the analytical nuclide migration model based on the continuously accumulated analytical data wasestablished. (Fig. 1)
In this method, the distribution of relative and normalized concentrations (transport ratios) is analyzed based on Bayesian statistics, and theuncertainty of the parameters (mean and standard deviation) is quantitatively represented. Thus, the migration rate can be represented as aprobability density distribution. Moreover, as a result, the certainty of the estimated inventory can be shown together with radioactivity.
Along with the accumulation of analytical data, the changes in the certainty of parameters can be investigated and used in the analysis plan.

Data frequency◼

Statistical setting― Maximum value―○――

95 %tile
50%ile5%tileMinimum value

Release rate inside
the building

Statistical setting of release rate
into the air inside the building
Release rate into the air inside thebuilding for each element group

Fig. 1 Flow of setting migration parameters based on Bayesian statistics©International Research Institute for Nuclear Decommissioning

199

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
iv. Statistical inventory estimation method

◼

◼

97.5%
97.5%
75%

75%

25%

25%

2.5%

Logarithmic average of transport ratio

2.5%

Bayesian estimation◼

The bootstrap method and the jackknife method, which are methods for frequency-wise parameter interval estimation, were studied in order to study thevalidity of introducing the Bayesian statistical methods, and as the bootstrap method was found to be more suitable when the amount of data is less, its
probability distribution parameters were compared with the Bayesian statistical method. (Fig. 1)
While applying these to the inventory estimation of the accident waste at Fukushima Daiichi, currently, there is a restriction, in that the amount of data thatcan be referenced is limited to about 10 to several dozen units. When the number of data is less, the methods based on frequency-based statistics havedisadvantages such as the possibility of giving smaller confidence intervals than needed, requiring verification by simulation, etc.
If the form of distribution of the target is known, the Bayesian statistical method can estimate the parameters even when there is little data. There is boththeoretical and practical experience, that this method enables parameter estimation even in situations where there is little actual data such as failure rates
in probabilistic risk assessment of nuclear power plants. In addition, it is possible to visualize the process of updating confidence intervals and improving
the extent of confidence by sequentially adding data.
From these, its applicability to the inventory estimation of accident waste in Fukushima Daiichi is considered high.

Bayesian estimation◼

Logarithmic standard deviation of transport ratio

Fig. 1 Comparison of statistical estimation results of probability distribution parameters (rate of semi-volatile nuclides released into the air inside the building)©International Research Institute for Nuclear Decommissioning200

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterization
iv. Statistical inventory estimation method◼ Using the results of the study, the inventory of various wastes was recalculated (Fig. 1). These results can beused for the study of disposal concepts in the next fiscal year.
Statistical estimation
Maximum value

Nuclide dose

Minimum value*

Estimation of FY2018

Maximum
value

Minimum
value

Fig. 1 Example of statistical inventory estimation results (Total radioactivity of KURION adsorption tower used for secondary waste generated fromcontaminated water treatment)©International Research Institute for Nuclear Decommissioning201

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data
◼ Results
 Analytical samples were either collected or obtained fromTEPCO, which were then transported and analyzed (Fig.
1). The obtained analytical data was publishedsubsequently (Fig. 2) and a database was created andmade available for test.

Chelating resin 1
outletChelating resin 2
outletActivated carbon
outletTitanate 2 outlet

Cerium oxide
outletSilver zeolite
outletTitanium oxide
outletCerium oxide
outletTitanate 1 outlet

Note) Shows the radioactivity concentration corrected forattenuation on Mar 11, 2011.
The plot points with outline only, indicate value below thelower detection limit.

Carbonate
precipitation
outletActivated carbon
outlet◼ Indicators to determine goal achievement
 Analytical samples should be collected, stored andmanaged, transported, and analyzed. Analytical data
should published on an ongoing basis. Missing analytical
data and samples should be indicated. (FY2019, FY2020)

Inlet (raw water)◼ Implementation details
 The samples for analysis were either collected or obtained
from TEPCO. The obtained analytical samples were stored
and managed in cooperation with TEPCO and weretransported to the analytical facility. Samples wereanalyzed at a licensed facility. (FY2019, FY2020) The obtained analytical data was published subsequently
and made available for test via the Internet. In addition, thedata that must be further expanded and enriched wasstudied in consideration of the accumulated data and theprediction on waste generation. (FY2019, FY2020)

Fig. 1 Collection of samples of secondary waste generated
from contaminated water treatment from the storage cask (HIC)Radioactivity concentration (Bq/L)◼ Goal To accumulate analytical data to establish a method forcharacterization.

Fig. 2 Radioactivity concentration at the outlet of the main equipment in
the multi-nuclide removal facility (expanded B System) #１#1 The 60th Session of Team for Countermeasures for Decommissioning and Contaminated Water Treatment / Secretariat Meeting, Nov 29, 2018.
©International Research Institute for Nuclear Decommissioning202

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data

◼ The analytical samples of contaminated secondary waste generated from contaminated water treatment were collected from thefacilities and from stored containers, respectively.
◼ Processed water from the cesium adsorption facility (SARRY II) and ALPS adsorbent were collected.
 The used adsorption towers were collected during the transfer and discharge to high-integrity containers (HIC). As the amountof contaminated water decreased, there was a decrease in the delivery frequency of adsorbents and hence samples werecollected from only 3 towers. (Table. 1)◼ The samples that were collected from HICs that were generated before April 2016 were studied in FY2018, and 29 HICs wereselected as sampling targets. In FY2018, 12 out of 29 units were sampled, and the remaining 17 units were sampled in FY2019.
(Table. 2)
Table. 1 Collection results of treated water andsecondary wastes generated from contaminatedwater treatment in ALPSClassification

Treatedwater

Adsorbent

Sampling date

Sample nameDec 3, 2019

SARRY II

Jan 30, 2020"

Mar 17, 2020"

Aug 26, 2019

ALPS III chelating
resin 1
ALPS III chelating
resin 2
ALPS I Titanate1

Sep 12, 2019
Nov 29, 2019

Table. 2 Collection results of samples that are being stored inside HIC
Facility

ALPS I

ALPS III

Maximum dose rate [mSv/h]
Surface of sample
HIC sampling
containerhole(about 10 ml)
γ
γ+β
γ
γ+β
0.7
60
0.0035
1.7
3.5
210
0.025
8.0
3.0
300
0.020
4.5
1.2
120
0.035
13
2.8
200
0.024
10
2.5
160
0.055
15
3.5
150
0.013
5.5
0.6
40
0.035
13
2.5
260
0.025
14
2.5
100
0.060
15
7.0
400
0.045
23
4.0
300
0.080
22

Name ofsampleSampling date

Numbe
r of
samples

Iron
coprecipit
ation
slurryDec 18, 2019
Dec 19, 2019
Dec 20, 2019
Dec 23, 2019
Dec 24, 2019
Dec 25, 2019
Jan 9, 2020
Jan 10, 2020
Jan 14, 2020
Jan 17, 2020
Jan 20, 2020
Jan 24, 2020

1
1
1
1
1
1
1
1
1
1
1
1

Side
surface of
HIC
γ
0.535
3.114
3.430
4.000
4.550
1.859
1.237
1.462
2.658
4.240
3.430
4.550

Carbonate slurryJan 23, 2020

1

6.583

3.0

180

0.75

220

Carbonate slurryJan 15, 2020
Jan 21, 2020
Jan 22, 2020

1
1
1

7.873
9.547
5.305

15
15
2.5

800
1000
170

1.1
0.80
0.40

350
350
130

Chelatingresin 2

Jan 27, 2020

3

0.018

0.002

0.1

0.00050.08©International Research Institute for Nuclear Decommissioning203

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data◼

Debris from Unit 2 RB (this report)
Debris from Unit 3 RB (this report)
Debris from Unit 1 RB (existing report)
Debris from Unit 2 RB (existing report)
Debris from Unit 3 RB (existing report)

137Cs Radioactivity
concentration (Bq/g)

TIP piping blocking
material
PCV sediments

Debris from Unit 1 RB (this report)

Debris from Unit 2 RB (this report)
Debris from Unit 1 RB (existing report)
Debris from Unit 2 RB (existing report)
Debris from Unit 3 RB (existing report)

Plot points with outline only, indicate
values below the lower detection limit

137Cs Radioactivity
concentration (Bq/g)

Radioactivity ratio with reference to 137Cs

Debris from Unit 1 RB (this report)

238Pu Radioactivity concentration (Bq/g)

90Sr Radioactivity concentration (Bq/g)

During the 65th Session of the Team for Countermeasures for Decommissioning and Contaminated Water Treatment / Secretariat Meeting(April 25, 2019), the results of analyzing debris samples (floor boring core and strippable paint accompanying decontamination) collectedfrom inside the Reactor Buildings in Units 1, 2 and 3 (Fig. 1), containment vessel sediment deposit samples from Unit 1 and TIP pipe
blocking material samples from Unit 2 (Fig. 2), were reported.

Nuclides➢ Regarding the 90Sr/137Cs ratio, the ratio tends to be high in the samples collected from the 1st floor of Unit 1(near the through hole X6) and the 1st floor of Unit 3.
➢ Regarding the 238Pu/137Cs ratio, the ratio tends to be high in the samples collected from the 1st floor of Unit 1(near the through hole X6) and the upper floor of the well plug on the 5th floor of Unit 2.
➢ For 90Sr and 238Pu, it was suggested that there were places where there was a difference in contamination
against Cs.

➢ The ratio of radioactivity based on 137Cs isremarkably high for 55Fe, 60Co and 63Ni in the TIP
pipe blocking materials. Meanwhile, it is extremely
low for actinide nuclides such as U and Pu.

Fig. 1 Nuclide concentrations detected in debris samples (floor boring core andstrippable paint accompanying decontamination) collected from inside theReactor Building (RB) (plot against 137Cs)

Fig. 2 Concentration of nuclides detected in thesediments of Unit 1 containment vessel(PCV) and Unit 2 TIP pipe blocking
material (ratio against 137Cs)➢ Even inside the containment vessel, the tendency ofcontamination varies greatly from place to place.

204©International
Research
Institute for
Nuclear(63Decommissioningrd session), Feb 28, 2019
# Waste sample analysis results, Team for Countermeasures for Decommissioning and Contaminated
Water Treatment/ Secretariat
Meeting【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data
◼ During the 67th Session of the Team for Countermeasures for Decommissioning and Contaminated Water Treatment /Secretariat Meeting (Jun 27, 2019), the results of analyzing stagnant water collected from Unit 4 R/B basement, Units 2 to 4
T/B basement and Units 1 to 4 Rw/B basement, and the sludge contained in it, were reported. (Fig. 1)
 A tendency for the sludge to be enriched with U and Pu nuclides was seen.
In addition, the analysis results of 237Np and 129I werereported for the stagnant water and treated water, forwhich the analysis results had already been reported.
○ Regarding 237Np, 18 samples including stagnant waterin Unit 1 Reactor Building, water near the water
surface of Unit 3 PCV, and water from the outlet ofSARRY, etc., were analyzed, and the detectedconcentration of 237Np was less than 9×10-3 Bq/cm3,
which was the regulatory concentration limit#. The
concentration in the water near the water surface ofUnit 3 PCV was 2.5×10-3 Bq/cm3, which was thehighest. Other detected concentrations were to theorder of 10-4 to 10-5 Bq/cm3.
○ From amongst the 47 analytical samples including
stagnant water in Unit 1 Reactor Building, water fromthe outlet of KURION, etc., 129I was detected in 1
sample of the water from the KURION outlet and 2
samples of the water from the AREVA outlet.
Radioactivity concentration was 1.6 x 10-1 Bq/cm3 forthe water from the KURION outlet and 2.7 x 10-1 Bq/cm3,
8.5 x 10–2 Bq/cm3 for the water from the AREVA outlet.

SludgeStagnant water(after filtration)

Radioactivity
concentration◼

# Concentration limit in the water outside the surrounding monitoredarea as stipulated in the “Notice defining the dose limit based on theprovisions of the Ministerial Ordinance for Commercial Nuclear
Power Reactors concerning the Installation, Operation, etc.”

Lowest basement floor in Unit 1 RadWaste Building

Lowest basement floor in Unit 2 RadWaste Building
Turbine Building

Lowest basement floor in Unit 3Note) Radioactivity concentration was calculated by dividing by the volume of stagnant water
before filtration. The plot points with only outline represent values that are below thedetection limit. Radioactivity concentration was corrected for attenuation on March 11, 2011.

➢ Many α-nuclides were detected in the sludge from the stagnant water (after filtration).
The α-nuclide concentration in the sludge of Unit 3 Turbine Building, in which 154Eu
was detected, tends to be higher than that in the Unit 1 and Unit 2 RadWaste
Buildings.
➢ 94Nb, 152Eu were not detected in any of the samples.

Fig. 1 Concentration of nuclides detected in the stagnant water (after
filtration) and sludge collected from the basement floor of each building

205©International
Research
Institute for
Nuclear(67Decommissioning
th session), Jun 27, 2019
# Waste sample analysis results, Team for Countermeasures for Decommissioning and Contaminated
Water Treatment/ Secretariat
Meeting【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data◼

The boring core on the outer wall of the Unit 2 Reactor Building was analyzed. These samples will help in finding the distribution ofcontamination present inside and outside the wall.
 Boring core samples were collected from two places in the upper part (around 5,300 mm in height) and two places in the middle part(around 3,000 mm in height) of the western side outer wall opening (width 5,230 mm x height 6,650 mm).
 With these analytical samples, the waterproofing material from the exterior of the boring core, and the outer side wall and the inner sidewall of the concrete part were analyzed, respectively. (Fig. 1, 2)◼ The ratio of 90Sr/137Cs was similar to the values reported so far. The ratio of 238Pu/137Cs was about an order of magnitude larger than thevalues reported so far. (Fig. 3)◼ This is the data on the local contamination distribution of Unit 2, and the distribution will be studied together with other data (rooftop, ceilingback, floor, etc.).
10000
Waterproofing material in outer wall
1.00E+07

4
1.00E+04

10

10-2
10-3

are below the lower detection limit

10-4
1.00E+03

1.00E+02
2

10

1.00E+01

1.00E+00
010

1.00E-01
1.00E+00 1.00E+01
1 1.00E+02 1.00E+03
3 1.00E+04 1.00E+05
5 1.00E+06 1.00E+07
7 1.00E+08

10

10

10

137Cs Radioactivity
137Cs放射能濃度（Bq/g）

concentration (Bq/g)

Sr-90
Fig. 2 External appearance of analytical sample(outer wall waterproofing material)

Debris from Unit 2 RB (this2号機RB瓦礫（本報告）report)Pu放射能濃度（Bq/g）

1.00E+05

10-1

Debris from Unit 2 RB (this2号機RB瓦礫（本報告）report)1号機RB瓦礫（既報）Debrisfrom Unit 1 RB (existing
Sr-90/Cs
report)Debrisfrom Unit 2 RB (existing
2号機RB瓦礫（既報）
report)Debrisfrom Unit 3 RB3(existing
3号機RB瓦礫（既報）10report)白抜きは検出下限値以下
Plot points with only outline indicate values that

10

238Pu Radioactivity
238
concentration (Bq/g)

Fig. 1 Overview of analytical sample

1090Sr放射能濃度（Bq/g）

Concrete core

1.00E+06
6

90Sr Radioactivity concentration (Bq/g)

Concrete in outerwall side
Concrete in innerwall side

1000
3
10

Debrisfrom Unit 1 RB (existing
1号機RB瓦礫（既報）
report)Debrisfrom Unit 2 RB (existing
2号機RB瓦礫（既報）
report)Debrisfrom Unit 3 RB3 (existingPu-238/Cs
3号機RB瓦礫（既報）
report)100

10

白抜きは検出下限値以下

Plot points with only outline indicate values that are
below the lower detection limit

1
1010
1

-1
100.1
0.01

0.001
-3
10

10-6

10-5

10-7 10-80.0001
1

101
10

100

10003
10

10000

7
1000005 1000000 10000000100000000

1010137Cs Radioactivity
137Cs放射能濃度（Bq/g）

concentration (Bq/g)

Pu-238Fig. 3 Concentration of nuclides detected in the sample fromthe western outer wall of the Unit 2 Reactor Building(Plotted with reference to 137Cs)©International Research Institute for Nuclear Decommissioning

206

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data◼

◼
◼

The used adsorbents, which are the secondary waste generated after contaminated water treatment from the multi-nuclide removal facility(ALPS), contain different nuclides depending on their types. Hence radioactivity analysis is important in order to study their treatment anddisposal methods. The absorbents were analyzed for the first time.
For activated carbon adsorbents, the analysis method was studied in advance. Due to the low concentration of radioactive Cs nuclides, γ ray analysis was performed directly without removing the Cs nuclides. For α-ray and β-ray analysis, a method of treatment in which a solid
sample is dissolved in acid, etc. to bring it in the liquid state (making a solution), was studied and applied. (Fig. 1)
As a result of the analysis, 60Co, 106Ru, 125Sb and 137Cs were detected and quantified (Table. 1). 54Mn, 90Sr, 94Nb, 134Cs, 152Eu, 154Eu, 238Pu,
239Pu+240
Pu, 241Am and 244Cm were not detected. From these facts, the nuclides that can be captured in the final stage of the process
became clear.
Sample<making a solution>

γ ray analysis
Measurement (Ge semiconductor detector)
54Mn, 60Co, 94Nb,

Thermaldecomposition

106Ru, 125Sb,

120 ℃ for 40 minutes
700 ℃ for 120 minutes

Table. 1 Concentration of radionuclide detected in activated carbon adsorbent
Radioactivity concentration 〔Bq/g〕

Sodium percarbonate
Sodium carbonate

134,137Cs, 152,154Eu

Alkali melting

800 ℃ for 30 minutes

Demineralized water
5 M nitric acid
Concentrated nitricacid

Filtration

0.1 μmMembrane filter

Chemical treatment

Chemical treatment

β Ray analysis

α Ray analysis• Measurement (gas flow type low
background β ray measuring device)
90Sr• Measurement (Si semiconductor detector)

Name ofsample

60

Co

106

Ru

125

Sb

137

Cs

(About 5.3
years)(About 374
days)(About 2.8
years)(About 30
years)ADC-AAL8-2(2.8 ± 0.1) ×101(2.8 ± 0.3) ×102(1.3 ± 0.1) ×
101< 6 × 10-2

ADC-AAL8-3(1.6 ± 0.1) ×101(2.1 ± 0.2) ×102(6.9 ± 0.3) ×
100< 5 × 10-2

ADC-EAL9-1(3.1 ± 0.1) ×101(3.8 ± 0.3) ×102(1.6 ± 0.1) ×101(7.8 ± 1.9) ×
10-2

238Pu, 239+240Pu, 241Am, 244Cm

Fig. 1 Analysis flow of activated carbon adsorbents©International Research Institute for Nuclear Decommissioning207

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data
◼ Current analytical samples

◼

Items

Analysis facility

Samples

Debris/contaminated water,
etc.

JAEA (Nuclear
Science ResearchInstitute)

Debris from concrete core on the west wall of Unit 2 Reactor Building refueling
floor

Secondary waste generated
from contaminated water
treatment

JAEA (Nuclear
Fuel Cycle
Engineering
Laboratories)

Stagnant water that contains building sludge, adsorbents used in multi-nuclide
removal system (Cerium oxide, activated carbon, silver zeolite etc.)

Contamination distribution,
etc.

JAEA (Oarai
Research andDevelopment
Institute)

Debris from cover soil temporary storage facility

Sludge, debris etc. frombuildings

NDC

Debris from buildings, debris from cover soil temporary storage facility, stagnant
water containing building sludge

Contaminants from PCV

NFD

Unit 1 PCV deposits, Unit 2 TIP piping blockage

Results of transportation of analytical samples from Fukushima Daiichi NPS to various analysis facilities
Transportat
ionfrequency

Date oftransport

Destination

Analytical samples

Transport 1

9/26/2019①JAEA (Nuclear Science Research Institute), ②
NDC① Stagnant water, treated water ②Debris

Transport 2

12/19/2019①JAEA (Nuclear Fuel Cycle Engineering
Laboratories), ②NFD

① Slurry from multi-nuclide removal system② Sludge from buildings

Transport 3

2/19/2020①JAEA (Nuclear Science Research Institute), ②
JAEA (Oarai)① Debris ② Debris©International Research Institute for Nuclear Decommissioning

208

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationv. Acquisition of analytical data
◼ Continuous implementation of operations on the database FRAnDLi (Fukushima Daiichi Radwaste Analytical Data Library) (Fig. 1).
 Relevant data released by TEPCO is also compiled in addition to the data acquired in this project (Table 1).
◼ The statistical data on accesses showed that there is constant usage throughout the year (Fig. 2).
◼ In addition to continuous expansion of data, improvements will be examined in the next fiscal year in consideration of long term
accumulation of analytical data.

