Subsidy Project of Decommissioning and Contaminated Water Managementin the FY2016 Supplementary Budgets

R&D for Treatment and Disposalof Solid Radioactive Waste
Accomplishment Report for FY2017

February 2019
International Research Institute for Nuclear Decommissioning (IRID)

All Rights Reserved. International Research Institute for Nuclear Decommissioning©International Research Institute for Nuclear Decommissioning

No.1

Contents
1. Project Overview and Research Approach
2. Characterization
3. Study on Predisposal Management
4. Study of Disposal Strategy and Safety Assessment
Methods Suitable for Characteristics of Solid Waste
5. Integration of R&D Results (Study on Waste Stream)
6. Schedule and Project Organization©International Research Institute for Nuclear Decommissioning

No.2

1. Project Overview and Research Approach©International Research Institute for Nuclear Decommissioning

General Overview

No.3- Background and Purpose of the R&D 

The project of decommissioning and the contaminated water management for the
Fukushima Daiichi Nuclear Power Station(NPS) of Tokyo Electric Power Company
Holdings, Inc. (hereinafter, TEPCO) are ongoing according to “The Mid-and-LongTerm Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi NPS”
(hereinafter, Mid-and-Long-Term Roadmap) and “The Progress Status and Future
Challenges of the Mid-and-Long-Term Roadmap toward the Decommissioning of
TEPCO's Fukushima Daiichi NPS*1.”



Under such a circumstances, research and development (R&D) of technologies forsolid radioactive waste treatment and disposal was performed for the nuclear
decommissioning and contaminated water management according to “The Mid-andLong-Term Roadmap and the Technical Strategic Plan 2017 for Decommissioning of
the Fukushima Daiichi NPS of TEPCO Holdings, Inc. *2 “(hereinafter, TechnicalStrategic Plan 2017).

*1. The 39th Session of Team for Countermeasures for Decommissioning and Contaminated Water Treatment/Secretariat Meeting (2017)*2. Technical Strategic Plan 2017 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 (NDF) in 2017
©International Research Institute for Nuclear Decommissioning

General Overview

No.4- Mid-and-Long-Term Roadmap PolicyThe basic concept of waste management* (excerpts from description of research anddevelopment (R&D) , partly reworded)–
The characteristics of solid radioactive waste, such as nuclide composition and activity
concentration, need to be understood, to study solid waste treatment and disposal
methods.
–

Generated solid waste shall be stored and managed by safe and streamline methods
and procedures based on their characteristics.

–

A method to determine waste stabilization and solidification methods (pretreatment
methods) on a rational basis shall be established, and pretreatment methods shall be
determined by the established method before determining the technical requirements
for disposal, in order to ensure the safety of solid waste storage and management.

–

To promote effective R&D on solid radioactive waste treatment and disposal, R&D
projects related to the characterization, treatment, and disposal of solid waste work
closely together. R&D is promoted by the sharing of research and issues among R&D
teams, overviewing all activities of solid waste management, and identifying required
R&D tasks.

The waste treatment and disposal measures, and their technical prospect of the safety shall be proposed by around FY2021.
*Revised on September 26, 2017
©International Research Institute for Nuclear Decommissioning

General Overview

No.5- Policy of Technical Strategic Plan 2017 Strategic proposals for solid radioactive waste treatment and disposal* (partly reworded)–
Focusing on waste characterization, storage and management as predisposal
management until the prospect of disposal can be obtained.
Item

Promotion of characterization

R&D task•

Establishment of a solid waste characterization method thatcomplementarily combines evaluation data based on analysis data andmigration models.

•

Optimization of analysis sample numbers, the simplification andspeeding-up of analysis methods, etc.

•

Study on estimation methods and management of the volume ofhydrogen gas production from the secondary wastes generated fromcontaminated water treatment during the storage and management ofsolid waste.