Table 1 Compilation of data from the last fiscal year
(As of Feb 14, 2020)
Data (Sample)

17

IRID/ JAEA

12

Contaminated waterTokyo Electric Power Company

68

Storage amountTokyo Electric Power Company

1012

Total for FY2019

-

1120

Total data items

-

11180

No. ofaccesses permonthFig. 1 Overview of database (FRAnDLi) for analysis
results

Jan

Published
contents (English)

Feb

Published
contents(Japanese)

EnglishwebsiteApr

JapanesewebsiteCommunication

No. of accesses permonth (Cases)

Database (CSV
format)Cumulative
access countCumulative accesscount (Cases)

Automatic generation

Dec

Uploadtool

IRID/ JAEA

Building sludge

Oct

Database

Sampleclassification
masterDebris

Nov[Server]

11

Sept

UploadIRID/ JAEA

July

Summary file
(Excel format)

Contaminated waterAug

Summary
file

Total stored
items

May

Data input

Data reporting organization

June

Materialpublished online

Fig. 2 Transition in number of times the
database is accessed (Including repeatedaccess)©International Research Institute for Nuclear Decommissioning209

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationvi. Evaluation and management of analytical data
◼ Study on the contamination of stored waste The current analytical data contains little data on alphanuclides and on I-129, which is important for theevaluation of safety of disposal. Study was conducted on
the utilization of waste storage data in order to expand
the analytical data for these nuclides.
 Waste storage data was organized according tocontamination classification in order to study the wastes
to be analyzed. (Fig. 1)
 The occurrence locations where alpha-contaminatedwaste has been recorded are limited to the area of Units
1 to 4. However, alpha-contaminated wastes recorded
are in very small amounts in the wastes generated from
Units 1 to 4 areas. (Fig. 2)◼ Study of sampling methods as considered necessary for
future analysis based on the organized data of waste
storage.
 The following 3 approaches were studied in connection
with the sampling methods using waste storage data:• Collection from the containers in the solid radioactive
waste storage facility• Collection from the existing storage area (Other thansolid radioactive waste storage facility)• Collection during work planned in the future

None
Beta contamination
present
Alpha
contamination
present
Alpha and beta
contamination
present
Total/total
Units 1 to 4
Commonfacilities

Units 5 and 6
Outside thepremises

Other premises
Water treatment
facility

Tank area
Wastes area

Fig. 1 Percentage of occurrence locations according
to contamination classification of waste storage data

Alpha and beta contamination
present
Alpha contamination present
Beta contamination present
None
No entries

Fig. 2 Contamination classification ratio for the wastes generated from Units 1 to 4

©International Research Institute for Nuclear Decommissioning210

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationvi. Evaluation and management of analytical data
◼ Study was conducted on types of debris using the debris types database compiled in FY2018 for the
sampling methods to be implemented in future.
 A conceptual study was conducted on the method of collecting samples from the containers inside the solidradioactive waste storage facility, and the method of collecting samples from the wastes that will be
generated in future (Fig. 1), and issues related to the facilities, operations and exposure during work wereidentified.
 It may be reasonable to conduct a study on collecting samples from the wastes that will be generated infuture on a priority basis, since sampling from the already stored wastes requires considerable manpower
Investigation ofRequest for cooperation from
Coordination of sampling
and time.
wastes produced
the Engineering Management
schedule with the prime
Advancepreparationfrom work

Group, TEPCO Holdings Inc.

contractor’s in-charge

Work preparation

Primary sampling
Accompany theprime contractor’s
work in-charge

Selection ofwastesAlpha contamination
measurement

Work site

Have the Sampling
No. recorded while
entering the relevant
wastes in the Debris
and felled tree
control sheet

Beta + gamma
contamination
measurement

Transportation ofwastesTemporary storage

Sampling point
entry

Sampling
process

Waste sampling
process•••

Secondarysampling

Seal the sample
container

Transport
ation

Storage atdelivery point

In case of alpha contamination, it is mandatory toproceed with the work within the house where alphazone is set
Use a band saw
Sample 10g - surface dose rate 1 mSv/h or lessFig. 1 Example of work flow for the method of collecting samples from the wastes that will be produced in future©International Research Institute for Nuclear Decommissioning211

【Confidential】(a) Characterization efficiency improvement① Establishment of method for characterizationvi. Evaluation and management of analytical data
◼ It is indispensable to establish an approach to the analysis plan in conformity with the characteristics of Fukushima
Daiichi wastes in order to continue the analysis. Hence, investigation was carried out in a joint effort with NNL in U.K.
◼ DQO process (Data Quality Objectives Process) is an approach to the analysis plan developed by U.S.
Environmental Protection Agency (EPA) for environmental restoration (Clearance)#. This approach is recognized as
the standard method in the field of radioactive wastes. (Fig. 1)◼ This approach is applied in the analysis of wastes
in U.K.
In addition,
it is also combinedwith Bayesian
DQO
Process
revisiondeveloped
for statistics forDevelopment of Statistically Underpinned
subjects with uncertain parent population. (Fig. 2)

the workshop
The Data Quality Objective Process

Sampling and Analysis Plans for F1 Wastes

Development o
Sampling and An

Step 1Planning
State the problem
EPA, “Guidance on Systematic
Using the Data Quality Objectives The workshop took the DQO stru
an approach, based on the seven
Step 2 Identify the goals
Process,” EPA QA/G-4, 2006
applied to the selected three wasdata
The DQO Process uses systematic planning and
Thetheapproach
captured all the re
Step 3 Identify
information inputs
review (historic and future data) for creating sampling
DQO process and provides feed
strategies for a wide range of purposes.
to early stages of the process sh

Step 4 Define
the boundaries
available
require this.
approach
The DQO Process is a seven-step planning
based on the scientific method that is used to define
data collection activities such as sampling or The approach enables demonstr
Step 5 Identify the necessary data
data may be sufficient and no fur
environmental monitoring.
required to address the problem
It provides the criteria that a sampling design shouldthat be the case.
satisfy, sample collection locations, tolerable decision
error rates and the number of samples to collect.
Step 6

Step 7

Fig. 1 DQO process of U.S. EPA

10Study the effects of additional analysis

Develop the analysis plan

Fig. 2 DQO Process applied in waste analysis in U.K.

11
* US EPA, “Guidance on Systematic Planning Using the Data Quality Objectives Process. EPA QA/G-4 (2006).

©International Research Institute for Nuclear Decommissioning212

(a) Characterization efficiency improvement① Establishment of method for characterizationvi. Evaluation and management of analytical data

【Confidential】◼ Case study was conducted by selecting debris, felled trees, waste generated from contaminated water treatment (slurry) as waste
characteristic to the Fukushima Daiichi NPS. Assuming the conditions in the U.K., evaluation was conducted on the required analysis
score by estimating the concentration distribution using Bayesian statistics.
 It is important that debris has a dose rate classification closer to concentration levels in disposal classification and it is necessary to
conduct analysis in stages.
 In case of felled trees, the required analysis points were estimated in comparison with the concentration levels in disposal
classification assuming the concentration after incineration. (Fig. 1)
 In case of slurry, radioactive concentration was high when compared to disposal classification and hence the necessity of analysis
was low. Since it is discharged from technologically-designed facilities, it is easier to characterize compared to debris, etc.
◼ Using this example as reference, it was decided to proceed with the study in the next fiscal year presuming the situation within Japan.

Fig. 1 Example of study conducted on the analysis score for felled trees using a method
incorporating DQO and Bayesian statistics©International Research Institute for Nuclear Decommissioning213

(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods◼

[Restricted to authorized
persons]

A study was conducted for the simplification and speeding-up of the current methods of analysis
with an aim to establishing analysis methods that will be used regularly for the analysis of accident
waste. (Figure.1)

Applicability of thesampling technology

Received items

Applicability of standard
analysis methods

Blank manufacturing
Samplemanufacturing
Crushing

Drying/Heating/Collection

Distillation

Distillation

Alkali fusion
Adsorption
separation

Adsorption
separation

Samplepreparation

Na
adsorption

Formation ofdeposits

Applicability of automation
technology

Acid decomposition(fully dissolved)

Adsorption
separation

Separation

Filtration
Evaporated
into drysolidifications

Adsorption
separation
Separation

Separation

Dilution

Adsorption
separation
Np-237, U (5 nuclides), Pu(5 nuclides), Am (3nuclides), Cm (3 nuclides)

Applicability of streamliningof the separation process

Si Detector
Green box indicates the location ofchanges during the ICP-MS application

Fig.1 Targets for analysis method improvement and strategies for the same
©International Research Institute for Nuclear Decommissioning214

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
i. Development of sampling technology◼ Results
 By ensuring the representativeness of the sample through evaluation of the distributionof radioactive concentrations on the surface of radioactive waste by a non-destructive
analysis, a methodology that would work with as minimal sampling as possible, wasdeveloped (Fig. 1).
 A device that would measure the distribution of the gamma ray releasing nuclides on thesample surface and a device that would collect the sample in the form of powder fromany location, were prototyped (Figures 2 and 3).
◼ Goal To highlight specific strategies along with manufacturing a mock-up device for sampling
and thereby, to identify the issues in the verification of functionality and in real-time
application in order to establish a practical sampling method.
◼ Implementation details
 The methods to decide the collection location and the methods to verify the sampling
accuracy and the validity of the sampling size were studied in FY2019 with respect to thesampling size from the statistically evaluated sample surface. Moreover, based on thefindings obtained from the prototype manufactured in FY2018, a mock-up device wasdesigned by carrying out tests to verify the selection of conditions etc. to optimize thedesign. During the verification testing related to the mapping device, optimization of
background correction methods and the optimization of the positional relation between
the samples and the detector was studied, so that highly sensitive and quantitative
measurements were possible. During the verification testing related to the sampling
device, collection methods that suit the material and the optimization of the design inorder to accurately collect the cut sample from any location, were studied.
 In FY2020, the mock-up device will be manufactured, and any points of improvements or
issues to be resolved will be selected from the point of view of site application.

◼ Indicators determining achievement of goals
 Specific strategies should be presented for a practical sampling method for the analysis
samples. (FY2019) A blueprint of the mockup device to be used for the sampling device should be prepared.
(FY2019)
 Any points of improvements or issues concerning the manufacturing of the actual
equipment should be presented. (FY2020)

Sampling forms
All

RandomLayer-wise

Last form

Evaluation of the necessary samples by using statistical methods

Condition A

Condition B(Probability distribution) (Indicators ofconcentration distribution)

A dose mapping technology is adopted
to verify the parameters required forstatistical evaluation.

Condition C(The indicators of concentration
distribution and nuclide concentration
have a one-to-one correspondence andSF evaluation is possible.)

Reduction

Fig.1 Illustration of sampling technology thatcan work with a minimal amount of sample

Fig.2 Prototype sample mapping
deviceDiaphragm
vacuum pump

Electric drill

Nilu Filter holder

Fig.3 Sampling device prototype©International Research Institute for Nuclear Decommissioning215

(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
i. Development of sampling technology◼

A mock-up device for the device that would measure the distribution of the
gamma ray-releasing nuclides on the sample surface (mapping device) was
designed based on the findings obtained from the prototype manufactured in
FY2018.
 The device consists of a sample feeding part, a sample fixing part, a shapemeasuring part and a radiation measuring part (Fig. 1).
 A system was designed to record the geometrical shape and the radiation
data of the target sample obtained by means of measurements.
◼ A mock test was conducted to examine which radioactivity other than that ofthe part to be measured would influence the measurement value. The
influence from the surroundings of the part to be measured was about 6%
higher, and it could be confirmed that there was sufficient resolution toevaluate the distribution. (A-B in Fig. 2). Even when the difference in theconcentrations was large, the influence was about 50% and the shades of theradioactive concentrations could be confirmed with high resolution (A’-C in Fig.
2).
◼ A study regarding a statistical methodology was conducted to enableevaluation with a minimal amount of sample, while ensuring therepresentativeness of the sample concentration on the basis of the data on the
distribution of the gamma ray-releasing nuclide concentration on the sample
surface.
 With regard to the distribution of radioactive concentrations on the sample
surface, evaluation formulas for the number of samples were studied
assuming two cases based on normal assumptions.

Shapemeasuring part
Radiationmeasuring part
Radiationmeasuring
container【Confidential】

Laser
displacement
meter IL-100:
Keyence
Sample fixing
part

Sample feeding partFig.1 Entire outline of the mock-up device of the device to measure the
distribution of the gamma ray-releasing nuclides on the sample surface(mapping device): Radiation level [Bq] ofpart to be measured: Radiation level [Bq]in the vicinity of thepart to be measured
A,A’: Measurement value in which
there is no influence of thesurrounding radiation level

B,C: Measurement value in which
the influence of surrounding
radiation level is consideredFig.2 Test concerning resolution for measurement (3 cm square)©International Research Institute for Nuclear Decommissioning216

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
i. Development of sampling technology◼

A mock-up device for the device that would performthe sampling of the analysis sample from thesampling surface was designed, based on thefindings obtained from the prototype manufactured
in FY2018.
 A drill material suitable for the target material wasstudied, and a drill blade was designed andprototyped with a shape that enabled the powder
generated during cutting to accumulate around thecutting hole instead of being dispersed in thesurroundings, and which could facilitate the easycollection of this cutting powder (Fig. 1).
◼ A concrete cutting test was conducted to evaluate
the validity of the study results. The collection rateof the powder was obtained (Table 1 shows theresults for a depth of 25 mm) by cutting at various
depths (25, 50, 75, 100 mm).
 A high collection rate of 96% or more wasobtained in case of all the cutting depths.
 It was understood that the relative standard
deviation (RSD) was 7% or less for all the cutting
depths, and the cutting powder could be collectedwith a high level of repeatability.

Nilu filter
To the diaphragmvacuum pumpCutting
powderDrill blade
Jig to prevent the dispersalof cutting powder

The cut powder issucked andcollected by the
Nilu filter.

Solid waste

Fig.1 Proposed study of the sampling mechanism
Table 1 Cutting powder collection test results (cutting depth 25 mm)

nNo.
数

Weight of the Nilu filter [g]【g】 コンクリートブロック重量
Weight of the concrete block [g]ニールフィルター重量【g】
Packing
風袋

With cutting

After theDifference
差引

Difference
to test 試験後
切削粉込
差引 Prior
試験前
powdertest

Cutting powder collection
切削粉回収率【%】rate [%]

1

143.81147.67

3.86

1857.52

1853.57

3.95

97.7

2

143.81147.503.69

1853.57

1849.79

3.78

97.6

3

143.81147.36

3.55

1849.79

1846.14

3.65

97.3

4

143.81147.293.48

1846.141842.51

3.63

95.9

5

143.81146.94

3.13

1842.51

1839.27

3.24

96.6

6

143.81147.133.32

1839.27

1835.82

3.45

96.2

Average
value
平均値

3.51

Average
平均値
value

96.9

標準偏差
deviation

Standard

0.24

RSD【%】

6.8©International Research Institute for Nuclear Decommissioning217

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
ii. Study on streamlining the separation process◼ Results
 The triple quadrupole ICP-MS (ICP-MS/MS) is believed to be useful from the viewpoint of streamlining of the separation process because the influence of
nuclides of the same atomic mass number (isobars) can be reduced by reaction with gas, and the device on its own has a high separation performance.
Therefore, the latest findings (information about lower limit of detection, effective solid phase extraction agent and effective reaction gas etc.) related toanalysis using the ICP-MS method, were organized (excluding actinides). As a result, it was possible to propose a separation process, which was changed
from beta-ray measurement using a liquid scintillation counter (LSC) requiring complicated pretreatment, to the ICP-MS method, which can simplify the
pretreatment process for eight nuclides. (Fig. 1)◼ Goals
 To evaluate the optimal separation and measurement conditions for the proposed separation process by testing with simulated radioactive waste samples.
 To propose a similar streamlined separation process for actinides.
◼ Implementation details
 In FY2019, the optimal separation and measurement conditions for the
separation process proposed in FY2018 were studied by testing with simulated
samples. Optimum liquidity and fluid volume of reagents during the extraction or
the separation process were studied for the separation conditions. Withreference to the measurement conditions (ICP-MS/MS), device conditions suchas the optimal types of gas and flow rate were evaluated for the separation from
hindering nuclides. Moreover, like FY2018, even in case of actinides, a rational
separation process presuming alpha ray spectrometry and ICP-MS, which isbelieved to be able to separate and measure multiple nuclides in batches, were
proposed, and calibration methods for that device were also proposed.
 In FY2020, in continuation with FY2019, tests using the simulated samples will
be conducted.
◼ Indicators for determining goal achievement
 Optimal separation and measurement conditions for a streamlined separation
process should be presented. (FY2019, FY2020).
 A list of the current analysis methods for actinides should be presented.
(FY2019)
 A proposed practical method concerning actinides should be presented.
(FY2019)
 Proposed calibration techniques for devices used for actinide measurement
should be presented. (FY2019)

Crushing

Drying/Heating/Collection

Distillation

Distillation

Alkali fusion

Separation

Samplepreparation

Separation

Acid decomposition

Adsorption
separation

Adsorption
separation

Adsorption
separation

Separationandpurification

Separationandpurification

Separationandpurification

Separation

Formation ofdeposits

Dilution

Dissolution ofdeposits
Formation ofdeposits

Separation
Separation

Dissolution ofdeposits
Separation

LSC: 13 Nuclides ⇒ 5 Nuclides
ICP-MS: 2 Nuclides ⇒10 Nuclides

Streamlining and speeding up ofprocesses
Decrease in exposure and waste

Fig.1 Outline of the separation process introducing the new ICPMS method©International Research Institute for Nuclear Decommissioning218

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
ii. Study on streamlining the separation process◼

Items to be evaluated during the tests to evaluate validity or feasibility were
decided (Table 1).
Validity evaluation testing of the separation process was conducted with theseparation process applying the ICP-MS method and with the radioactivity
measurement method#.

# The beta ray measurement nuclides (36Cl, 41Ca, 63Ni, 90Sr) for which the application of the ICP-MS method isdeemed unsuitable from the viewpoint of detection sensitivity

•
•
•
•

Collection rate evaluation
Evaluation of the effects ofspectral interference concerningisotopes
Evaluation of the effects ofspectral interference concerning
isobars
Evaluation of the effects ofspectral interference concerning asample matrix

Radioactivity measurement method•••

Collection rate evaluation
Evaluation of nuclides that could
hinder measurements due tosimilarities in chemical behavior,etc.
Evaluation of the decontamination
coefficient of hindering nuclides

Zr is effectively separated from theisobars (93Nb,93Mo) by using the NH3 gas

Conditioning (Pure water 5mL, 1M HNO310mL)
Addition of sample
Acid wash (1M HNO35mL)
Elution of Zr by oxalic acidNH3 flow rate
H2 flow rate

Elution of Nb by high concentration of oxalic
acid

Pure water
Elution of Mo by Nitric acid

Zr resin

Zr collection liquid -> ICP-MS/MS
measurement
Mo collection liquid -> ICP-MS/MS
measurement

Fig.1 Separation process of 93Zr and 93Mo

Nb count◼

For the chemical separation of 93Zr and 93Mo, a method with a high selectivity for Zr
and Mo while separating out the hindering Nb, was proposed (Fig.1).
The conditions to separate Nb were found during the analysis of ICP-MS/MS, whenammonia was used as a reactive gas (Fig. 2, 3).

ICP-MS method

Zr count◼

Table 1 Evaluation items for the validity or feasibility tests

NH3 flowrate

Fig.2 Separation of Nb and Zr by means of areactive gas

Count

◼

gas

Fig.3 Separation of 93Zr from 93Nb, 93Mo bymeans of ICP-MS/MS©International Research Institute for Nuclear Decommissioning219

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methodsiii. Development of automation technology◼
Results
 The performance of the manufactured automated systems was compared with
operations carried out by skilled personnel (Fig. 1). A series of analyticaloperations were carried out for Ni assuming the most complex Ni-63 separation
operations and the utility of the automated systems was proven as similarcollection rates and accuracy were achieved when the collection rates werecompared (Table 1).
◼
Goal To evaluate the feasibility of the separation process of nuclides to be measured
using the manufactured automated systems.
◼

Implementation details
Separation and purification tests using the automated systems were conducted
and their feasibility was evaluated with respect to the separation process which
was studied under "(ii) Streamlining of the separation process" throughout FY2019
and FY2020. Since the aim of these tests is to evaluate the feasibility under theactual analytical conditions, the tests will be conducted by using simulated
radioactive waste samples.
Among the separation processes, the types or quantities of the solid-phaseextraction agents and the types and concentration of reagents that will be used foreach of the elements subject to separation during the solid-phase extraction
operation, will be different. For this reason, it is necessary to evaluate the feasibilityof the operation conditions of the automatic solid-phase extraction system byselecting the column size and optimizing the reagent flow rate during theseparation operation according to each target element. In particular, if there is alarge difference between the rate of reagent addition to the resin and the liquid
passing rate in the resin, the reagent may accumulate at the top of the column,
overflowing the reagent, or mixing with the reagent in the next step, which may
have a large impact on the separation accuracy. Therefore, it is important toevaluate the optimum operation content under each analysis condition.
In addition, if a problem that cannot be sufficiently addressed within the operating
condition range of the current automation device is found in the test from theviewpoint of accuracy and speed, improvement of the device will be considered tosolve the problem.

◼
Indicators for determining goal achievement
 The feasibility evaluation results for the separation operation using theautomated system for the streamlined separation process should be presented.
(FY2019 and FY2020)

Analytical operation units
Dissolution operation
Sorting and distribution operations
Reagent addition operation
Filtration operation
Heating and solidification by evaporating
and drying operation
Constant melting operation
Ion exchange and solid-phase extraction
operation

Automatic solid-phase extraction system
Improvement points: It is now
possible to use columns in 3
sizes

Improvement points: System improvements such that it ispossible to insert air between the reagent addition
operations so that there is no mixing between the various
chemicals

Fig.1 Configuration of the automation technology and the improved
automatic solid-phase extraction equipment
Table 1 Performance evaluation of the automated system(Comparison of collection rates during the Ni-63 analysis)
Operations

Ni Collection
ratesStandard
deviation

Automated systems

89.8%

2.9%

Skilled analysts

87.0%

2.2%

©International Research Institute for Nuclear Decommissioning220

(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methodsiii. Development of automation technology
◼
◼
◼

◼

【Confidential】

The validity and feasibility of the examined automated equipment was evaluated with respect to the size of columns filled with resin and thereagent liquid passing speed.
The automatic solid-phase extraction equipment consists of a column unit, a pump unit and a control computer. It is possible to process 4
samples at the same time (Fig. 1).
The applicability of the automatic solid-phase extraction equipment was studied by targeting the analysis methods which combined theextraction chromatography with the ICP-MS/MS (Table 1). As for the repetition accuracy of the Zr collection rate, an excellent result was
obtained when compared with the relative standard deviation of 2.6%. when done manually.
A phenomenon wherein the eluant might not spread throughout the resin was observed (Fig. 2). In FY2020, a further speeding-up will be
studied by optimizing the liquid passing speed along with making further improvements in the repetition accuracy and the collection rate
through improvements in this phenomenon.
Table 1 Reproducibility of extraction andseparation of Zr and Mo with the automaticsolid-phase extraction equipment
Pump unit (4 systems)

Zr collection rate [%] Mo回収率[％]
Mo collection rate [%]Zr回収率[％]
Pump A ポンプA

73.8
82.9
84.5
80.4
7.1

94.5
95.0
80.1
89.8
9.4

70.3
82.6
75.0
76.0
8.1

92.5
94.6
93.6
93.5
1.1

相対標準偏差Relative
Standard Deviation

74.3
76.0
82.3
77.5
5.5

91.2
88.9
89.9
90.0
1.3

平均
Relative
Standard Deviation
相対標準偏差

78.0
5.1

91.1
4.7

Pump B ポンプB

Column unit

Day 1

Pump C ポンプC
Average

Control computer

平均

相対標準偏差Relative
Standard Deviation
Pump A ポンプA
Pump B ポンプB

Day 2

Pump C ポンプC
Average

平均

相対標準偏差Relative
Standard Deviation
Pump A ポンプA
Pump B ポンプB

Day 3

Fig.1 Automated system with the automatic solid-phaseextraction equipment at the core
(Part of the ion-exchange and solid-phase extraction)

Pump C ポンプC
Average

全体

Average

平均

The flow of the eluant
was not uniform.
Fig.2 State of the column after
the Mo elution operation©International Research Institute for Nuclear Decommissioning221

【Confidential】(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
iv. Establishment of standard analytical techniques◼
Results
 In order to evaluate the feasibility of the proposed separation
process for the beta ray measurement nuclides, separation andpurification tests were carried out for the target nuclide elements.
Excellent results were obtained for both the collection rates as wellas the standard deviations, and it was therefore expected that it
could be applied to radioactive waste analysis (Fig. 1). Moreover,since beta nuclides generally require complex chemical separation
operations prior to measurements, techniques to train engineers
who will start new analysis on this topic were studied.
 A Quality Assurance guideline draft was created, which could bereflected in a system conforming to the ISO9001 Standard, andwhich takes into account the swift verification and validation ofstandard analysis methods (Fig. 2) for a wide variety of analysis
samples that are characteristic of radioactive waste.
◼
Goal To evaluate the feasibility of the methods of analysis using thestreamlined separation process, in consideration of quality
assurance in the form of radioactive waste analysis method.

Sr
Resin

Sample (Including Sr and Ni carriers)

Cleaning
solution

Ni
Resin

EluantCleaning Sr fixed quantity
solution

Sr

88.1%

2.8%

Ni

95.8%

3.7%

Waste
Ni fixed quantity
Fig.1 Example of separation and purification test (System separation of the Sr and Ni nuclides)

Sample (matrix)

Elements(nuclides)

Dissolution(liquefaction)

Separation

Concrete

Acid dissolution Ion exchange

Ashes

Columnseparation

Alkali fusion

Conversion into Separation of
ashes
deposits

Soil[Speeding up of the verification and validation of
the standard analysis methods]

Measurement

Developmentof an analysis
methodVerificationand validationof the analysis
methodExternal
ExpertCommittee

Standard
analysis
methodCooperation

Analysis Office for the 1F
radioactive waste
Past

Verification and validation conditions
for the analysis method◼
Indicators for determining goal achievement
 Feasibility evaluation results for the radioactive waste analysis
methods with respect to the analysis methods using thestreamlined separation process should be presented. (FY2019
and FY2020)

Standard
deviation

Eluant[Streamlining the verification and validation of the standard analysis methods]◼
Implementation details
 Throughout FY2019 and FY2020, each parameter corresponding
to the test conditions shown in Fig. 2, was evaluated for the
analysis method that uses the separation process examined in"(ii) Streamlining of the separation process", and its feasibility as aradioactive waste analysis method was assessed. If it isdetermined that the verification and validation conditions are notsatisfied, such as the lower limit of detection cannot be obtained
due to the influence of hindering nuclides, the influence removal
method or alternative analysis method will be considered.

Collection
rate

Selectivity and uniquenessMeasurement range
Calibration and traceability

Cooperation
New

Verification ofanalysis
methodExternal analysis office

Inclination
Linearity
Detection lower limit/
quantitative lower limit
Toughness
Accuracy

Developmentof the analysis
methodVerificationand validationof the analysis
method(In the analysis
office) Standard
analysis method

Analysis Office for the 1F radioactivewaste

Fig.2 Specific methods of streamlining and speeding up verification and validation©International Research Institute for Nuclear Decommissioning222

(a) Characterization Efficiency Improvement
 Simplification and speeding-up of analytical methods
iv. Establishment of standard analytical techniques【Confidential】◆ Presentation of proposed specific evaluation items regarding the verification and validation conditions forthe standard analysis methods (Table 1).
Table 1 Verification and validation methods for standard
analysis methods
Dissolution(liquefaction)

Separation

ConcreteAciddissolution

Ion exchange

Ashes

Alkali fusion

Columnseparation

Soil

Conversion into
ashes

Separation ofdeposits

Sample (matrix)

Elements (nuclides)

Verification and validation conditions for the analysis method

Selectivity and uniquenessMeasurement range

Calibration and traceability

Measurement

Study of specific evaluation items (Proposed)[Dissolution]①: Dissolution/ leaching rate and their repetition accuracy②: Influence on solubility or on the repetition accuracy due to differences in the sample
properties or sample elements[Separation]③#: Collection rate and repetition accuracy④ : Necessity of enrichment operation to obtain the necessary sensitivity⑤#: Influence on the collection rate or on the repetition accuracy due to differences in thesample properties⑥#: Do the nuclides believed to hinder measurements satisfy the decontamination coefficientrequired for the Fukushima Daiichi radioactive waste analysis methods? (Beta ray
measurement nuclides)⑦ : Evaluation using methods confirming that the hindering nuclides can be removed fromthe measurement sample (Gamma ray nuclide measurement from the measurement samples)(Beta ray measurement nuclides)⑧#: Evaluation of spectral interference by comparison of the B.G counts of operation
BLANK test and the simulated sample (ICP-MS measurement nuclides)

Inclination# : Executed with validity evaluation from the “Study on
streamlining of the separation process"Linearity[Measurements]
Detection lower limit/
quantitative lower limit
Toughness

Accuracy⑨ : Impact of the difference in the matrices of the calibrated sample and the measurement
sample on the measurement values⑩ : Device calibration and its measurement accuracy that ensures traceability⑪ : Evaluation of the method detection lower limit/ quantitative method lower limit⑫ : Linearity of calibration curve including the quantitative lower limit (ICP-MS
measurement nuclides)⑬ : Range of energy for the beta ray measurements (Beta ray measurement nuclides)©International Research Institute for Nuclear Decommissioning223

【Confidential】(b) Development of sampling technology◼

Past results (FY2018)➢➢➢

A conceptual design was created for the sampling devices
inside the adsorption towers of the cesium adsorption apparatus(KURION) and the secondary cesium adsorption apparatus(SARRY). (Fig. 1)
A testing apparatus was manufactured for the collected
elements and tests were conducted to gather the zeolite
samples from inside the adsorption tower to verify the collection
performance of the sampling tool. (Fig. 2)
On the basis of the test results, the structure of the sampling
tool was studied and a basic design was developed. (Fig. 3)

Rod

Sampling toolFig. 2 Testing apparatus for the collected elements
Sampling device

Coring device

Modified sampling
tool
resin modelWork standWork dock
Coring device

Sampling device

Fig.3 Sampling tool design
Fig.1 Conceptual design of the sampling device©International Research Institute for Nuclear Decommissioning224

【Confidential】(b) Development of sampling technology◼ GoalTo study the sampling devices compatible with the cesium adsorption apparatus (KURION) as well as thesecondary cesium adsorption apparatus (SARRY) for the collection of cesium adsorbent, and to test andevaluate the element technology necessary for designing the mock-up devices.

◼ Implementation details⚫

FY2019During the study of sampling devices, information such as the contents inside the adsorption towers, stageof generation, water flow time, surface dose rate etc. was organized and depending on the analysis
requirements, the adsorption towers to be collected, were selected. Together with this, the collection conditions
were reviewed. Based on the study results of the sampling devices compatible with both, the cesium
adsorption apparatus (KURION) as well as the secondary cesium adsorption apparatus (SARRY) adsorption
towers for which the conceptual design was created, items for the element tests including an opening methodof the adsorption tower, collection of samples, closing of the opening. required for the mock-up device wereselected, the required element test devices and the simulated adsorption towers for testing were designed andmanufactured, and test preparations were carried out.
⚫

FY2020Once the element test device is ready, element tests will be conducted and evaluation will be carried out formaking openings of the adsorption towers, collection of samples, and closure of the opening. Based on theevaluation results, a perspective will be obtained on the collection methods and devices applicable at the site.

◼ Indicators for determining goal achievement

FY2019Selection of element test items necessary for designing the mock-up sampling
device and preparation for testing.
(Target TRL at completion: Level 3→4)

FY2020Completion of element tests necessary for designing the mock-up device andevaluation of the same.
(Target TRL at completion: Level 3→4)
[Excerpts from the Subsidy Project Manuals]©International Research Institute for Nuclear Decommissioning225(b) Development of sampling technology【Confidential】- Conditions for selecting and collecting the adsorption towers (1/4) Selection policies for the samples to be collected from the adsorption towers were studied in order toselect the adsorption tower to be used for collection of samples.
◼ Analysis and evaluation requirements➢

Alpha nuclides have been detected during the analysis of stagnant water. Since the alpha nuclides havea very high sorption distribution coefficient, there is a possibility that they exist in the vicinity of theadsorbent surface.

➢

The results of the interim report on safety evaluations during disposal indicate the importance ofconfirming the existence of predominant nuclides (C-14, I-129 etc.) during disposal.

➢

The properties of the adsorption tower will change significantly depending on the types of adsorbent,
composition of the stagnant water, operating conditions etc. The assessment and evaluation of individual
differences is important.

➢

The adsorption towers, which were used during the initial stage of the earthquake, should be collectedon priority because the concentration of nuclides in the stagnant water is high in such adsorption towers.

◼ Approach to representativeness➢Regarding the representativeness of the collected samples, two approaches can be considered: one
focused on the adsorption towers (individual differences between the adsorption towers) such as the
structure of the adsorption tower, period when it was used, adsorbent used, and the content ofradionuclides, and the other focused on the adsorbent location (spatial non-uniformity) such as theradionuclide distribution inside the adsorption tower.
©International Research Institute for Nuclear Decommissioning226

【Confidential】(b) Development of sampling technology- Conditions for selecting and collecting the adsorption towers (2/4) ◼ Individual differences between the adsorption towers➢ There are 10 or more types of adsorbents used for the adsorption towers.
➢ The composition of stagnant water changes with the passage of time and the composition of thecontaminated water passing through the adsorption tower also changes as per the "Structure” and"Arrangement" of the adsorption tower which acts as the treatment system.

➢ Therefore, the nuclide composition and the radioactivity concentration for each adsorbent solid varies. It has
actually been indicated that the variation in the surface dose of the adsorption towers is large, and that theindividual differences are large.
➢ Especially, the nuclide composition captured by the adsorbent has a large impact on treatment method
selection and the treatment equipment design, and it is necessary to understand the trends.
Types of adsorbents(Surface Modifying Zeolite)(Hershlite)

Surface dose rate

KURION adsorption towers
SARRY adsorption towers
No. of towers: Approx. 800 units No. of towers: Approx. 200 units(Silver added Zeolite)(Synthetic Zeolite)(Titanate silicate system)

Silica sandDate of removal of the adsorption towers

Fig.1 State of storage of the used adsorption towers

Fig.2 Example of data organization of the KURION adsorption
towers (Contents, dose rate and date of generation)©International Research Institute for Nuclear Decommissioning227

【Confidential】(b) Development of sampling technology- Conditions for selecting and collecting the adsorption towers (3/4) ◼ Spatial non-uniformity➢ In FY2015, Cs adsorption distribution measurements (JAEA/ IRID#) (Fig. 2) were carried out by filling IE-96
adsorbent in a full-scale SARRY adsorption tower internal mock-up facility (Fig. 1).
•

It was found that the Cs concentration distribution in the direction of depth accurately matched withthe analytical value.

•

It was confirmed that there was almost no concentration distribution in the circumferential direction.

D

C
B

A
B
C

Axial direction

E

E

軸
方
向

A

2017/1/2414

平衡吸着Cs濃度（Cs7.7ppm）
Cs concentration
at equilibrium adsorption (Cs 7.7 ppm)
（ゼオライト充填密度0.726[g-80%RH/ml-充填層）(Zeolite
filling density 0.726 [g-80%RH/ml – filled layer])

12
10

Axial
direction
軸方向
A A
Axial
direction
軸方向
B B
Axial direction C
軸方向 C
Radial direction D-E
径方向
D-E (Langmuir
ZAC
analysis
ZAC解析（Langmuir式）
equation)

8

6
4
2

0
10

20
30
40
50
Depth from 充填層上面からの深さ[cm]the top side of theSide viewFig.1 SARRY adsorption tower internal mock-up facility

2017/1/24

100

Depth from the top side of the filled layer [cm]平衡吸着Cs濃度（Cs7.7ppm）
Cs concentration
at equilibrium adsorption (Cs 7.7 ppm)

10

Axial
direction (19 cm from the center)
軸方向（中心から19cm）
Axial
direction (32 cm from the center)
軸方向（中心から32cm）
Axial
direction (23 cm from the center)
軸方向（中心から23cm）
Radial
径方向direction
ZAC
analysis (Langmuir equation)
ZAC解析（Langmuir式）

1
0.1
0.01

(Zeolite filling density

0.001（ゼオライト充填密度0.726 [g-80%RH/ml –0.726[g-80%RH/ml-充填層）
filled layer])

0.000100

Top view

Adsorbed Cs concentration
吸着Cs濃度[mg-Cs/ml-充填層][mg-CS/ml-filled layer]

D

Adsorbed Cs concentration
吸着Cs濃度[mg-Cs/ml-充填層][mg-CS/ml-filled layer]

 The concentration distribution of the entire adsorption tower can be presumed if theadsorption concentration of one place is actually measured.

60

filled layer
Cs adsorption distribution measurement value(linear display)

70

10

20

30

40

50

60

70

Depth from 充填層上面からの深さ[cm]the top side of thefilled layer
Cs adsorption distribution measurement value(Semi-logarithmic display)

Fig.2 Cs adsorption distribution measurement value ##IRID (2017): “Subsidy Project of Decommissioning and Contaminated Water Management in the FY2014 Supplementary Budgets, R&D for Treatment and Disposal of
Solid Radioactive Waste (Final report)” pp. 3-20 – 3-22, pp. 4-381 – 4-386©International Research Institute for Nuclear Decommissioning228

【Confidential】(b) Development of sampling technology- Conditions for selecting and collecting the adsorption towers (4/4) ◼ Selection of adsorption towers➢➢➢

The adsorption towers used during the first stage ofthe operations when the concentration of nuclides inthe stagnant water was high, have a higher priority.
The SARRY adsorption towers used in the firststage of operations correspond to the flange type.
Samples are collected giving priority to theadsorption towers installed on the upstream side.

◼ Number of adsorption tower sampling points➢

Collect samples from one point on each adsorption
tower.

◼ Collection location➢

The surface range is set from the surface of theadsorbent (top end) to 100 mm.

Table1 Conditions for selecting and collecting the adsorption towers
Selection and collection conditions
for the adsorption towers

Items

Types

The adsorption towers used during the first stage of theoperations are given priority.
Especially, those on the upstream side of the system aregiven priority.

Structures

KURION：All adsorption towers have a common structure.
SARRY：The flange type adsorption tower is applicable.

Collection
targets

Amount of collection

10 to 20 mL/sample

No. of sampling points per
tower

One point in the circumferential direction x One point in thedepth direction(One point per adsorption tower in total)

Collection depth from thesurface

Surface to about 100 mm

Position

Diameter of opening inadsorption tower
Method of closing the
opening

KURION： Top plate side of the adsorption tower
SARRY: Adsorption material filling port flange

Approx.φ 50 mm
A pressure resistance of 1.37 MPa is ensured in theclosed part.
The closing plug is a detachable structure.

©International Research Institute for Nuclear Decommissioning

229

(b) Development of sampling technology【Confidential】- Selection of element test items (1/3) ◼ Assumptions during actual work
Risks were evaluated assuming actual work at the site and the necessary element test items and the element test device
were studied.
[Existing Crane]
Rated load: 24 t or more
Lift: 7.5 m or more
[Integrated
sampling device][Work Dock]

Install the adsorption tower
on the work stand with theexisting crane.

With the existing crane,
detach the shield lid andload the [Work dock] on topof the adsorption tower.

With the existing crane,
connect the [Work dock]with the [Integrated
sampling device].

Bore holes in the top plateof the adsorption tower withthe [Integrated sampling
device].

Collect the sample byinserting the sampling toolfrom the bored hole.

Attach the closing plug on
the bored hole.

[Work Stand][Adsorption tower](Inside the work stand)

Structures

Functions

Existing Crane

To install the adsorption tower, the integrated sampling device
and the work dock.

Integrated sampling
deviceA device that remotely executes a series of work processes
such as making openings, sample collection, and closing of the
opening by switching the vertical slide unit and tip tool.

Work dockIts function is to shield the upper part of the adsorption tower,
connect the adsorption tower with the sampling device, secure
the boundaries and control the shaking of the rod core during
making openings.

Work standInstallation of the adsorption tower and for shielding the sides.
A scaffold for the workers to work on.