•

Study on methods to store and manage of solid wastes generated byfuel debris retrieval

Establishment of pretreatment
selectionmethod considering the possibilityof disposal•

Establishment of a method for selecting waste treatment based onsafety evaluation results of in-process wastes for multiple waste
treatment.

Promotion of effective R&D by overviewing
all activities of solid waste management•

R&D is promoted by sharing research progress and issues among
projects, overviewing all activities of solid waste management, andidentifying required R&D tasks.

Thorough storage and management* Technical Strategic Plan 2017 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company
Holdings, Inc., by Nuclear Damage Compensation and Decommissioning Facilitation Corporation©International Research Institute for Nuclear Decommissioning

Approaches to R&D- R&D Planning and Implementation -No.6

 Specific plans were established for each of the following four main R&D
items respectively: (1) Characterization of solid waste; (2) Predisposal
management of solid waste (thorough storage and management); (3) Studyof disposal strategy and safety assessment methods suitable for thecharacteristics of solid waste (establishment of pretreatment methods
considering disposal strategy); and (4) Integration of R&D outcomes (thepromotion of effective R&D by overviewing all activities of solid waste
management).
– The plans were developed based on the characteristics of waste.
– The plans were created in reference to the process chart of the TechnicalStrategic Plan 2017 to ensure the steady implementation of the Plan andto achieve the goals.
– The plans were implemented under appropriate role assignments andwith active information exchange to enable the concurrent progress of
R&D activities.
 Assessment indexes for achieving goals were established and implemented.
©International Research Institute for Nuclear Decommissioning

No.7

Approaches to R&D- Estimated Characteristics of Waste Generated by the Fukushima Daiichi
Accident for the Consideration of R&D Approaches  Waste generated out of control due to the accident
 Contamination originated from nuclear fuel in the reactor core of Unit 1 to Unit 3 * The amount of produced waste will fluctuate with the varying status of decommissioning work.
 Currently only a limited amount of samples are available due to the access difficulty.

Dispersion and diffusion

Water treatment
systemRubble, felled trees, etc.
RubbleFelled trees

Contaminated
water

Soil

Large volume, spreading into wideareas
Trees and soil a poor experience intreatment and disposal
Mainly surface contamination bydispersion and diffusion with some
penetrating contamination by way ofaccumulated water

Secondary wastes generated fromcontaminated water treatment
Secondary waste
fromcontaminatedwater treatment

Generated waste from fueldebris retrieval
Large amount and high radioactivity
Currently some waste is difficult tosample due to the access difficulty.

Replaced pipes,
storage tanks, etc.

A poor experience in treatment anddisposalSome waste at the site is difficult tosample.
The amount of produced waste andnuclides can be partially estimated basedon system features*Contamination originated from activated materials and reactor operation waste may be included.
©International Research Institute for Nuclear Decommissioning7

Approaches to R&D

No.8

Role-Sharing among Relevant Organizations, and Scope and Viewpoint of Study
Formulation of total
strategy

NDF[Scope]•Design, production and introduction

Implementation of
long-term storage[Viewpoint]• Priority, safety, economic
rationality and otherrequirements indecommissioning work

TEPCO

Materialization of waste management(from generation to storage)

TEPCO

Formulation of waste management(treatment and disposal)

NDF/TEPCO

Information flow

IRID

[Viewpoint]• Technical feasibility (Safety)(Basic information needed for evaluation from an
economically rational viewpoint, to be collectedby parameter surveys, etc.)[Scope]• Measures for issues to be toward theintroduction of pretreatment methods• Study of technical matters required fordesignStudy on predisposal
management(Storage and management)• Study of safety measures required forthe storage of highly radioactive waste• Evaluation of technology to stabilize
the secondary wastes generated fromcontaminated water treatment[Risk reduction measures during storage]

Integration of R&D results(Study of waste streams)

Characterization

Study on predisposal
management(Applicability assessment,
reduction of waste amount)• Assessment of applicability depending on
the characteristics of solid waste• Research on waste amount reduction
technologies• Comprehensive assessment of progress, theconsistency of the results and remaining
issues based on research results• Coordination of R&D themes that willcontribute to future R&D• Collection and management of analysis data
• Improvement of the accuracy of analyticalevaluation methods• Summary of a comprehensive inventory
evaluation• Response to materials with impact on
disposal, etc.