Fig.1 Equipment layout

Fig..2 Work steps (KURION Adsorption tower)©International Research Institute for Nuclear Decommissioning230(b) Development of sampling technology◼ Risk evaluation
Risk items【Confidential】- Selection of element test items (2/3) Table 1 Evaluation of the risks involved in a series of work processes
Risk details

Risk reduction measures

Spread of contamination in the surroundings due to the samplingheads after collection

Evaluating the surface adhesions (contamination) on the sampling apparatus
after collection and appropriate curing and enclosure

Diffusion of contaminants from inside the adsorption tower during thecollection work

Securing the boundaries during the collection work

Leakage of contaminants from the closed part

Ensuring appropriate sealing

Exposure

Excessive exposure of workers due to the operation andmaintenance of the sampling device

Shielding (shielding by the work dock) the upper part of the adsorption tower
Shortening of human working hours

Hydrogen ignition

Ignition of hydrogen generated in the adsorption tower

Ventilation and measurement of hydrogen concentration inside theadsorption tower

Change in properties of the flange gasket of the adsorption tower
due to frictional heat or the shearing force during making openings

Reducing the frictional heat by appropriate conditions of making openings
and lubrication

Difference in the plate thickness, or in case of the flange type tower,
difference in the assumed number of flange plates

Remote verification by camera leaving a margin for the cutting ability andstrokeForeign material intrusion

Mixing of cutting splinters in the adsorption tower

Collecting by suction and assessing the cutting conditions so that the mixing
is minimal

Falling or flying objects

Fall during the installation operation of the work dock and the ISM

Use of special hoisting attachments and surveillance with camera

Damage during transportation of the adsorption tower

Use of special hoisting attachments

Interference with the internal piping during collection

Checking the insides with a camera after making openings

Interference due to shaking of the shaft during making openings andcollecting samples

Assessing the amount of shaking of the shaft beforehand

Equipment parts caught during making openings and closing the
opening

Understanding the conditions under which making openings or closure canbe carried out in a stable manner

Pressure resistance lacking in the closing plug

Process control based on a prior pressure test

Failure in withdrawal of the rods due to failure in return of thesampling head vane (damage)Study of the vane structure
Study on a head shape that does not require vanes

Contamination

Change in properties of
the adsorption tower

Equipment damage

Failure in integrity of theadsorption tower#: Indicates items that must be evaluated through actual tests and are considered as test items in this research.
©International Research Institute for Nuclear Decommissioning

231

(b) Development of sampling technology【Confidential】- Selection of element test items (3/3) ◼ Element test items
Table 1 Element test items
Test items

Test for making
openings

Collection test

Closure test

Installability confirmation
test

Test goals

Measurement items

・Verification of operating conditions
・Determining the need for shaft holding mechanism
・Reflection in the study on the working hours

・Load factor of motor (rotation direction and vertical direction)
・Durability of cutting blade
・Rotation speed and feed speed
・Time required for making openings
・Cutting splinter status and cutting oil supply conditions
・Status of shaking and vibration of the shaft

・Verification of effect on the subsequent work processes (collection
and closure)・Verification of validity of the blade for making openings

・Shape of the hole (circular, coaxial etc.)
・Amount of cutting splinters and cutting oil mixed in the adsorption
tower

・Evaluation of workability of joints
・Validity of equipment layout
・Determining the necessity of contamination prevention measures

・Replaceability and time of replacement of the tool for making
openings
・Scope of work from the glove port
・Dispersion status of the cutting splinters and cutting oil
・Identification of any other issues

・Verification of operating conditions
・Verification of required performance (10 to 20 ml)
・Determining the necessity of contamination prevention measures

・Rotation speed and feed speed
・Amount of collection
・State of sample dispersion

・Evaluation of workability of joints
・Understanding the information input to exposure evaluation(distance and time)

・Handling properties and handling time of collection tools
・Scope of work from the glove port
・Identification of any other issues

・Verification of operating conditions

・Closing torque value
・Feed speed・Verification of the structure of the closing plug・Setting the operating conditions

・Pressure resistance (test pressure: Approx. 2 MPa)
・Loosening torque value for detachment

・Evaluation of workability of joints
・Validity of equipment layout

・Installability and installation time of the closing plug・Scope of work from the glove port
・Identification of any other issues

・Structural verification of the parts aligned with the work dock・Measurement of the work dock installation time・Installability of work dock in the simulated adsorption tower(alignment)
・Identification of issues from the viewpoint of actual work

・Structural verification of the parts that match the position of theelement test device

・Installability of the device on the work dock・Identification of issues from the viewpoint of actual work
©International Research Institute for Nuclear Decommissioning232

【Confidential】(b) Development of sampling technology- Format of the element tests ➢➢Step 1: Preparation for element tests ... Execution of "Design and manufacturing of the element test device" and “Preliminary
testing (making openings, sample collection and closing of the opening)"
Step 2：Element tests... Verification of the feasibility of a series of work processes with the element test device

Preparation for element tests

Element tests(Step 1) Ongoing. Scheduled for execution by the first half of FY2020.

(Step 2) To be executed in the second
half of FY2020.

Organization of the conditions presumed for thedesigning of the element test device

Designing and manufacturing of theelement test device
Design and manufacture the device (Fig. 1)that is required to carry out the element tests.

Verify the feasibility of a series of work
processes related to sampling by using
the element test device (Fig. 3).

Preliminary testing
Verify the basic performance necessary formaking openings, sample collection andclosing of the opening through preliminary
testing (Fig. 2).

Making openings

Collection

ClosureFig.2 Preliminary testing
Integrated sampling
device for the test

Work dock for the test Simulated
adsorption tower

Fig.1 Configuration of the element test device

Reflection of theresult of thepreliminary test in
the design

Fig.3 Illustration of the element test
device©International Research Institute for Nuclear Decommissioning233

【Confidential】(b) Development of sampling technology- Preparation for element tests - Designing and manufacturing of the element test device (1/7) Presuming the site environment and based on the study of the collection conditions and the sampling
procedures and the results of the preliminary tests, a detailed design was created for each component
required to perform the element tests, such as making openings of the adsorption tower, collection ofsamples, closure of the opening etc., and manufacturing of the equipment was started.
◼ Configuration of the element test device
Table 1 Configuration and functions of the element test device
Configuration

Integrated sampling
device for the test

FunctionsIntegrated sampling
equipment for the test

To make openings, collect and close byswitching the tool attached to the tip of therotating rod that operates in the verticaldirection.

Work dock for the test

To fix the integrated sampling device for thetest and align it with the upper surface of theadsorption tower.

Simulated adsorption
tower

To reproduce the superstructures of the
SARRY and the KURION adsorption towers.

Work stand for the test

To support the device during testing andprevent any falls. It is the scaffold usedduring work, such as switching the tools etc.,
around the integrated sampling device forthe test.

Work stand for the
test

Work dock for the
test

Simulatedadsorption tower

Fig.1 Configuration of the element test device➢ Since the aim is to perform the element tests in a cold environment, the boundary securing functions and theshielding functions will be omitted for every equipment, but the profile such as the shield thickness, which affects
the workability, will be reproduced.
©International Research Institute for Nuclear Decommissioning234

【Confidential】(b) Development of sampling technology- Preparation for element tests - Designing and manufacturing of the element test device (2/7) ◼ Integrated sampling device for the test
➢ Making openings, collecting and closing by switching the tool.
➢ The tool is switched manually through the glove port.
Table 1 Specifications of the integrated sampling device for the test
Items

Specifications

Basis of the specifications

1100 x 950 x 3200 mm

The external appearance is decided based on the assumed
crane lifting height, the necessary stroke, the workability
through the glove port, etc.

External appearance

Vertical
drive

Drive
methodVertical drive with ball screw

High positioning accuracy due to high rigidity.

Motorspecifications

Output: 0.75 kW
Torque: 20 Nm
Rotation speed: Up to 250rpm

Depends on the machine tool manufacturer's findings.
To suit the feed speed of the blade for making openings.

Stroke2000 mm

Depends on the machine tool manufacturer's findings.
To suit the cutting speed of the blade for making openings.

・Tool for opening(φ 60, 50, 40 mm)
・Cutting splinter sucking
function・Cutting oil spraying function

・Bore holes of multiple diameters forming stepped holes.
・Installation of a suction mechanism so that the cutting
splinters do not get mixed inside the adsorption tower.
・Installation of an oil spraying mechanism for cooling andlubrication.

Collection

Sampling tool
(Collection capacity max. 20
mL)

The collection capacity is set at 20 mL, since the target value
for collection is 10 to 20 mL.

ClosureInsertion of the closing plugwith the closure tool

The closing plug is remotely installed in the stepped holethat is formed.

Motorspecifications

Making openings

Suctioncollectionunit

It should be possible to insert the sampling tool till a depth of
100 mm from the adsorbent surface.

Output: 5 kW
Torque: 300 Nm
Rotation speed: Up to 150
rpm

Rotational
drive

Motor forrotational drive

Tools formakingopenings,
collection andclosureMotor forvertical drive

Glove port

Fig.1 Internal image of the integrated sampling
device for the test
©International Research Institute for Nuclear Decommissioning235

【Confidential】(b) Development of sampling technology- Preparation for element tests - Designing and manufacturing of the element test device (3/7) -◼ Detailed design of the integrated sampling device for the test (illustration)
COUP E C-C
ECHE LLE 1 : 2

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D

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E

B
517

102

505

100

112

101

522

123

117

512

126

513

125

115

127
107

514

107
116

108

109

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114

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124

519

519
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521

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518

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E

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505

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120

121

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514

C

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C

105

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119
108

107

502

101

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504

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106

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B

125

515

523

525

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509

520

507

510
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FO L I O : 1 /2

ECI - AQ- 120- A

F ORM AT : A0

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S-ISM
Asse mbly

T e l : 0 4 .9 0 . 2 8 .8 3 . 7 7

Etu d es Co n ce p tio n In g en ierie

e c i @ e c i -m e c a .c o m
8 4 1 1 0 Va i s o n -l a -Ro m a i n e

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e M a c h in e s s p e c ia le s

Ec h : 1 : 5

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Fig.1 Overall layout of the integrated sampling device for the test

F ORM AT : A0e c i @ e c i -m e c a .c o m
8 4 1 1 0 Va i s o n -l a -Ro m a i n e

ont age de M achines speciales

SIMPLIFIED MOCK-UP
S-ISM
Asse mbly
P1335-22-000

Ce d ocum ent

est la p r opr iet ede neATOpeut
X , ilet r e ut ils e, r epr od uitque
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ou m
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FO L I O : 2 /2

ECI - AQ- 120- A

523

Fig.2 Diagram of the partial assembly of the rod drive mechanism©International Research Institute for Nuclear Decommissioning236

【Confidential】(b) Development of sampling technology- Preparation for element tests - Designing and manufacturing of the element test device (4/7) ◼ Making openings: Flow of cutting oil: Collection flow of cutting splinters

Table 1 Functions of the tool
Requisite functions

Device functions

・Remotely operate the tool to perform cutting

Details of the A Part

120
0

To be able to remotely bore
holes on the KURION and the
SARRY adsorption towers at
positions where the sampling
tool will be inserted. (Fig. 1)(tool switching is a manual operation).
・Friction control and cooling by spraying the
cutting oil (Fig. 2).
・Collection of the cutting splinters and thecutting oil by suction (Fig. 2).
・Holding mechanism for controlling the shakingof the rod core.

Positions ofmakingopenings
φ50-60Counter-bored hole

Collection of
the cutting
splinters andoil by suction

Holdingmechanism on thework dock-side
Position of
making an
opening

φ 50-60
Counter-bored hole
Lower welding plate φ 40

Cutting blade

Fig.1 Positions of making openings and oepnings sectional view
(Above: KURION adsorption tower, Below: SARRY adsorption tower

Details of
the B Part

Oil spraying

Fig.2 Illustration of opening tool
©International Research Institute for Nuclear Decommissioning

237

(b) Development of sampling technology【Confidential】- Preparation for element tests - Designing and manufacturing of the element test device (5/7) -◼ Tool for making openings
Tool ƒÓ40 on extensi on tube

Tool samplin g on ext ension t ube

Tool ƒÓ50 on extensi on tube

From the cutting oil spraying
unit

Tool ƒÓ60 or plug on extensi on tube

Motor for rotational drive

501
501

501

107

501

105

106

107

103

104

Rod coupling

107

107

Quantity of
injected oil
Speed of injection

Supplied air pressure

To the suction-collection unit
101

100

Rod for making openings
103

Blade for making
openings

108
P lug interfa ce
106

102
P lug
S ee : P 1335- 26-020

SuctionpressureSuction flowrate

Supply of cuttingoilCollection of
cutting powderA
I NDI CE

2 7 /1 1 /2 0 1 9AGR
DATE

DESS I NE

BV E

PGO

VERI FI E

APPR O UVE

F i rs t e d i ti o n
O BSE RVATI O NS

ATOXS amp ling too l
S ee : P 1335- 26-010

N‹ :E.C.I. MECA
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Asse mbly
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est la p r opr iet e de ATO Xne, il peut et r e ut ils e, r epr od uit , com
quem uni
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Fig.1 Tool design for making openings

FOLI O : 1/1

E C I - A Q - 120- A

Co n c e p t i o n , Re a l i s a t i o n t aegt Me d e M a c h in e s s p e c ia le s

E c h : 1 :1

The cutting oil supplied by the "Cutting oil supply unit" issupplied to the tip of the blade via the center of the rod, and thiscutting oil along with the cutting powder gets sucked into thesuction-collection unit from the outer side of the rod.
Fig.2 Configuration of the mechanism for making openings©International Research Institute for Nuclear Decommissioning238

【Confidential】(b) Development of sampling technology- Preparation for element tests - Designing and manufacturing of the element test device (6/7) -◼ Collection (Sampling)➢

Samples from inside the adsorption tower will be collected by inserting a sampling tool through the bored holes (φ 50)

Table 1 Collection (sampling) tool functions
Requisite functions
It should be able toremotely collect 10 to
20 mL sample fromthe vicinity of thesurface

Device functions
Collection position(Fig. 1)
Sampling tool(Fig. 2)The stroke is set such that samples can be collected from the surface (in the vicinity of
100 mm) of the adsorption tower.
Stroke the sampling tool by rotating it clockwise (closing the sampling port), and when it
reaches the sampling position, rotate it counterclockwise to open the sampling port and at
the same time slide out the vane and scrape off the adsorbent for sampling. (Collection
port volume: Approx. 20 mL).

Approx.100

Vane

Collection
port closed

Surface layer
Collection position(a) SARRY(b) KURION
Fig.1 Sample collection position

Collection
port open

Fig.2 Vane opening-closing type sampling tool
©International Research Institute for Nuclear Decommissioning

239

(b) Development of sampling technology【Confidential】- Preparation for element tests - Designing and manufacturing of the element test device (7/7) ◼ Closure◼
◼

To perform the closure operation by inserting the closing plug in the bored holes after the collection is over (Fig. 1, 2).
During routine storage, no pressure is applied, but it is assumed that pressure is applied on the adsorption tower due tothe method of the pulling out operation. Therefore, studies were conducted for a closing plug with the same pressure
resistance performance as the adsorption tower.

Table 1 Functions of the closing tool
Requisite functions

Device functions

Installation
methodBy rotating the shaft while pressing the closing plug into the
stepped hole, the steel spring is extruded, the polyurethanepart is expanded, and the bored hole is closed. The steel
spring can also be removed by rotating the shaft in thereverse direction.

Specifications

・Max. pressure: Approx. 1.37 MPa (Maximum Working
Pressure of the adsorption tower)
・ Closing port diameter: φ 50 mm
・Compatible with both KURION and the SARRY adsorption
towersIt should be possible to close thebored holes remotely

Closing rod (Also acts as the rod
for φ 60 hole)

Rotating
shaft

Conversion connector: There isa 12.7 [mm] angular ball lockbetween the closing plug andthe connecting part due to which
it is possible to remotely
separate the plug

Steel spring

Closing plug

Tapering
part
Tool part(a) Structure of the closure tool

Polyurethane(for testing byhardness)(b) Structure of theclosing plug

Fig.1 Closing plug (left: image, middle: disassembly image, right: prototype)Fig.2 Installation of the closing plug
(Left: KURION, Right: SARRY)©International Research Institute for Nuclear Decommissioning240

【Confidential】(b) Development of sampling technology- Preparation for element tests - Preliminary test for making openings ◼ Preliminary test for making openings➢ Preliminary test for making openings of a steel plate was carried out by installing a rod for opening anda prototype blade for the preliminary test on the machining center (Fig. 1).
➢ It was found that it was possible to reduce the size of the cutting splinters (approx. 5 mm or less) byshaping the blade for making openings.
➢ It was confirmed that it was possible to improve the accuracy of the bored holes by installing the holding
mechanism.
➢ An insight was obtained on the bored holes from the preliminary test results, and this will be reflected in
the element test device.

Vertical machining
centerPrototype holding
mechanism

Opening rod for the
preliminary test (Approx. 700
mm)

Prototype blade

Preliminary test equipment formaking openings

Appearance of preliminary
test for making openings

Installation of holding
mechanism

Condition of the bored holes

Fig.1 Preliminary test for opening©International Research Institute for Nuclear Decommissioning241

【Confidential】(b) Development of sampling technology- Preparation for element tests - Preliminary test for collection (1/2) ◼ Preliminary test for collection➢ The preliminary test for collection was carried out with the element test device for collection which had been
manufactured in the previous years (Fig. 1).
➢ The vane opening-closing sampling tool was newly manufactured in order to deal with the issue of insufficientamount of collection encountered during the previous years’ preliminary test for collection.
➢ During the previous years’ test, only one kind of property was set for the collection target (zeolite). Presuming an
actual test, the preliminary test for collection was executed for multiple test targets (Fig. 2).
Motor forrotational driveCollection
port

Motor forvertical drive

Sampling head
# Executedwith 4 types

Test
tower
SimulatedsampleTest targets①

Itaya zeolite Z-05 Particle size 0.5 to 1 mm Dry②

Itaya zeolite Z-12 Particle size 0.5 to 2 mm Dry③

Itaya zeolite Z-13 Particle size 1 to 3 mm Dry④

Itaya zeolite Z-05 Particle size 0.5 to 1 mm Moist⑤

Itaya zeolite Z-12 Particle size 0.5 to 2 mm Moist #⑥

Itaya zeolite Z-13 Particle size 1 to 3 mm Moist#: Samples collected during the previous years’ tests

Fig.1 Element test device for collection

Fig.2 Test targets©International Research Institute for Nuclear Decommissioning242

【Confidential】(b) Development of sampling technology- Preparation for element tests - Preliminary test for collection (2/2) ◼ Results of the preliminary test for collection
The results of the preliminary test for collection are shown below (Fig. 2) using various sampling tools (Fig. 1).
The optimal values from the previous years’ tests were used for the operating conditions during collection.
The vane opening-closing typehad satisfactory collection
under all conditions

22.00

All the collected amounts ofthe dry sample weresatisfactory

20.00
18.00

乾燥
Dry

16.00

Moist
湿潤

SpiralFlat

Fixed vane type

Vane opening-closing type

採取量 [g] = [ml]Collection amount

14.00

12.00
Targetcollection amount
採取量の目標値

10.00

10 [g] = 10 [ml]※かさ密度:1.0[g/ml]*Bulk
density

8.00
6.00
4.00

Operating conditions
Collection rotation speed: 50 rpm
Feed speed: 20 mm/min

2.00

Fig.1 Various sampling tools

0.00
Dry
Moist
Moist
Moist
乾燥Moist湿潤 Dry
乾燥Moist湿潤 Dry
乾燥Moist湿潤 Dry
乾燥Moist湿潤 Dry
乾燥Moist湿潤 Dry
乾燥Moist湿潤 Dry
乾燥Moist
湿潤Dry
乾燥Moist
湿潤Dry
乾燥Moist
湿潤Dry
乾燥
湿潤Dry
乾燥
湿潤Dry
乾燥
湿潤
Z-05

Z-12
スパイラル
Spiral

Z13

Z-05

Z-12

Z13

Z-05

Z-12

Fixedベーン固定式
vane type
Flat
Toolツール形状および吸着材
shapes and adsorbents
フラット

Z13

Z-05

Z-12

Z13

Vane opening-closing
ベーン開閉式 typeFig.2 Results of the preliminary test for collection➢

It was possible to collect sufficient amounts of sample with all the tool forms since the dry sample had a high fluidity.

➢

Since the moist samples had a high adhesiveness, the samples could not be collected in a stable manner, except in case of the vane opening-closing type of tool.

➢

The vane opening-closing type had satisfactory collection under all the sample conditions.
However, zeolite was found to be caught in the opening and closing mechanism of the vane opening-closing type. Moreover, since there is a risk of interference from ordamage to the vane at the time of withdrawing the tool from the bored holes, the shape will be further reviewed and study on the risk reduction measures will continue.

©International Research Institute for Nuclear Decommissioning243

【Confidential】(b) Development of sampling technology- Preparation for element tests - Preliminary test for closure ◼ Preliminary test for closure➢ The preliminary test for closure was carried out by mounting the prototyped closing plug on the pressure
resistance tester (Fig. 1).
➢ The closing plug was mounted with a manual tool, pressure was applied with a hydraulic pump and thepressure values were noted for the leakage from the closing plug.
➢ It was confirmed that with respect to the maximum working pressure of 1.37 Mpa for the adsorption tower,
there was no leakage from the prototyped closing plug even at 1.4 MPa.
➢ A study will be conducted for the optimization of the material and structure with an aim to developing aclosing plug which will not leak in a pressure resistance test for approx. 2 Mpa, which is 1.5 times of 1.37
MPa.

Prototyped closing plug

Preliminary test for closure

Mounting status of the closing
plug

Closing plug when it has
risen up due to internalpressure

Fig.1 Preliminary test for closure©International Research Institute for Nuclear Decommissioning244

【Confidential】(b) Development of sampling technology- Summary (1/2) -Table 1 Outline of the collection conditions◼ Overview of the results and issues in the current year
Collection targets➢ Collection conditions and adsorption tower selection policy
The collection conditions and the selection policies for adsorption towers (KURION,SARRY) were studied (Table 1). Note that these will be reviewed based on the future
results of analysis.

➢ Selection of the element test items

The element test items for the series of operations namely, making openings, collection andclosure, were selected.

The contents of the element tests for a series of work processes (making openings,
collection and closure of the adsorption tower) were studied assuming the actual work
conditions. Moreover, in order to conduct the test, the designing of the element test device
was completed and its manufacturing was started (Fig. 1 and 2).

Overall layout

Integrated sampling device
for the test

Fig.1 Element test device

Work dock for the test

Collection conditions➢ Design of the element test device

Sampling tool

Items

Specifications

Types

Priority to the adsorption tower used during thefirst stage of operations
Especially, priority for the upstream-sideadsorption tower

Structure

KURION: Compatible with all adsorption towers
SARRY： Flange typeNo. of sampling
points

One point each

Depth

Depth of 100 mm from the surface of theadsorbent

Amount10 to 20 mL

Opening
diameter

φ 50 mm

Closing plug

Pressure resistance at 1.37 MPa (short term),
Detachable

Closing plug

Simulated adsorption tower(KURION)

Fig.2 Components of the element test device©International Research Institute for Nuclear Decommissioning245

【Confidential】(b) Development of sampling technology- Summary (2/2) ◼ Overview of the results and issues in the current year
➢ Execution of preliminary tests•••

Preliminary tests were conducted for making openings, sample collection and closing of the opening ( Fig. 1 to Fig. 3).
A perspective of an opening method was obtained from the preliminary test.
There will be an ongoing study regarding the methods of collection and closure.
The vane openingclosing type hadsatisfactory collection
under all conditions

22.00
20.00
18.00

乾燥

16.00

湿潤

採取量 [g] = [ml]

14.00

12.00
採取量の目標値
10 [g] = 10 [ml]※かさ密度:1.0 [g/ml]

10.00
8.00
6.00
4.00
2.00
0.00
乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤 乾燥 湿潤
Z-05

Z-12
スパイラル

Z13

Z-05

Z-12

Z13

Z-05

フラット

Z-12
ベーン固定式

Z13

Z-05

Z-12

Z13

ベーン開閉式

ツール形状および吸着材

Fig.1 Preliminary test for making
openings

Fig.3 Status of the preliminary test forclosure➢ Risk evaluation••

The risks involved in a series of work processes were identified. The risks were evaluated through the element tests.
The exposure during the series of work processes was evaluated. The validity of the assumed working hours was evaluated through theelement tests.

◼ Overview of the plan for the following year
➢
➢

➢
➢

The sampling tool will be improved based on the results of the preliminary test for sample collection.
The preparations for the element tests will be completed after manufacturing the element test equipment and simulated adsorption
towers etc., which are required for the element tests.
The series of element tests for making openings in the adsorption tower, collecting samples and closing an opening of the adsorption
tower will be conducted.
The possibility of applicability for collection methods and equipment that are applied at the site will be provided based on the test results.

©International Research Institute for Nuclear Decommissioning246

【Confidential】

2. Project Details
d. Integration of R&D Results©International Research Institute for Nuclear Decommissioning247

【Confidential】

d. Integration of R&D- Study on waste stream Input◼ Results
 By FY2018, methods were developed for integrative evaluationof consistency in results, progress status and pending issues
by reflecting the latest results obtained through previous
research (Fig. 1). Further, consistency in results, progressstatus and pending issues were presented in these developed
methods.
◼ Goals
 To incorporate the results obtained in other researches into
waste stream and perform integrative evaluation of consistency
in results, progress status and pending issues.
◼ Implementation details (Details are shown on the next page) In FY2019, options were developed for treatment methods fromthe perspective of both treatment and disposal, by reflecting
into the waste stream the requirements on the basis of the R&D
results of disposal (waste conditioning, waste preliminarystorage) as also R&D results of treatment (pre-treatment,
treatment, re-processing). Further, the entries in the Input
compilation table were revised for each target waste (Fig.2).
 Waste stream will continue to be updated in FY2020 byincorporating the issues and R&D results obtained in FY2019
and the results of evaluating progress status, consistency inresults and pending issues will be presented.
◼ Indicators for determining goal achievement
 Waste stream should be updated. (FY2020) The input compilation table should be updated. (FY2020)

Details of study

OutputEnsuring completenessConsolidate required
information and issues(Treatment flow)

Waste-wise input
compilation table

Information required forthe study conducted ateach step, and issues

Research results
Mid-and-long-termroad-map
TEPCO HD Storage& Management Plan

Refining the timelines
Set the study time at eachstep for each type of waste

Organizing the priority
order and methods
Confirm the overall pictureof waste management

Time axis
compilation table

Waste stream (overall
picture of waste
management)

Fig. 1 Overview of the study on the waste stream

 Describe the information input from upstream to downstream.
For example) Present the amount of generated raw waste, raw
waste specifications, and container specifications as information
input from “waste” to “pre-treatment, treatment, re-processing”①Describe the details studied in each step.
For example) Processing capacity andprocessing methods were studied in “pretreatment, treatment, re-processing”.

 Describe the information input from downstream to upstream. For
example) Present the necessity of level classification, the allowableamount of gas generation, and the study of the effect of materials
considered to be harmful to disposal, as information input from“disposal, re-usage" to "pretreatment, treatment, re-processing".

Fig. 2 Example of input compilation table©International Research Institute for Nuclear Decommissioning248

【Confidential】

d. Integration of R&D- Background and goals Background
1.

Various kinds of waste generated in Fukushima Daiichi and each has its own requirements from a safety perspective
effected on the site. In addition, properties of some waste are uncertain.
✓

2.

These wastes should be processed and stored safely and it is necessary to provide a path for the disposal of thesewastes.
✓3.

As of end of FY2018, IRID has prepared a list of 890 types of solid wastes inclusive of already existing wastes and wastes
expected to be generated in the future. There may be a further increase in types of wastes.

It has been decided to have a “Technical perspective concerning the policy on treatment, disposal and its safety” in the Mid-andLong-Term Roadmap by around 2021.
① In this connection, it is necessary to develop the flow (waste stream) from the generation of solid waste, to its
storage and treatment (segregation, volume reduction, stabilization etc.) and eventually to long term storage or
disposal.
② In this subsidy project, we are working towards narrowing down the waste streams having multiple options
depending on the characteristics of wastes and various efforts are being made to identify issues.

Meanwhile, it requires a lot of time to complete research pertaining to characterization of wastes and safe methods for
waste treatment and disposal.
✓ It is necessary to conduct R&D in a reasonable and efficient manner.
① In this subsidy project, we are striving to promote streamlining and efficiency of overall R&D while providing
information feedback mutually for each and every aspect of R&D. (Role of “Integration of R&D results”)

Target setting for integration of R&D results in this subsidy project•

The following targets will be presented toward the Mid-and-Long-Term Roadmap Third Phase(From FY2021 onwards).
✓ Updated input compilation table and proposal for waste streams as of then
✓ Evaluation results of the issues to be addressed when developing a waste stream©International Research Institute for Nuclear Decommissioning249

【Confidential】

d. Integration of R&D- Overall integration flow (formulation of waste stream) in this subsidy project -#The blue callouts are the main items studied in FY2019

Input

Contents for study
廃棄物の発生前処理・処理・再加工

一次保管

保管

廃棄体化

廃棄前保管

Organize necessary
information and issues

処分・再利用（安定化処理）前処理・処理・再加工

汚染区分：XI-4
一次保管

廃棄物の発生保管

廃棄体化
Ａ（安定化処理）廃棄物の発生R&D results (inventory etc.)

汚染区分：XI-4
一次保管

廃棄物の発生

多核種 HIC
前処理・処理・再加工（スラリー）（安定化処理）汚染区分：XI-4
一次保管

多核種 HICHIC 保管
前処理・処理・再加工（スラリー）※1

HIC 保管

基本フロー（堰設置等）

漏洩拡大防止（堰設置等）

Ｃ

基本フロー

水分
低減

※2
HIC 残材

水分低減※2
HIC 残材

固型化
解体・分別
金属
Ｅ
（SUS 補強体等）

保管

HIC 残材

保管

解体・分別
金属

保管
除染

保管 除染

（SUS 補強体等）
金属

核種本
455

発生時期

10 本
発生理由

外形

処理水から評価
発生量

復旧対策

2014年12月末
8 本
サンプル数

455 本
核種本
10

1 本
汚染水処理

94 本
6 本
処理水から評価-

濃度(Bq/g)
本
-

○

本

線量
セシウム吸着装置

本

○

33○本

-

本

○

2○本
α核種検出
94○本

-

本
参照文献
本

○

廃棄物リスト

6○本

3.3E+14
Bq/カラム
処理水から評価
0○本
本 Cs137
号機、
構内エリア

発生場所 -

本

廃棄物名称

H

保管
保管場所

可燃物

保管形態
吸着塔
考え方

第二仮保管施設
化学組成
SMZ

含水率

一時保管施設（第一施設）
有害物等
一時保管施設（第四施設）

0 本
Cs137
0 本

留意事項
考え方
留意事項

仮保管施設

○

TSG

ろ過砂
第二仮保管施設
1
廃棄物名称

-

本

○

高温焼却炉建屋

11
15

*2

15

※8
14

60

14

137

10

容器保管
8

廃棄体化

再加工 ・線量(中) ；<10 （Bq/t） Co，<10 （Bq/t） Cs※6・線量(低)；検討が必要
Ｇ
60

137

保管
再利用

Ｆ

再加工
可燃物

減容

（PE 容器等）

減容

保管

保管

Ｇ
容器保管

減容

※6

廃棄体化

減容

※6

Ｇ

保管

容器保管

Ｇ’

減容

容器保管

※7
Ｇ’
容器保管

※9

※7

廃棄体化

容器保管

※9

※6

保管

※6

保管
再利用

廃棄体化 廃棄確認

廃棄体化
廃棄確認

廃棄確認

各ステップへのインプットする情報
廃棄物の発生再利用

考え方

廃棄物の発生

ピット処分

※7

保管
ピット処分

保管
ピット処分

※7

保管
ピット処分

廃棄確認

廃棄確認

ピット処分

ピット処分

減容

－

－

れば再資源化。(※9)
【個別事項】
－

保管形態

TSG

Challenges in andinformation needed to
study each stage of thestream

第二仮保管施設

ろ過砂

第二セシウム吸着装置 IE-96
IE-911
EN-101HS-726

減容性
ガス発生の程度

漏えい・飛散の程度
取出し・物流ルート
核種組成・核種濃度（検認、処
廃棄体化
廃棄前保管
処理した廃棄物（性状・量）
分の観点）
取出し・物流ルート
材質・化学組成
二次廃棄物（性状・量）減容性
線量率（敷地境界の観点）

前処理・処理・再加工

保管

廃棄体化

減容性
ガス発生の程度
漏えい・飛散の程度
取出し・物流ルート
廃棄体化
廃棄前保管
処理した廃棄物（性状・量）減容性
ガス発生挙動漏えい・飛散の程度
処分・再利用検認シナリオ・方法

二次廃棄物（性状・量）
減容性
減容性
ガス発生挙動ガス発生の程度
漏えい・飛散の程度
処分・再利用漏えい・飛散の程度
検認シナリオ・方法

廃棄前保管

処分・再利用処分・再利用
（未設定）
・処分区分
・廃棄体技術基準

取出し・物流ルート

廃棄体を健全に保つ観点での保管への要求技術基準への適合レベル区分の必要性の有無
核種濃度制約
発熱量制約

望ましい保管期間

レベル区分の必要性の有無
ガス発生許容量
処分に有害と考えられる物質の影響有無検討

望ましい分別程度
廃棄物管理を具体化するための放射能濃度Matters and details thatneed to be studied,status of initiatives,
necessary actions

検認の観点からの調査要求
処分区分の観点からの調査要求（処分影響物質、安全評
価用ソースターム設定、等）処分・再利用Set the study time for eachstep of each waste
Review of timeline compilation
table
廃棄物分類

Reflecting the progress of
R&D

廃棄物の発生一時保管

S1_圧力容器ー（未設定）
S2_格納容器金属ー（未設定）
S3_格納容器コンクリートー（未設定）
S4_建屋内金属
ー（未設定）
S5_建屋内コンクリートー（未設定）
仮設処理装置、タンク S6_瓦礫金属
ー
（発生中）
類
S7_瓦礫コンクリート
ー（発生中）
可燃性雑固体、伐採木 S8_可燃物
ー（発生中）
S9_吸着塔１
ＫＵＲＩＯＮ，ＳＡＲＲＹ （発生中）
S9_吸着塔２
モバイル浄化装置（発生中）
S10_多核種除去装置３－ スラリー
（発生中）
S10_多核種除去装置３－ 吸着材（発生中）
水処理２次廃棄物
S10_多核種除去装置４－ 処理カラム（発生中）
S11_除染装置スラッジ
ＡＲＥＶＡ（発生終了）
S12_フィルタb
高性能多核種（発生中）
S13_濃縮廃液
エバポ濃縮廃液（発生停止中）S14_デブリ取出し前の発生廃
横アクセス
至近の３か年
棄物(横アクセス)
デブリ取出し前の発生廃棄物
S14-1_デブリ取出し前の発
縦アクセス
至近の３か年
生廃棄物(縦アクセス)

（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（保管中）
（保管中）
（保管中）
（保管中）
（保管中）
（保管中）
（保管中）
（保管中）
（保管中）
（発生中）
（発生中）

汚染土壌等

（未設定）
（未設定）
（未設定）

解体に伴って発生する可能性のある廃棄物

前処理・処理・再加
工

保管

廃棄体化

（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（計画済み）（計画済み）（未設定）（計画済み）（計画済み）（未設定）（焼却処理中）
（保管中）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
至近３か年のうち早期 至近３か年のうち早期 （未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
至近３か年のうち早期 至近３か年のうち早期 （未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）

Timeline compilation
table

金属（ＳＦＰ、サイトバンカ
保管物等）
樹脂

保管廃棄物

廃棄前保管

Period when the study ofeach stage of waste willbe essential (expected)

処分・再利用

（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）

（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）
（未設定）

至近の３か年（要否含めて）至近の３か年

至近の３か年

（未設定）

（未設定）

（未設定）

至近の３か年（要否含めて）至近の３か年

至近の３か年

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

（未設定）

Current processing flow
proposal and integrated
stream proposal

Mid-and-Long-Term
Roadmap
廃棄物の発生前処理・処理・再加工

一次保管

保管

廃棄体化

廃棄前保管

Check the overview of waste
management

処分・再利用（安定化処理）廃棄物の発生前処理・処理・再加工

汚染区分：XI-4
一次保管

保管（安定化処理）廃棄物の発生汚染区分：XI-4
一次保管

多核種 HIC
前処理・処理・再加工（スラリー）

多核種 HICHIC 保管
前処理・処理・再加工（スラリー）（安定化処理）多核種 HIC

汚染区分：XI-4（スラリー）

多核種 HICHIC 保管（スラリー）
基本フロー

※1

基本フロー

基本フロー
HIC 残材保管

TEPCO HD
Storage andmanagement plan, etc.

廃棄体化
Ａ

廃棄前保管

処分・再利用※7

HIC 保管
保管
廃棄体化
廃棄体化
廃棄前保管
処分・再利用※3 Ａ※7
漏洩防止拡大保管
廃棄確認
HIC Ｃ
保管
保管
廃棄体化
廃棄前保管
処分・再利用Ｂ※7
容器保管
漏洩防止拡大保管
廃棄確認（堰設置等）
処分
Ｃ水分
HIC 保管 ※2
HIC 保管
保管
廃棄体化
※6
低減
Ａ
Ｂ
※3
※8
Ｄ
取出し
※7
容器保管
漏洩拡大防止基本フロー
処分※1
漏洩防止拡大保管
廃棄確認（堰設置等）*1
水分
HIC 保管
【線量区分値】政令濃度上限値，埋設規則*2 から設定
廃棄体化
※6
固型化保管
Ｃ
※2
低減
11
※3
※8
Ｂ
・線量(ＴＲＵ相当)*1；>10 （Bq/t）α濃度Ｄ
漏洩拡大防止
取出し
容器保管
15
14
*2
60
Ｅ
漏洩防止拡大保管*1
処分
廃棄確認（堰設置等）・線量(高Ｌ１相当)；>10 （Bq/t）
Co，>10 （Bq/t）137Cs【線量区分値】政令濃度上限値，埋設規則*2 から設定
固型化
水分
Ｃ ※2
※6
15
14
*2
11
；<10 （Bq/t）60Co，<10 （Bq/t）137Cs
保管
HIC 残材 低減
・線量(ＴＲＵ相当)*1；>10 ・線量(高)（Bq/t）α濃度
Ｂ 解体・分別
※8
Ｄ
10
8
取出し
容器保管
*2
60
15
60 ；<10 14
Co，<10（Bq/t）137Cs
Ｅ
・線量(高Ｌ１相当)*2；>10 ・線量(中)（Bq/t）
Co，>10 （Bq/t）（Bq/t）137
Cs
処分
*2
水分
【線量区分値】政令濃度上限値*1，埋設規則
から設定
固型化
※6
15
14
※2
・線量(低)；検討が必要*2
Ｆ・線量(高)；<10（Bq/t）60Co，<10 （Bq/t）137Cs
低減
保管
解体・分別
11
HIC 残材※8 8・線量(ＴＲＵ相当)*1；>10 （Bq/t）α濃度
Ｄ
10
*2
60
・線量(中)；<10 （Bq/t）
Co，<10 （Bq/t）137Cs
除染
再加工
再利用
15
14
*2
Ｅ
*1
・線量(高Ｌ１相当)；>10（Bq/t）60Co，>10 （Bq/t）137Cs【線量区分値】政令濃度上限値，埋設規則*2から設定
固型化
・線量(低)；検討が必要
Ｆ 15
14
*2
※9
※411
※7
；<10 （Bq/t）60Co，<10 （Bq/t）137Cs*1 ・線量(高)保管
解体・分別・線量(ＴＲＵ相当)；>10 （Bq/t）α濃度
金属
保管
容器保管
廃棄体化
減容
保管
10
8
除染
*2
60
再加工・線量(中)再利用15
14；<10 （Bq/t）
Co，<10 （Bq/t）137Cs
Ｅ
・線量(高Ｌ１相当)*2；>10 （Bq/t）60Co，>10 （Bq/t）137Cs（SUS 補強体等）※61514・線量(低)；検討が必要※9137
Ｇ※4・線量(高)*2Ｆ※7；<10 （Bq/t）60Co，<10 （Bq/t）
Cs
解体・分別
金属
保管
容器保管
廃棄体化
減容
保管
10
8
除染
再加工 ・線量(中)*2；<10 （Bq/t）60Co，<10 （Bq/t）137Cs
再利用
廃棄確認
ピット処分
（SUS 補強体等）※6・線量(低)；検討が必要
Ｆ
Ｇ
※9
※4
※7
※7
保管 除染
容器保管
廃棄体化
減容
保管
再加工
再利用
保管
廃棄確認
容器保管
廃棄体化
ピット処分
減容
保管
可燃物
※6
（PE 容器等）
※9
※6
Ｇ
※4
※7
※7
Ｇ’
容器保管
廃棄体化
減容
保管
保管
容器保管
廃棄体化 廃棄確認
減容
保管
可燃物
ピット処分
廃棄確認
ピット処分
※6
（PE 容器等）
※6
Ｇ
Ｇ’
※7
保管
容器保管
廃棄体化 廃棄確認
減容
保管
ピット処分
廃棄確認
ピット処分
【共通事項】【共通事項】【共通事項】【共通事項】【共通事項】※6
・ＨＩＣ容器内で保管。
－
・必要な場合は廃棄体化処理用の容器 ・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲ
※7
Ｇ’
(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
容器保管
廃棄体化
減容
保管
考え方
【個別事項】・スラリーの取出し。(※2)(※6) 【共通事項】 廃棄確認管。(※7)ピット処分 【共通事項】として扱う前提。(※8)
【共通事項】【共通事項】【共通事項】
・漏えいリスクの低減。(※3)
【個別事項】※6－ －・ＨＩＣ容器内で保管。
・必要な場合は廃棄体化処理用の容器
・スラリーは処分に適合するよ【個別事項】
・α 核種を含む廃棄物はＴＲ ・ＨＩＣ外側容器は可能であ
留意事項
Ｇ’
・ＨＩＣ外側容器は切断し保管。必要に－－れば再資源化。(※9)
HIC 保管

保管（安定化処理）汚染区分：XI-4
一次保管

廃棄物の発生HIC 残材

Fig. 1 Overall integration flow

減容性
ガス発生の程度
漏えい・飛散の程度
取出し・物流ルート
核種組成・核種濃度（検認、処
廃棄体化
廃棄前保管
分の観点）
処理した廃棄物（性状・量）
取出し・物流ルート
材質・化学組成
減容性
二次廃棄物（性状・量）
線量率（敷地境界の観点）
ガス発生挙動取出し・物流ルート
減容性
漏えい・飛散の程度

Input compilation tablefor each waste

R&D results (treatment etc.)

Unification of input
information byusing the List of
Wastes and Waste
Management Sheet

一時保管

核種組成・核種濃度（検認、処分の観点）
材質・化学組成（検認、処分の観
前処理・処理・再加工 取出し・物流ルート
保管
点）
減容性
線量率

Feedback on disposal studies(WasteManagement
Sheet)

一時保管

核種組成・核種濃度（検認、処分の観点）
材質・化学組成（検認、処分の観 取出し・物流ルート
保管点） 前処理・処理・再加工
線量率

Review of input compilation table
廃棄物の発生

各ステップへのインプットする情報

留意事項

一時保管

核種組成・核種濃度（検認、処分の観点）
材質・化学組成（検認、処分の観 取出し・物流ルート点） 前処理・処理・再加工
保管
線量率

処分・再利用廃棄前保管（未設定）（未設定）
取出し・物流ルート
減容性
・保管場所
・処分区分
ガス発生挙動取出し・物流ルート
・保管方法
・廃棄体技術基準
漏えい・飛散の程度処理した廃棄物（性状・量）・線量率（敷地境界の観処分・再利用検認シナリオ・方法
二次廃棄物（性状・量）点）
減容性
廃棄前保管（未設定）
ガス発生挙動（未設定）
減容性
取出し・物流ルート
廃棄物の発生
一時保管
処分・再利用ガス発生の程度
・保管場所
漏えい・飛散の程度
・処分区分
減容性
ガス発生挙動廃棄体を健全に保つ観点での漏えい・飛散の程度
取出し・物流ルート
・保管方法
減容性の観点での要求
検認シナリオ・方法
・廃棄体技術基準
核種組成・核種濃度（検認、処核種組成・核種濃度（検認、処ガス発生の程度
漏えい・飛散の程度
保管への要求処分・再利用処理した廃棄物（性状・量）・線量率（敷地境界の観ガス発生の程度分の観点）
廃棄体化（未設定）
ガス発生の観点での要求分の観点）
検認シナリオ・方法
取出し・物流ルート
漏えい・飛散の程度
取出し・物流ルート（未設定）
材質・化学組成（処分の観点）
材質・化学組成（検認、処分の観 取出し・物流ルート
二次廃棄物（性状・量）点）
漏えい・飛散の程度
取出し・物流ルート
・廃棄体化技術
漏えい・飛散の観点での要求
廃棄前保管（未設定）処理した廃棄物（性状・量）
・処分区分
減容性
線量率
点）
核種組成・核種濃度（検認、処処理した廃棄物（性状・量）検認の観点での要求
ガス発生の程度
・保管場所
二次廃棄物（性状・量）検認シナリオ・方法
・廃棄体技術基準
減容性
線量率
分の観点）
ガス発生挙動廃棄体を健全に保つ観点での二次廃棄物（性状・量） 取出し・物流ルート
減容性の観点での要求技術基準への適合漏えい・飛散の程度
取出し・物流ルート
・保管方法
核種組成・核種濃度（検認、処ガス発生の程度
材質・化学組成保管
漏えい・飛散の程度
保管への要求分の観点）
廃棄体化（未設定）
ガス発生の観点での要求レベル区分の必要性の有無処理した廃棄物（性状・量）・線量率（敷地境界の観減容性
線量率（敷地境界の観点）
取出し・物流ルート
取出し・物流ルート
減容性
廃棄前保管（未設定）漏えい・飛散の程度（至近３か年で早期）減容性の観点での要求
検認シナリオ・方法
材質・化学組成（処分の観点）
・廃棄体化技術
漏えい・飛散の観点での要求
核種濃度制約
ガス発生の程度
減容性
二次廃棄物（性状・量）点）
核種組成・核種濃度
核種組成・核種濃度（検認、処処理した廃棄物（性状・量）
ガス発生の程度
・保管場所
・保管場所
ガス発生の観点での要求
線量率
材質・化学組成（保管の観点）
検認の観点での要求
発熱量制約
分の観点）
ガス発生挙動廃棄体を健全に保つ観点での取出し・物流ルート
漏えい・飛散の程度
望ましい保管期間
二次廃棄物（性状・量）
取出し・物流ルート
漏えい・飛散の程度
取出し・物流ルート
・保管方法
・保管方法
漏えい・飛散の観点での要求減容性
線量率（敷地境界の観点）
材質・化学組成
漏えい・飛散の程度
保管への要求処理した廃棄物（性状・量）減容性の観点での要求技術基準への適合
核種組成・核種濃度（検認、処処理した廃棄物（性状・量）・線量率（敷地境界の観・線量率（敷地境界の観検認の観点での要求
保管
ガス発生の程度
線量率（敷地境界の観点）
検認シナリオ・方法
減容性分の観点）
廃棄体化（未設定）
ガス発生の観点での要求 二次廃棄物（性状・量）レベル区分の必要性の有無
二次廃棄物（性状・量） 取出し・物流ルート
点）
点）（至近３か年で早期）減容性の観点での要求漏えい・飛散の程度
取出し・物流ルート
材質・化学組成（処分の観点）
ガス発生の程度
核種組成・核種濃度・廃棄体化技術
漏えい・飛散の観点での要求 核種濃度制約
核種組成・核種濃度（処理の観・保管場所
ガス発生の観点での要求処理した廃棄物（性状・量）
線量率
材質・化学組成（保管の観点） 取出し・物流ルート
漏えい・飛散の程度
望ましい保管期間
望ましい保管期間
減容性
前処理・処理・再加工
減容性の観点での要求検認の観点での要求
発熱量制約
点）
・保管方法
漏えい・飛散の観点での要求 レベル区分の必要性の有無
減容性の観点での要求技術基準への適合二次廃棄物（性状・量）
核種組成・核種濃度（検認、処線量率（敷地境界の観点）処理した廃棄物（性状・量）
ガス発生の程度
材質・性状・化学組成（同上）（至近３か年で早期）ガス発生の観点での要求・線量率（敷地境界の観検認の観点での要求分の観点）
廃棄体化（未設定）
ガス発生の観点での要求レベル区分の必要性の有無
取出し・物流ルート
漏えい・飛散防止への要求
ガス発生許容量
保管
形状・寸法（同上）
二次廃棄物（性状・量）
取出し・物流ルート
漏えい・飛散の程度
取出し・物流ルート
・処理目的
漏えい・飛散の観点での要求
減容性 ・廃棄体化技術
材質・化学組成（処分の観点）点）
漏えい・飛散の観点での要求
核種濃度制約
再利用用途（至近３か年で早期）線量率（作業、敷地境界の観
減容性の観点での要求処理した廃棄物（性状・量）・処理技術
検認の観点での要求
ガス発生の程度
核種組成・核種濃度
線量率
核種組成・核種濃度（処理の観検認の観点での要求点） 発熱量制約
・保管場所
ガス発生の観点での要求
二次廃棄物（性状・量） 取出し・物流ルート
材質・化学組成（保管の観点）
前処理・処理・再加工
減容性の観点での要求レベル区分の必要性の有無
漏えい・飛散の程度
望ましい保管期間
望ましい保管期間点）
・保管方法
漏えい・飛散の観点での要求 レベル区分の必要性の有無
線量率（敷地境界の観点）
材質・性状・化学組成（同上）
ガス発生の観点での要求
ガス発生許容量保管
一時保管（保管継続）
処理した廃棄物（性状・量）
物量 （至近３か年で早期）取出し・物流ルート
漏えい・飛散防止への要求
ガス発生許容量
減容性
減容性の観点での要求・線量率（敷地境界の観検認の観点での要求・処理目的
漏えい・飛散の観点での要求 望ましい分別程度
処分に有害と考えられる物質（至近３か年で早期） 形状・寸法（同上）
減容性の観点での要求 材質（長期安定性、腐食、水
二次廃棄物（性状・量）・保管場所
望ましい分別程度
望ましい分別程度
再利用用途
ガス発生の程度
核種組成・核種濃度点）
ガス発生の観点での要求
線量率（作業、敷地境界の観・処理技術
検認の観点での要求の影響有無検討
・保管場所
ガス発生の観点での要求 素）・保管方法
廃棄物管理を具体化するため 廃棄物管理を具体化するため廃棄物管理を具体化するため
材質・化学組成（保管の観点） 取出し・物流ルート
核種組成・核種濃度（処理の観漏えい・飛散の程度
望ましい保管期間
漏えい・飛散の観点での要求点）望ましい保管期間
形状・寸法（目視）・保管方法
漏えい・飛散の観点での要求・線量率（敷地境界の観の放射能濃度の放射能濃度
前処理・処理・再加工
減容性の観点での要求レベル区分の必要性の有無 の放射能濃度
線量率（敷地境界の観点）点）処理した廃棄物（性状・量）検認の観点での要求
レベル区分の必要性の有無
線量率（敷地境界の観点）
・線量率（敷地境界の観物量 （至近３か年で早期）検認の観点での要求点）
一時保管（保管継続）
材質・性状・化学組成（同上）
ガス発生の観点での要求
ガス発生許容量
二次廃棄物（性状・量） 取出し・物流ルート
減容性の観点での要求漏えい・飛散防止への要求
ガス発生許容量点）
形状・寸法（同上）
・保管場所望ましい分別程度
望ましい分別程度
望ましい分別程度
材質（長期安定性、腐食、水・処理目的
漏えい・飛散の観点での要求 望ましい分別程度
処分に有害と考えられる物質
ガス発生の観点での要求
再利用用途
核種組成・核種濃度（処理の観線量率（作業、敷地境界の観素）・保管方法
廃棄物管理を具体化するため 廃棄物管理を具体化するため廃棄物管理を具体化するため廃棄物管理を具体化するため減容性の観点での要求・処理技術検認の観点での要求の影響有無検討
漏えい・飛散の観点での要求
減容性の観点での要求レベル区分の必要性の有無
物量の最新情報点）点） 前処理・処理・再加工
形状・寸法（目視）の放射能濃度の放射能濃度の放射能濃度
レベル区分の必要性の有無 ・線量率（敷地境界の観 の放射能濃度
廃棄物（発生継続）
前処理・処理・再加工に必要廃棄体化に必要な性状調査
ガス発生の観点での要求検認の観点での要求
材質・性状・化学組成（同上）（至近３か年で早期）ガス発生の観点での要求
ガス発生許容量
一時保管に必要な性状調査
保管に必要な性状調査要求
線量率（敷地境界の観点）
点）
取出し・物流ルート
漏えい・飛散防止への要求
ガス発生許容量・廃棄物の物性把握
な性状調査要求
要求
漏えい・飛散の観点での要求
一時保管（保管継続）
物量
形状・寸法（同上）・処理目的
漏えい・飛散の観点での要求 処分に有害と考えられる物質
要求減容性の観点での要求
再利用用途
検認の観点での要求
・保管場所
望ましい分別程度
望ましい分別程度
望ましい分別程度
望ましい分別程度
材質（長期安定性、腐食、水
線量率（作業、敷地境界の観・処理技術
検認の観点での要求の影響有無検討
ガス発生の観点での要求
素）検認の観点からの調査要求点）
・保管方法
廃棄物管理を具体化するため 廃棄物管理を具体化するため 廃棄物管理を具体化するため廃棄物管理を具体化するため減容性の観点での要求
物量の最新情報
漏えい・飛散の観点での要求
廃棄物の発生
一時保管
前処理・処理・再加工
保管
廃棄体化
廃棄前保管
形状・寸法（目視）
廃棄物（発生継続）
前処理・処理・再加工に必要廃棄体化に必要な性状調査
ガス発生の観点での要求処分区分の観点からの調査要・線量率（敷地境界の観 の放射能濃度の放射能濃度の放射能濃度の放射能濃度
一時保管に必要な性状調査
保管に必要な性状調査要求
一時保管（保管継続） 線量率（敷地境界の観点）
検認の観点での要求
物量・廃棄物の物性把握
な性状調査要求
要求
漏えい・飛散の観点での要求 求（処分影響物質、安全評点）
減容性の観点での要求
要求・保管場所
望ましい分別程度
望ましい分別程度
望ましい分別程度
望ましい分別程度
材質（長期安定性、腐食、水検認の観点での要求
価用ソースターム設定、等）
ガス発生の観点での要求
各ステップへフィードバックする要求事項
素）・保管方法
廃棄物管理を具体化するため 廃棄物管理を具体化するため 廃棄物管理を具体化するため廃棄物管理を具体化するため
漏えい・飛散の観点での要求
減容性の観点での要求検認の観点からの調査要求
形状・寸法（目視）・線量率（敷地境界の観 の放射能濃度の放射能濃度の放射能濃度の放射能濃度
物量の最新情報
黒字：国プロ（2018年度までの）の研究計画外
廃棄物の発生
一時保管
前処理・処理・再加工
保管
廃棄体化
廃棄前保管
処分・再利用検認の観点での要求廃棄物（発生継続）
前処理・処理・再加工に必要廃棄体化に必要な性状調査 ガス発生の観点での要求処分区分の観点からの調査要
線量率（敷地境界の観点）点）
一時保管に必要な性状調査
保管に必要な性状調査要求
赤字：国プロ（2018年度までの）の研究計画にて検討予定（未着手）
至近３年で早期に決定する範囲
・廃棄物の物性把握
な性状調査要求
要求
漏えい・飛散の観点での要求 求（処分影響物質、安全評
要求
青字：着手済み、一部検討済み
各ステップへフィードバックする要求事項
検認の観点での要求
価用ソースターム設定、等）
減容性の観点での要求 緑字：検討済み
検認の観点からの調査要求
物量の最新情報廃棄物（発生継続）
前処理・処理・再加工に必要廃棄体化に必要な性状調査 ガス発生の観点での要求処分区分の観点からの調査要
黒字：国プロ（2018年度までの）の研究計画外
廃棄物の発生
一時保管
前処理・処理・再加工
保管
廃棄体化
廃棄前保管
処分・再利用一時保管に必要な性状調査
保管に必要な性状調査要求
・廃棄物の物性把握
な性状調査要求
要求
漏えい・飛散の観点での要求 求（処分影響物質、安全評至近３年で早期に決定する範囲
赤字：国プロ（2018年度までの）の研究計画にて検討予定（未着手）要求
価用ソースターム設定、等）
青字：着手済み、一部検討済み 検認の観点での要求
各ステップへフィードバックする要求事項
緑字：検討済み
廃棄物の発生
一時保管
前処理・処理・再加工
保管
廃棄体化
廃棄前保管
処分・再利用
黒字：国プロ（2018年度までの）の研究計画外
赤字：国プロ（2018年度までの）の研究計画にて検討予定（未着手）
至近３年で早期に決定する範囲
各ステップへフィードバックする要求事項
青字：着手済み、一部検討済み
緑字：検討済み
黒字：国プロ（2018年度までの）の研究計画外
赤字：国プロ（2018年度までの）の研究計画にて検討予定（未着手）
至近３年で早期に決定する範囲
青字：着手済み、一部検討済み
緑字：検討済み
各ステップへのインプットする情報※7

廃棄体化 廃棄確認

各ステップへのインプットする情報※9

第二仮保管施設

KH

1

処分
※8

処分

廃棄確認
*1
【線量区分値】政令濃度上限値，埋設規則*2 から設定
・線量(ＴＲＵ相当)*1；>10 （Bq/t）α濃度

11

吸着塔

空容器
建屋
設備 等○
一時保管施設（第一施設）
0○本
本
一時保管施設（第一、第四施設）
本 吸着塔
１～４号機PP内 高温焼却炉建屋
第二セシウム吸着装置 IE-96
一時保管施設（第四施設）
１～４号機PP内
焼却工作建屋
セシウム吸着装置
SMZ
本○
IE-911
H
本○
EN-101
AGH
HS-726
本○
EH

１～４号機PP内

廃棄確認

保管
※6
Ｂ

*2
60
Ｅ
*1
処分
廃棄確認
・線量(高Ｌ１相当)；>10 （Bq/t）
Co，>10 （Bq/t）137Cs【線量区分値】政令濃度上限値，埋設規則*2 から設定

KH

○

1 本
○
8.1E+14
Bq/カラム場所、フロア本○
8.1E+14 Bq/カラム
本○保管

・ＨＩＣ外側容器は切断し保管。必要に応じて除染。(※4)

一時保管施設（第一、第四施設）
空容器
吸着塔

廃棄確認

保管

1. Proposed waste stream that
integrates the processing flowthat can be presented currently
2. Issues in waste stream
formulation

Ｂ

※7
容器保管

Ｄ

※6
固型化保管
Ｂ

【共通事項】【共通事項】【共通事項】【共通事項】【共通事項】※6
・ＨＩＣ容器内で保管。
－
・必要な場合は廃棄体化処理用の容器 ・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲ
※7
Ｇ’
(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
容器保管
廃棄体化
保管
【個別事項】・スラリーの取出し。(※2)(※6)
管。(※7)ピット処分として扱う前提。(※8)
【共通事項】【共通事項】【共通事項】【共通事項】 廃棄確認
【共通事項】
・漏えいリスクの低減。(※3)
【個別事項】【個別事項】・ＨＩＣ外側容器は可能であ※6－ －・ＨＩＣ容器内で保管。
・必要な場合は廃棄体化処理用の容器 ・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲ
Ｇ’・ＨＩＣ外側容器は切断し保管。必要に－－れば再資源化。(※9)(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
応じて除染。(※4)
【個別事項】
考え方【個別事項】・スラリーの取出し。(※2)(※6)
として扱う前提。(※8)
廃棄確認 管。(※7)
ピット処分－【共通事項】【共通事項】【共通事項】【共通事項】－
・漏えいリスクの低減。(※3)
【個別事項】 【共通事項】【個別事項】・ＨＩＣ外側容器は可能であ
留意事項
・ＨＩＣ容器内で保管。
－
・必要な場合は廃棄体化処理用の容器
・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲれば再資源化。(※9)
・ＨＩＣ外側容器は切断し保管。必要に－－(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
応じて除染。(※4)
【個別事項】【個別事項】・スラリーの取出し。(※2)(※6)
管。(※7)
として扱う前提。(※8)－
【共通事項】【共通事項】【共通事項】【共通事項】【共通事項】
・漏えいリスクの低減。(※3)
【個別事項】 ・スラリーは処分に適合するよ【個別事項】・ＨＩＣ外側容器は可能であ・ＨＩＣ容器内で保管。 － －・必要な場合は廃棄体化処理用の容器
・α 核種を含む廃棄物はＴＲ
・ＨＩＣ外側容器は切断し保管。必要に－－れば再資源化。(※9)(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）【個別事項】・スラリーの取出し。(※2)応じて除染。(※4)(※6)
管。(※7)
として扱う前提。(※8) 【個別事項】－－
・漏えいリスクの低減。(※3)
【個別事項】【個別事項】・ＨＩＣ外側容器は可能であ

AGH
EH

容器保管
※7

（PE 容器等）

有害物等（PE 容器等）

仮保管施設
場所、フロア
1 等
設備

本
○
公開可否
１～４号機PP内 放射能濃度
焼却工作建屋

1 本 ○：可 ×：否
-

含水率

保管場所

発生場所

号機、

3.3E+14
Bq/カラム
本
○建屋
構内エリア

1 本

可燃物

（PE 容器等）

可燃物

保管

Cs137
36 本
吸着塔

本

化学組成
線量

○：可 ×：否
○
α核種検出
参照文献
○ 廃棄物リスト(List of Wastes)

33吸着塔
本
2 本

将来見込

分析
本
最大放射能濃度
本
濃度(Bq/g)
本

36 本
発生時期
処理水から評価
Cs137
1 本 外形
発生理由
復旧対策

汚染水処理

本

放射能濃度

公開可否

将来見込
分析

8 本 最大放射能濃度
-

Ｄ

減容 *2

※4

※4

保管

（SUS 補強体等）

発生量
2014年12月末

漏洩防止拡大保管
廃棄体化
処分・再利用

廃棄前保管

漏洩防止拡大保管
水分
廃棄体化
低減

保管

(Processing flow withplenty of options)（SUS 補強体等）

金属

サンプル数

※2

Ｄ
容器保管
固型化
※6
Ｂ 解体・分別処分・再利用※7

廃棄体化
処分・再利用※7；<10 （Bq/t） Co，<10 （Bq/t） Cs
・線量(ＴＲＵ相当)*1；>10 ・線量(高)（Bq/t）α濃度
※8
10
8
*2
60
15
60 ；<10 14Ｅ
Co，<10（Bq/t）137Cs
・線量(高Ｌ１相当)*2；>10 ・線量(中)（Bq/t）
Co，>10 （Bq/t）（Bq/t）137
Cs
処分*2【線量区分値】政令濃度上限値*1，埋設規則
から設定
固型化
※6
15
14
・線量(低)；検討が必要*2
Ｆ・線量(高)；<10（Bq/t）60Co，<10 （Bq/t）137Cs
11
解体・分別
※8 8
・線量(ＴＲＵ相当)*1；>10 （Bq/t）α濃度10*260・線量(中)；<10 （Bq/t）
Co，<10 （Bq/t）137Cs
除染
再加工
15
14
*2
Ｅ
・線量(高Ｌ１相当)；>10（Bq/t）60Co，>10 （Bq/t）137Cs*1【線量区分値】政令濃度上限値，埋設規則*2から設定
・線量(低)；検討が必要
Ｆ 15
14
*2
※411
※7
；<10 （Bq/t）60Co，<10 （Bq/t）137Cs*1 ・線量(高)・線量(ＴＲＵ相当)；>10 （Bq/t）α濃度
保管
容器保管
廃棄体化
減容
保管
10
8
除染
*2
60
再加工・線量(中)再利用15
14；<10 （Bq/t）
Co，<10 （Bq/t）137Cs
・線量(高Ｌ１相当)*2；>10 （Bq/t）60Co，>10 （Bq/t）137Cs※61514・線量(低)；検討が必要※9137
Ｇ※4・線量(高)*2Ｆ※7；<10 （Bq/t）60Co，<10 （Bq/t）
Cs

保管

Ｄ

基本フロー

HIC 残材

漏洩防止拡大保管
容器保管

保管

HIC 保管
Ｃ（堰設置等）

廃棄体化
Ａ
取出し

Ｃ
HIC 保管

漏洩防止拡大保管
水分
廃棄体化
低減

ＣHIC 保管
※2

※3
取出し

取出し

※3

Ａ
※3
取出し

漏洩拡大防止

基本フロー
※1

※1

廃棄前保管

HIC 保管
廃棄前保管

廃棄体化
Ａ

※3

漏洩拡大防止（堰設置等）

HIC 保管（スラリー）

多核種 HIC（スラリー）

保管※1
保管
漏洩拡大防止（安定化処理）多核種 HIC

汚染区分：XI-4

HIC 保管Output※1※1
保管

漏洩拡大防止漏洩拡大防止廃棄体化
Ａ
取出し(Refined processing flow)
金属（SUS 補強体等）

金属

保管（SUS 補強体等）
可燃物（PE 容器等）
保管

可燃物（PE 容器等）
考え方(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
応じて除染。(※4)
【個別事項】【個別事項】・スラリーの取出し。(※2)(※6)
として扱う前提。(※8)
廃棄確認 管。(※7)
ピット処分－【共通事項】【共通事項】【共通事項】【共通事項】－
・漏えいリスクの低減。(※3)
【個別事項】 【共通事項】【個別事項】・ＨＩＣ外側容器は可能であ・ＨＩＣ容器内で保管。
－
・必要な場合は廃棄体化処理用の容器
・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲれば再資源化。(※9)
・ＨＩＣ外側容器は切断し保管。必要に－－(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
応じて除染。(※4)
【個別事項】【個別事項】・スラリーの取出し。(※2)(※6)
管。(※7)
として扱う前提。(※8)－
【共通事項】【共通事項】【共通事項】【共通事項】【共通事項】－
・漏えいリスクの低減。(※3)
【個別事項】【個別事項】・ＨＩＣ外側容器は可能であ・ＨＩＣ容器内で保管。
－
・必要な場合は廃棄体化処理用の容器 ・スラリーは処分に適合するよ・α 核種を含む廃棄物はＴＲ
・ＨＩＣ外側容器は切断し保管。必要に－－れば再資源化。(※9)(※1)
【個別事項】に収納。
う固型化またはそのまま保Ｕ廃棄物（地層処分対象）
応じて除染。(※4)
【個別事項】【個別事項】・スラリーの取出し。(※2)(※6)
管。(※7)
として扱う前提。(※8)－－
・漏えいリスクの低減。(※3)
【個別事項】【個別事項】・ＨＩＣ外側容器は可能であ・ＨＩＣ外側容器は切断し保管。必要に－－れば再資源化。(※9)
応じて除染。(※4)
【個別事項】
－

留意事項
考え方
留意事項
考え方
留意事項

廃棄物の発生（堰設置等）※3

Detailing waste classification, and refining the processing flow

直
近
５
年
以
内
に
方
針
を
決
め
る
べ
き
範
囲

一時保管

前処理・処理
・再加工

保管

廃棄体化

廃棄前
保管

処分
・
再利用

廃棄前
保管

処分
・
再利用

水処理二次廃棄物
吸着塔（ＫＵＲＩＯＮ、ＳＡＲＲＹ、モバイル浄化装置）、
多核種除去装置（吸着材、吸着塔）、フィルタ（高性能多吸着塔、ＨＩＣ保管施設（ストリームＳ９、Ｓ１０、Ｓ１２）＋

大型廃棄物保管庫

多核種除去設備（スラリー）（ストリームＳ１０）（ストリームＳ９、Ｓ１０、Ｓ１２）

水処理２次廃棄物の処理・保管・廃棄体化のための設備、施設（ストリームＳ９、Ｓ１０、Ｓ１１、Ｓ１２、Ｓ１４）

除染スラッジ

Ｐ／ＭＢ Ｄピット（ストリームＳ１１）

濃縮廃液（蒸発濃縮廃液）

一次保管（ストリームＳ１３）

除染スラッジ（ストリームＳ１１）

２次廃棄物等増設固体廃棄物貯蔵庫（ストリームＳ６）

固
体
廃
棄
物
の
保
管
管
理
計
画
等
で
方
針
が
決
ま
っ
て
い
る
範
囲

滞
留
水
処
理
等
復
旧
対
策

一時保管
（ストリームＳ６）
テント、覆土、屋外集積、屋外
容器収納、固体庫

がれき類減容処理設備増設固体廃棄物貯蔵庫（ストリームＳ６）（ストリームＳ６）

Waste stream(Dash board)

可燃物一時保管

可燃性雑固体（ストリームＳ８）

伐採木

一時保管

仮設処理装置・タンク等

焼却灰保管（ストリームＳ８）（ストリームＳ８）

金属、コンクリート等の廃棄体化のための設備、施設

焼却炉前処理設備
増設雑固体焼却設備

一時保管（ストリームＳ８）

汚染土壌等

雑固体焼却設備（全ストリーム）（全ストリーム）

汚染土一時保管施設

除染設備

一時保管

２次廃棄物等２次廃棄物等

時
間
を
か
け
て
方
針
を
決
め
て
い
く
範
囲

原
子
炉
・
建
屋
の
解
体
廃
棄
物

解体に伴って発生する可能性のある廃棄物

金属廃棄物、コンクリート等の処理・保管

設備、施設（ストリームＳ１、Ｓ２、Ｓ４、Ｓ１４）

２次廃棄物等２次廃棄物等２次廃棄物等

ストリームＳ１４

デブリ取出し前の発生廃棄物

次の処理フローと同じ扱いの組み合わせ

Overall integration (Bird's-eyeview update)
圧力容器（ストリームＳ１）、格納容器金属（Ｓ２）、格納容器コンクリート（Ｓ３）、建屋内金属（Ｓ４）、建屋内コンクリート（Ｓ５）、多核種除去装置（Ｓ１０）、フィルタ（Ｓ１２）

燃料デブリ・燃料デブリ付着
物 保管施設

燃料デブリ付着物

保
管
廃
棄
物

金属

（ＳＦＰ、サイトバンカ
保管物等）

保管

保管廃棄物については
今後検討が必要

本図は、検討の進捗を反映して見直していく

：運用中

：計画中

樹脂

©International Research Institute for Nuclear Decommissioning250

d. Integration of R&D- Overview of 1F waste treatment summarized at present in this subsidy project –【Confidential】(Waste stream dashboard)
We are trying to integrate the whole picture and accumulate the R&D results, but it is necessary to further refine the processing flow
despite lack of information, in order to have an insight into the feasibility of treatment and disposal.

Pre-treatment,
treatment,
reprocessing

Storage

Preliminarystorage ofwaste

Waste conditioning

Disposal/
re-useThe blue frame
represents the area
where deeper study
is required

Secondary wastes generated from contaminatedwater treatment

Restoration measures for stagnant water treatment

Concentrated liquid waste
slurry (Evaporative
concentrated liquid waste)
Adsorption tower (KURION,SARRY, mobile purification device),multi-nuclide removal system(slurry, adsorbents, adsorption
tower), filter (high performance
multi-nuclides)

DecontaminatedsludgeDebris types

Miscellaneous,
combustible solid
wastes (Protective
clothing, combustible
substances)

Primary storage(Stream S13)
Adsorption tower, HIC
storage facility (Stream
S9, S10, S12)

Large store room for
wastes (Stream S9,
S10, S12)

Equipment and facilities for the treatment, storage, andconditioning of secondary wastes generated from contaminatedwater treatment (Stream S9, S10, S11, S12, S13, S14)

Treatment and storage of slurry from the multinuclide removal system (S10)P/MB D pit(Stream S11)

Temporary storage(Stream S6, S7)
Tents, soil cover,open storage,
outdoor container
storage, solid waste
facility

Treatment and storage of decontaminated sludge(S11)
Extended storage
facility for solid waste(Stream S6, S7)

Temporarystorage (Stream
S8)

Contaminated soil etc

Temporarystorage

Volume reduction
treatment facility(Stream S6, S7)

Miscellaneous solid
waste incineration
facility (Stream S8)

Temporary storage of
combustible materials(Stream S8)Felled trees

Temporary water
treatment equipment,tank etc

Temporary storage
facility forcontaminated soil

Temporarystorage

Extended storage facility
for solid waste (Stream S6,
S7)
Incinerated ash
storage (Stream S8)

Waste
preliminarystorage (Allstreams)

Equipment, facilities
for waste conditioning
such as of metals andconcrete (All streams)

Treatment prior toincineration, Miscellaneous
solid waste incineration
facility

Decontaminationequipment

Dismantling waste fromreactor and buildings

Secondary waste

Storagewaste

Scope of time-consuming policy decisions

Temporary storage

Disposal andre-useScope of storage management plan for solid wastes for which
the policy has been decided

Scope of the study on policies

Generation of waste

Secondary waste

Waste likely to begenerated due todismantling

Treatment, storage, equipment, facility for metallic waste, concretes (Stream
S1, S2, S10, S14)

Secondary waste
Secondary waste

Stream S14, Combination of handling same as next processing flow
Pressure vessel (Stream S1), containment vessel metal (S2), containment vessel
concrete (S3), metals inside buildings (S4), concrete inside buildings (S5), multi-nuclide
removal system (S10), filter (S12)

Waste generated
before debris removal

Metals (SFP, site
bunker stored matter,etc.)
ResinsSecondary waste

Storage

Further study on stored
waste is required

This diagram will be revised to reflect the
progress in the study

Fuel debris deposits

Fuel debris and fuel debris
deposits, storage facility

Operation
inprogress
Planning
inprogress

Fig. 1 Waste stream dashboard (Created in FY2018)©International Research Institute for Nuclear Decommissioning251

d. Integration of R&D【Confidential】- Example of study on refining the treatment flow and current issues(Example of HIC slurry) ① Multiple options for pre-treatment,
treatment and re-processing
Generation ofwaste
廃棄物の発生多核種除去装置
Multi-nuclide
removal
HICsystem
HIC (Carbonate
slurry, co（炭酸塩スラリー、
precipitated slurry)

Temporary
一時保管
storage

HIC storage
HIC 保管

Pre-treatment, treatment, re前処理・処理・再加工
processing(Stabilization
process)（安定化処理）HIC 保管
HIC
storage

Prevention of spread
漏洩拡大防止of leakage

Storage to prevent
漏洩防止拡大保管
spread of leakage

Segregation

分別

A: It is necessary to provide feedbackon the study results of waste types(characteristics of wastes etc.) in orderto refine the options for pretreatment,
treatment and re-processing.

HIC 残材保管

金属

ストリームＳ６（SUS 補強体等）
Iron co-precipitation slurry

Storage into aseparate
別容器収納container

Decrease in水分moisture
低減
content

固型化
Solidification

除染Dismantling/
解体・分別
segregation

Carbonate slurry

Sample image of slurry from the multi-nuclide removal system (Pictures represent simulants)
Combustible
可燃物
material (PEcontaineretc.)（PE 容器等）

Storage
保管

Continuation of
現状保管継続existing storage

鉄共沈スラリー）HIC Image(Before SUS
reinforcement)
基本フロー
Basic flow

② Multiple options for the methods
related to waste conditioning

ストリームＳ８
Stream S8

容器保管 storage
Container

B: It isnecessary toprovidefeedback on the
study results ofideal form ofstorage(anticipatedrequirements) inorder to refine
the options for
pretreatment,
treatment, reprocessing.
C: It isnecessary toprovidefeedback on the
study results ofideal form ofstorage(anticipatedrequirements) inorder to refine
the options for
waste
conditioning

廃棄体化
Waste
conditioning

Preliminary storage
廃棄前保管of waste

Container
容器保管storage

Disposal/
re-usage
処分・再利用Confirmation
廃棄確認of waste

充填
Filling

Segregation
分別

Filling
充填
Cementkneading
and
ｾﾒﾝﾄ混錬固化
solidification

Container
容器保管storage

Disposal
処分

Melting and
溶融固化
solidification
Glass
ｶﾞﾗｽ固化hardening

砂充填
Sandfilling

D: Feedback from thestudy on disposal isrequired in order torefine the options for
waste conditioning.
(Minimum requirements
for waste, such asmechanical strengthand multi-nuclide
confinement
performance)

Fig. 1 Example of refining the treatment flow andcurrent issues (Example of HIC slurry)©International Research Institute for Nuclear Decommissioning252

【Confidential】

d. Integration of R&D- Study policies for current issues 1.

It is necessary to have consistency in waste classification, which will be usedduring the study of each stage from characterization to disposal.

<Study Policy>
a. To re-organize the classification in order to link it to the radioactivity
evaluations in the distant future.
b. To study integrated management by consolidating the inventory information
etc. into a waste list.
c. To revise the already created treatment flows into flows considering thenuclides or the physical amounts etc.
2.

It is necessary to determine the specific details of the ideal form of storage.

<Study policy>
a. To study the ideal form of storage and examine the pre-treatment, treatment
and re-processing required for it.
3.

It is necessary to indicate the minimum requirements of a waste form.

<Study policy>a. First of all, disposal concepts are considered to confirm the feasibility, in casethat a waste form are not required the best confinement performance.

©International Research Institute for Nuclear Decommissioning253

【Confidential】

d. Integration of R&D- Study policies for current issues –1. Waste classification

In order to link the solid waste at 1F to the radioactivity evaluations in the distant future, it wasconsidered appropriate to broadly classify waste under the following four types. The definitions(proposed) and characteristics of each type of waste were studied. These classifications were also
defined in detail.
(1) Dismantling waste, (2) Types of debris, (3) Secondary wastes generated from contaminatedin the atmospherewater
treatment,(4)Inside
Other(Including the waste during Released
operations)
Fuelfrom withinthethe building
and oceanCore internals
core

Soil

Released in the air inside thebuilding

Debris (outside the building)

Fuel debris

Felled trees
Waste from pressure vessel [before the accident: equivalent to
L1](Contamination due to adherence of fuel debris + stagnant
water)

Stagnant
waterContaminated water treatment

Waste from pressure vessel [before the accident: equivalent to
L1](Contamination due to stagnant water)

Waste from pressure vessel [before the accident: equivalent to
L2](Contamination due to adherence of fuel debris + stagnant
water)
Waste from pressure vessel [before the accident: equivalent to
L2](Contamination due to stagnant water)

Waste from containment vessel (concrete) [before the accident:
equivalent to L2](Contamination due to adherence of fuel debris + stagnant water)

Waste from containment vessel (metal) [before the accident: equivalent to
L2](Contamination due to adherence of fuel debris + stagnant water)
Waste from containment vessel (metal) [before the accident: equivalent to
L2](Contamination due to adherence of fuel debris + stagnant water)

Waste from containment vessel (concrete) [before the accident:
equivalent to L2](Contamination due to adherence of fuel debris + stagnant water)

Waste from containment vessel (metal) [before the accident:
equivalent to L3](Contamination due to stagnant water)

Waste from containment vessel (concrete) [before the accident:
equivalent to L3](Contamination due to stagnant water)
Waste from containment vessel (concrete) [before the accident:
equivalent to L3](Contamination due to stagnant water)

Waste from containment vessel (metal) [before the accident:
equivalent to L3](Contamination due to stagnant water)

Adsorption tower

Waste inside the building (metal) [before theaccident: equivalent to L3](Contamination due to stagnant water)

Decontaminationequipment
Multi-nuclide removal facility

Dismantling
waste

Waste inside the building (metal) [before
the accident: equivalent to L3]

Waste inside the building (metal) [before the accident: nonradioactive](Contamination due to stagnant water)

Waste inside the building (metal) [before
the accident: non-radioactive]

Waste inside the building (concrete) [before theaccident: equivalent to L3](Contamination due to stagnant water)

Waste inside the building (concrete)

Waste inside the building (concrete) [before the accident: nonradioactive](Contamination due to stagnant water)

Waste inside the building (concrete) [before#[before the accident: equivalent to L3]
the accident: non-radioactive]
Fig.1 Classificationof waste based on contamination

Treated water# R&D concerning the treatment and disposal of solid waste, Interim report (April 5,2018), cited from Classification of waste based on contamination.
©International Research Institute for Nuclear Decommissioning254

【Confidential】

The study will continue in FY2020

d. Integration of R&D- Study policies for current issues 1. Waste classification(1) Dismantling waste
Definitions(proposed)

The waste which is managed by storage in container units depending on various waste
properties and when the waste is generated, in accordance with the decommissioning plan.

Characteristics

The waste present at 1F even before the accident and some of which is radioactively
contaminated, etc.

(2) Debris types
Definitions(proposed)

Waste that consists mainly of metal and concrete, but organic matter (gum tape and curing
generated during work) is also mixed in it, and the contents are not managed.

Characteristics

The waste present at 1F even before the accident, and variation in the properties of theinventory and materials is large.

(3) Secondary wastes generated from water treatment
Definitions(proposed)

Waste such as used adsorbents generated from contaminated water treatment facility. Notethat it does not include the dismantling waste of the contaminated water treatment facility itself.

Characteristics

Properties such as inventory and water content differ depending on when the waste is
generated and the water treatment facility at the source of waste generation, but the waste issegregated and managed at the time of generation for each type of adsorbent.

(4) Others
Definitions(proposed)

The waste that is not included in the above-mentioned dismantling waste, types of debris or thesecondary wastes generated from contaminated water treatment.

Characteristics

Waste that differs significantly in terms of material or inventory, physical amounts, properties
etc. and which is difficult to define uniformly.
©International Research Institute for Nuclear Decommissioning255

d. Integration of R&D- Study policies for current issues 1. Waste classification【Confidential】

The detailing of waste classification was carried out jointly with the operators.
For conformity with the input compilation table, a waste management sheet was created foreach type of waste and the mechanism to improve the searchability of information was studied.
In FY2019, an example was created for the consolidation of information into the waste
management sheet for the main secondary wastes generated from contaminated water
treatment.

Fig. 1 Input compilation table

The Waste Management Sheet for each type ofwaste is maintained so that the details or progressof the R&D results for each item in the input
compilation table can be referenced.

Fig. 2 Waste Management Sheet©International Research Institute for Nuclear Decommissioning256

【Confidential】

d. Integration of R&D- Study policies for current issues 2. Ideal form of storage
From “Basic approach to nuclear safety" (Atomic Energy Society of Japan (AESJ))
Purpose of Nuclear Safety
The purpose is to protect humans and the environment from the harmful impact of radiation
resulting from nuclear power facilities and their activities.
Basic Principles of Nuclear Safety

Protection of humans and the environment fromthe risk of radiation

Nuclear
SafetyCategory

1

2

3

Responsibility
andManagement

Protection ofhumans andtheenvironment

Confinementof radiation
risk source

Principles
1

Obligation to safety

2Role of Government

3

Role of regulatory organizations

4

Leadership and management for safety5

Development of safety culture

6

Explanation on the validity of the nuclear power facilities
and their activities

7

Risk control for humans and environment and continuous
efforts for the same

8

Prevention of accidents and impact mitigation

9

Preparation and response during emergencies

Principles related to disaster prevention
measures such as evacuation etc.
(which will be studied in future)

10

Protection measures for reducing the existing radiation
risk or the uncontrolled radiation risk

Principles related to natural radiation
sources (which will be studied in future)

Principles related to organizations
and systems (which will be studied
in the future)

It is necessary to decrease the amountof the site boundary radiation dose andprevent any leakage or diffusion duringstorage based on these viewpoints©International Research Institute for Nuclear Decommissioning257

【Confidential】

d. Integration of R&D- Study policies for current issues 2. Ideal form of storage
Ideal form of storage considering nuclear safety (Proposed)◼ Ensure the shielding of radiation.
◼ Ensure the confinement of radioactive materials. (Maintain in a state where there is no leakage or diffusion,and leakage or diffusion does not take place easily)◼ When the gases are released in a controlled manner, ensure that the radioactive materials havebeen sufficiently diluted by the ventilation and air-conditioning system of the storage facility.
(When hydrogen is discharged through the container vent etc.)

Using the above principles, the form of storage for each type of waste will be studiedin consideration of the future waste conditioning treatment and ensuring safety duringstorage, etc.
For instance, following are the assumptions for the secondary wastes generated from contaminated water
treatment such as slurry etc. illustrated by a specific example in FY2019.
The ideal form of storing the other types of waste will also be studied in FY2020.
⚫ The study assumes that the radiation is shielded at the storage facility.
⚫ The study assumes that the radioactive materials are confined inside containers and the storage
facilities.
The amount of free water is reduced as much as possible so as to lower the uncertainty (risk) of
any further leakage or diffusion. The study assumes that solidification treatment is not performed
considering the effect on future waste conditioning treatment.
⚫ It is assumed that the gas generated from waste is released in a controlled manner. The study
assumes that it is ensured that the radioactive materials have been sufficiently diluted by theventilation and air-conditioning system of the storage facility.

The study will also be continued from the viewpoint of functional requirements with respect tothe ideal form of storage.
©International Research Institute for Nuclear Decommissioning258

d. Integration of R&D- Study policies for current issues 3. Minimum requirements for waste packages【Confidential】

Identification of factors that have a large impact on refining the wasteconditioning treatment for the waste stream(1) "Necessity of Solidification" has a particularly large impact on the study ofwaste streams.
(2) "Necessity of Mineralization and Detoxification" has a large impact on
refining of the waste stream

Study of the waste stream when the best waste package performance is not required.
① In order to obtain a technical outlook, a case of waste conditioning treatment wherein the
introduction of equipment is as limited as possible, and which is feasible up to disposal, will beassumed as Case 1 (Base case) in the waste stream.
② First of all, in order to minimize the waste conditioning process (equipment introduction), a waste
stream will be developed wherein waste package performance such as confinement of waste
during waste disposal is not required, and the feasibility of this waste stream will be studied.
③ In Case 1, volume reduction is not considered while identifying issues. (Equipment that has been
introduced or is planned to be introduced is excluded)④ Other cases will also be studied based on Case 1.

©International Research Institute for Nuclear Decommissioning259

d. Integration of R&D【Confidential】- Study policies for current issues 3. Minimum requirements for waste packages
Based on the set pre-conditions and through the study of treatment and disposal processes anddiscussions with the concerned parties, it was decided that the waste packages must have harmful gap
countermeasures at the very least.
➢ It was assumed that the other waste package functions (confinement, strength etc.) will be ensured by meansof the containers or at the disposal sites.
This material presents Case 1 (Case where performance is not required during waste conditioning (only the harmful gap countermeasures are executed)) for the waste stream and shows the flow that is assumed fromstorage to disposal. It does not target any specific waste. More detailed processes and designing of individual devices must be studied separately for each type of waste in the future.
Is it necessary to confine gases
during transportation within the
1F premises?

Disposal facility

This material presents Case 1 (Case where performance is not required during waste conditioning (only the harmful gap countermeasures are executed)) for the waste stream and shows the flow that is assumed from storage todisposal. It does not target any specific waste. More detailed processes and designing of individual devices must be studied separately for each type of waste in the future.
What are the important safety requirements for
Is it necessary to solidify thethe waste packages? In order to confirm
Waste conditioning treatment facility (No segregation
The assumed flow from the generation of waste
waste prior to waste conditioning
Storage facilities
whether the safety requirements have been
only filling and solidification)
till storage will also be necessary in future
for the fall countermeasures?
fulfilled, which data is collected from where andMoreover, is confinement ofwhat is ensured?
[Transport Container Handling Area *1/ Truck[Truck yard/Transport Container
gases
necessary?
[Storage Area]
yard]
Handling Area *1][Receiving Inspection Area] *1[Outdoor (Within the 1F premises)]
Transport device
Transport container (Prevents the
Measurementofscattering or leakage ofradioactive
solidifications or liquids andVentilation andconcentration, and otherensures the confinement of gases
Airmeasurements
Waste(hydrogen,noble
gases)
andtheconditioning
Storageshielding function during transport)
SystemContainer[Truck yard/
Transport Container
Handling Area *1][Acceptance test/ Waste preliminary storage area] *1

Ventilation andAir-conditioning
SystemDamage verification /Tag verification / Other
waste verification[Waste burial area]

During geological disposal, aren’t theharmful gap countermeasuresunnecessary?
To begin with, why are only the gap
countermeasures particularly necessary?

Transport
Equipment*1 Unmanned remote operations depending on the surface dose rate of the storage container and the properties of the waste.
*4 It is assumed that there is technology that can perform all types of measurements appropriately irrespective of the shape of the*2 The waste storage container dispatched from the storage facility is assumed to be sound from the viewpoint of strength andcontainers or the state of the contents. The same applies below. Here, the operation guidelines for waste verification by the Nuclear
confinement. Note that the on-vehicle container lifted with the crane in the above diagram is shown as an example, but if theRegulation Authority assumes that the measurement records submitted by the nuclear power stations pertaining to the maximum
hanging handle of the waste storage container has deteriorated and cannot be used, then it is assumed that the container isradioactive concentrations, the surface density limits and the surface dose rate are measured at the waste conditioning treatment
picked up into the vehicle by means of a forklift. It is assumed that in case of other equipment also, studies will be conduc ted forfacility.
each type of waste considering the individual conditions of each equipment. The same applies below.
Since there is no segregation or mixing of the waste, it was assumed that the verification that the maximum radioactive conce ntration*3 In the above diagram, details pertaining to the transport container such as the opening/closing of the lid, temporary plac ement,
has not been exceeded, is carried out prior to the solidification process. The surface density limits and the surface dose ra te for thesurface contamination check prior to dispatch from the facility have been omitted. Moreover, in the above diagram, the transport
waste verification are measured after waste conditioning. All other essential tests are assumed to be executed together with the above.
container handling area and truck yard have been shown in the same room, but in reality, the two work activities should be ca rriedout in two
Further deliberation on waste conditioning
different rooms of the building as per the controlled area classification and this description has been omitted.
processisneededfor each type of waste,
The same applies below.

including sorting.

[Waste Conditioning Process Area (Filling and Solidification)] *1
To begin with, are the gap(★Diagram iscountermeasures necessary? Filling materiala typical
silo
example)Kneading
Lid openingwater
Lid closing
Kneading
devicedevicemachineMeasuringcleaning water
deviceKneading
machine[Dispatch
Investigation Area][Transport Container Handling Area *1/ Truckyard][Outdoors(Inside/Outside the *6 Under this assumption, it is presumed that the control of leaching of nuclides from waste packages is ensured at the disposal site. Confinementof nuclides in the waste packages is not expected.
1F premises)]*7 Based on the study results for the FY2016 Supplementary Budget “R&D for the treatment and disposal of solid waste”, study of storage
measures for high-level radioactive waste and the study of countermeasures for generation of hydrogen, it is assumed here that the hydrogen
To the waste pre- generated from the waste package is discharged outside the container via the container vent. Until the period of backfilling at the disposal site,storage/ disposal scavenging is carried out inside the disposal site as a hydrogen countermeasure.
*8 It is assumed that a cement-based filling material is injected between the wastes. The harmful gaps between the wastes are removed by thefacility
waste conditioning treatment.
*9 It is assumed that there is technology which can perform all types of measurements appropriately irrespective of the shape of the containers or
the state of the contents. The same applies below. Here, the operation guidelines for waste verification by the Nuclear Regulation Authority
assume that the measurement records submitted by the nuclear power stations pertaining to the maximum radioactive concentrati ons, the surface
density limits and the surface dose rate are measured at the waste conditioning treatment facility.
Since there is no segregation or mixing of the waste, it was assumed that the verification that the maximum radioactive concentration has not been
exceeded, is carried out prior to the solidification process. The surface density limits and the surface dose rate for the waste verification are*7 Every time a waste storage container is converted into a waste package container, it ismeasured after waste conditioning. All other essential tests are assumed to be executed together with the above.
*5 The filling and solidification process has been illustrated with a diagram, referring to the FY2014 Research Report on
assumed that even when the bolts or packing has deteriorated, the lid can be opened andDevelopment of technology for the treatment and disposal of accident waste, Appendix 4.3.2-1 Waste treatment and waste
closedbyappropriate
treatmentsuchasreplacement
of
the
bolt
or
packing.
Or
itisassumedconditioning technology catalog No.12 Mortar filling technology.
*6 The waste package performance is ensured in the process (in this diagram, the filling material, kneading water mixing cond itions that the filling material can be injected without opening the lid.
or the filling amount), and at the time of disposal, by verification of the operation records. This diagram assumes a wasteconditioning process which can process without removing the moisture or any disposal impacting substances contained in thewaste to be treated. (In case, it is necessary to remove the moisture or any solidification inhibiting or accelerating substances prior
to solidification, then besides the contents stated in this diagram, drying and segregation, as well as verification and vali dation of
the drying and segregation, are necessary)
Surface contamination
density/ Surface dose
rate measurement

Conversion of storage containers to waste package containers, and process verificationand validation for waste package performance were set as the preconditions. The
verification of these preconditions will be needed in future in order to verify feasibility.

©International Research Institute for Nuclear Decommissioning260

【Confidential】

d. Integration of R&D- Proposal for refining the treatment flow The refining of the treatment flow was carried out for the main secondary wastes
generated from contaminated water treatment based on the study results for ideal formof storage and for the minimum functions required for waste packages.

Generation of waste

Temporary storage

Pre-treatment, treatment, reprocessing

Storage

Waste conditioning

Preliminarystorage of
waste

Disposal and reuseContainer storage

Confirmationof waste

Contamination
classification

The classification was
detailed and summarized in
the Waste Management
Sheet.

Multi-nuclide
HIC (slurry)

HIC storage

Continuation ofcurrent storage

HIC storage

Prevention of escalationof leakage

Storage for preventing
escalation of leakage

Segregation

Storage inseparate
container

Containerstorage

Injection offilling
material
SegregationStorage

Decontamination

Disposal

Meltsolidification

Solidific
ation
Material remaining
in HIC

Container storage

Cement meltsolidification

Reductionof watercontent

Basic flow

Injection offilling material

VitrificationSandfilling

Dismantling andsegregation[Legend] Existing options / options for which equipment is being
Metal (SIS
reinforcement
material, etc.)

Combustibles(PE containers,
etc.)

Stream S6planned by the operator
Options narrowed down based on hypotheticalconditions
Other options

Stream S8

From the TEPCO HD plan and the study of the ideal formof storage, it was assumed that the moisture in thesecondary wastes generated from contaminated water
treatment will be reduced.
It was assumed that the waste will not be solidified.

From the study of the minimum requirements forthe waste packages, it was assumed that fillingmaterial will be injected as the minimum
response. In this case, it was assumed that astorage container will be converted into a waste
package container.

Geological disposal (marginal depth ordeeper) was assumed from the results ofthe study on the disposal concept.
Moreover, the waste conditioning
treatment facility is assumed to carry out
the process of verifying and validating
waste packages.

©International Research Institute for Nuclear Decommissioning261

【Confidential】

d. Integration of R&D- Example of reflection into the waste stream (Dashboard) PreStorage
保管

reprocessing

Restoration measures such as treatment of stagnant water, etc.

Secondary wastes
generatedfrom water
水処理二次廃棄物
treatment

滞
留
水
処
理
等
復
旧
対
策

Concentrated liquid waste
濃縮廃液スラリー
slurry (Evaporative（蒸発濃縮廃液）
concentrated liquid waste)
Adsorption tower (KURION,
吸着塔（ＫＵＲＩＯＮ、ＳSARRY, mobile purification
ＡＲＲＹ、モバイル浄化装device), multi-nuclide置）、多核種除去装置（スremoval system (slurry,
adsorbents, adsorption
ラリー、吸着材、吸着塔）、
tower), filter (highフィルタ（高性能多核種）
performance multi-nuclides)

一次保管storage
Primary（ストリームＳ１３）(Stream S13)
Large waste
Adsorption tower,
storagevaultsHIC
storage facility ＋
吸着塔、ＨＩＣ保管施設
大型廃棄物保管庫（ストリームＳ９、Ｓ１０、Ｓ１２）（ストリームＳ９、Ｓ１０、Ｓ１２）(Stream S9, S10,(Stream S9, S10,
S12)
S12)

P/MB D pit

Decontaminated
除染スラッジ
sludgeＰ／ＭＢ Ｄピット(Stream S11)
（ストリームＳ１１）

Additional solid
Temporary storage(Stream S6, S7)
一時保管
Tent,
cover soil,（ストリームＳ６、Ｓ７）
outdoor accumulation,
テント、覆土、屋外集積、
outdoor container
屋外容器収納、固体庫
storage, solid waste
vault

がれき類
Types
of debris

Flammable
miscellaneous
solid waste
可燃性雑固体(protective（保護衣、可燃物）
clothing,
combustibles)

waste storage
増設固体廃棄物貯蔵庫
vault (Stream S6,
S7)

Dismantling waste fromreactor core and building
Stored waste
保
管
廃
棄
物

Secondarywaste, etc.

２次廃棄物等処分 reDisposal/
・usePreliminary
HIC（酸化チタ
storage in HICン）廃棄前保(Titanium管
oxide)

ピット
Pit
disposal
処分

Preliminary
吸着塔（多核storage inadsorption
tower
種除去装置）(multi-nuclide
廃棄前保管removal system)

Medium
中深度
depth
処分
disposal

再利用

Equipment for

HICへの充填材注入設
injecting fillingmaterial備into HIC

Additional solid waste
増設固体廃棄物貯蔵庫
storagevault (Stream（ストリームＳ６、Ｓ７）
S6, S7)

The red frame is an
example of reflectionof the study results
for refining thetreatment flow

Incinerated ash
焼却灰保管
storage(Stream（ストリームＳ８）
S8)

Incinerator
pre-treatment facility
焼却炉前処理設備
増設雑固体廃棄物焼却設備
Additional miscellaneous solid
waste incineration facility

Temporary
汚染土一時保管施設
storagefacility forcontaminated soil

Temporary
一時保管
storage

Equipment for
injecting filling
角型鋼製容器への充填
material
into the材注入設備

square steel
container

廃棄前保管

Decontamination
facility
除染設備

地層
処分

(中深度以深)Additional solid

Waste that is
解体に伴って発生する
likely to be
可能性のある廃棄物

generated during
dismantling

Volume reduction
treatment
減容処理設備removal facility)

Equipment for injecting
角型鋼製容器への充填
filling material into the
材注入設備
squaresteel container

Decontamination
除染スラッジ処理
sludgetreatment
２次廃棄物等
Secondarywaste, etc.

Miscellaneous solid
雑固体廃棄物焼却設備
waste incineration（ストリームＳ８）
facility (Stream S8)

TemporaryTemporary
仮設処理装置・タンク等
一時保管
storagetreatment
deviceand tank, etc.
Secondary waste, etc.
２次廃棄物等

原
子
炉
・
建
屋
の
解
体
廃
棄
物

Dehydration
スラリーの脱水処理
treatmentof slurry

S6, S7)

Temporarystorage
一時保管（ストリームＳ８）(Stream
S8)

Contaminated
汚染土壌等
soil, etc.

Facility for storing secondary waste
水処理二次廃棄物保管施設
generatedfrom contaminated water
treatment（ストリームＳ６、Ｓ７）
equipment (Stream

Preliminary
廃棄前ofstorage保管
waste

Equipment for injecting

吸着塔（SARRY,KURION、
filling material into theadsorption tower (SARRY,
多核種除去装置）へのKURION,
multi-nuclide
充填材注入設備（ストリームＳ６、Ｓ７）

Temporary
可燃物一時保管storage of（ストリームＳ８）
combustibles(Stream S8)

伐採木Felledtrees

Dehydration
濃縮廃液スラリーの
treatment of theconcentrated
脱水処理
liquid waste slurry

Waste
廃棄体化
conditioning

Disposal atintermediate depth or
moretreatment,
前処理・処理treatment,
・再加工

Temporarystorage
一時保管

Preliminary storage

Generationof waste
廃棄物の発生waste storage
増設固体廃棄物貯蔵庫

Equipment for

vault

200Lドラム缶への充填
injecting filling
材注入設備
material
into 200L

Secondarywaste, etc.

２次廃棄物等

drum can

燃料デブリ取出しに伴Waste generated
う廃棄物

in association
with fuel debris
retrievalFlame-retardantliquid waste
難燃性液体廃棄物(PCB,
EHC,(PCB,EHC等)
etc.)
Metal
(items金属
stored in SFP,
（ＳＦＰ、サイトバンカ
保管物等）
site
bunker)Resin
樹脂

Facility
for storing fuel debris and fuel
燃料デブリ・燃料デブリ付着物* 保管施設※
debris deposits
Storage
保管

Contaminated
汚染した難燃物
flame retardant
substances

Fuel debris燃料デブリ付着物
deposits：運用中、発生中（低線量瓦礫や伐採木が地層処分となるなど）(Low dose debris or felled trees undergo geological

Stored waste

Storage
保管

保管廃棄物についてはneeds to be
今後検討が必要studied in the

futureFacility
for storing fuel
燃料デブリ・燃料デブリ
debrisand保管施設
fuel debris
付着物deposits：計画中This diagram will be revised based on
本図は、検討の進捗を反映して見直していく
the progress of the study.

disposal, etc.)
現状の統合結果では、：研究中、将来検討、将来発生
The
current integrated results are not quite convincing.
：デブリ関連納得感は得られにくい（デブリ取出しと要調整）※燃料デブリ取り出しに関する研究開発の検討結果を反映していくものとする
2019/9/23東芝ｴﾈﾙｷﾞｰｼｽﾃﾑｽﾞ作成

Fig.1 Example of reflection in the waste stream (Dashboard)©International Research Institute for Nuclear Decommissioning262

【Confidential】

d. Integration of R&D- Summary Summary of FY2019 Results
1.
2.
3.

The waste classification was reviewed and defined in detail, and summarized in the Waste
Management Sheet.
The ideal form of storage (proposed) was studied and the required treatments were examined.
The cases in which the best waste package performance is not expected were studied, various
R&D results were incorporated and a waste stream proposal that indicated the current status
and required the least amount of treatment (the load on the disposal side is large) was
developed.

Main issues in the third stage highlighted through the R&D in FY2019✓✓

The integrated results of the R&D outcomes for Case 1 suggested based on the current evaluation thatcountermeasures such as solidification or man-made barriers would become necessary for near surface ordeeper disposal even for the comparatively low radiation wastes such as the low radiation debris (surface
dose rate of 0 to 5 μSv/h) or the felled trees. A further study is necessary from a rational viewpoint.
Especially, regarding the settings for important nuclides (C-14 and I-129) which are below the detection
lower limits, it is important to improve the degree of certainty of the inventory presumptions that are done on
the basis of the contamination mechanism.
During the study of Case 1, it was assumed that the harmful gap countermeasures were necessary during
the waste conditioning treatment, and the use of the current storage containers as waste package
containers and the verification and validation of the waste package performance processes were assumed
to be the preconditions. In order to ensure feasibility, it is necessary to verify the technical support forthese preconditions with respect to each type of waste.

FY2020 Plan (Proposed)
A waste stream will be developed as Case 2, in case that existing concepts
concerning the light water reactor can be applied to waste management inFukushima Daiichi NPS. (Excluding disposal)©International Research Institute for Nuclear Decommissioning263

【Confidential】

3. R&D Management©International Research Institute for Nuclear Decommissioning264

【Confidential】- Human resource development, knowledge gathering, clarification of
conditions and specifications, setting indicators ◼ Mid-and-long-term human resource development
 The IRID Symposium 2019 was held in Iwaki city, Fukushima Prefecture on August 1,
2019. At the symposium, an exhibition panel was provided to introduce research results
and future plans of the project for promoting understanding of research and supporting
educational activities.
 At the Fall Session of the Atomic Energy Society of Japan (Sept 13 - Sept 15, 2019),
presentation on 12 research results were delivered to promote understanding ofresearch and support educational activities.
◼ Gathering of knowledge from Japan and overseas
 Cooperation with research institutions (CRIEPI) has been continued since FY2014.
Furtheremore, information on the same cases was obtained from research institutions
overseas and investigation on the information was conducted. Useful information was
acquired from experts and experienced persons.
 The IRID design review meeting was conduced to confirm the validity of the output thatis an objective of the project. (on June 4, July 19, December 10, 2019 and February 25,
2020)◼ Clarification of test conditions and development specifications
 The IRID design review meeting was conducted to exchange opinions with experts
from and outside IRID concerning preconditions, hypothetical conditions and details of
the project output, etc. on June 4, 2019. The test conditions and specifications wereclarified.
◼ Setting an index for determining goal achievement
 To start an index setting, an index for determining the achievement of project goals
were considered, and the targeted Technology Readiness Level (TRL) was determined.
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【Confidential】- Coordination with other R&D, and research and operation management◼ Coordination with the reactor decommissioning task and other R&D
 Information related to the progress made by each project and the challenges faced wereshared by conducting regular meetings between concerned persons from related projects.
Moreover, the mechanism (archive) for sharing project plans, management of information on
results, which has been developed until now, will continue to be developed and utilized.
 In order to carry out thorough research and development related to treatment and disposal ofwaste generated during fuel debris retrieval, without leaving anything out, the R&D teamparticipated in the working of the Canister Project and the Retrieval Project so as to be able toshare information from the respective projects.
(April 22, May 20, June 17, 2019 and February 18, 2020)◼ Research management
 Persons concerned with this project met (Waste treatment and disposal project coordination
meeting) every month on a regular basis. During the meeting the R&D schedule and progressstatus were shared, and in addition, policies for resolution of issues encountered were
decided and accordingly measures were implemented.
April 5, May 10, June 6, July 4, August 2, September 5, October 3, November 5, December 6,
2019 and January 9, February 2, March 2, 2020##Meeting conducted in writing◼ Reflection of site information of Fukushima Daiichi, etc. and operation management
considering on-site applicability
 A “Memorandum of understanding related to initiatives for decommissioning Fukushima
Daiichi” was exchanged between IRID and TEPCO, the required management system wasdeveloped in cooperation with TEPCO and appropriate safety measures were implemented.

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【Confidential】- Project report, communication of information, alternative plans -◼ Project report The first debrief meeting was held on October 29, 2019 and in addition, a progress
report was submitted to the MRI Secretariat every month-end.
◼ Enhancing communication of information
 The results of this report were posted on the IRID website in an attempt tocommunicate information to the general public in an easy-to-understand manner.
Moreover, the results of this project were communicated by holding presentations at
various conferences held within Japan and overseas by AESJ or overseas institutions,etc.
◼ Advance preparation of alternative plans Alternative plans were studied in advance as required, in preparation for circumstances
during the project period, wherein the project does not progress as planned. However,such circumstances did not arise during this project.

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