Study of disposal strategy andsafety assessment methods
suitable for characteristics ofsolid waste• Detailed investigation and evaluation of disposal
methods inside and outside Japan• Study of solid waste disposal strategy suitable for
their characteristics and methods to assess thesafety of the strategy, such as a safety assessment
scenario[Comprehensive waste management measures]©International Research Institute for Nuclear Decommissioning

No.9

2. Characterization©International Research Institute for Nuclear DecommissioningNo.10Contents of Report(I) Collection and Management of Analysis Dataa. Identification of contamination distribution(a) Development of analysis plan(b) Sampling and transportation(c) Conducting analyses(d) Study of waste classification based on analysis
datab. Development of sampling technology(a) Sampling of secondary wastes generated fromcontaminated water treatment, including sludge(b) Study of sampling methods in the Reactor
Buildings.

c. Streamlining of analysis methods(a) Study on the migration behavior and the contamination
mechanism of radionuclides(b) Study on the representativeness of analysis data(c) Reselection of nuclides to be analyzed(d) Study on more efficient and reasonable analysis
methods
d. Analysis data management(a) Establishement of analysis database(b) Organize and update waste data
(II) Accuracy Improvement of Analytical Evaluation Methods"Comprehensive inventory evaluation" will be summarized in
FY2018.

(III) Measures for Materials with Impact on Disposal
Correspondence of action plan with the Technical Strategic Plan 2017*
Item/FY
Main events planned in thecurrent Roadmap

I. Characterization
1.Collection and managementof analysis data
2. Accuracy improvement ofanalytical evaluation methods
3. Summary of comprehensive
inventory evaluation
4. Response to materials with
impact on disposal, etc.

2nd period (the period until the start of fuel debris retrieval)
2014

2015

2016

2017

2018

2019/after

Summarizing basic concepts about treatment and disposal [C]Preparation for sampling
rubble, ALPS, soil,
incineration ash, and highly Sampling of rubble, ALPS, soil, materials in the reactor building,incineration ash, and highly radioactive materials, the streamlining
radioactive materials, aswell as data releaseof sampling methods, and the creation of a database
Development of methods to evaluate thesecondary wastes generated fromcontaminated water treatment, rubble,
felled trees, and soil

Development andupdate of analysis plan

Improvement of the accuracy of analytical inventory
evaluation that takes into account the variation ofanalysis results
Comprehensive assessment of analysis data and radioactive
inventory estimates, inventory estimation, and the establishmentof updating procedure

Collect and organize points of view pertaining to the provisional
value of acceptable concentration in predisposal managementand disposal facilities

Actions in
response toprogress ofsampling andanalysis
Upgrading ofevaluation
methods

Preparation foranalysis andevaluation of
impact* Technical Strategic Plan 2017 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 (hereinafter referred to as NDF) in 2017.
©International Research Institute for Nuclear DecommissioningNo.11(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (a) Development of analysis plan
FY

Implementation plan2017

• A mid- and long-term analysis plan will be developed based on the following
information: the target schedule and the timings of judgment specified in the Mid-andLong-Term Roadmap, analysis data accumulated to date and knowledge about various
contamination behaviors, and the foreseeable availability of analysis samples.
• The following samples will be collected in analysis targets: rubble, soil, incineration ash,secondary waste generated from contaminated water treatment, contaminated water
stagnated at the basement of the building, etc.
• An annual analysis plan will be developed.

• A mid- and long-term analysis
planning• An annual analysis planning2018• An annual analysis plan will be developed.

• Annual analysis planning.

Implementation details

LegendBlue: Plan
Red: Result

Goal achievement index

FY2017
4

5

6

7

8

9

10

11

12

1

2

Notes

3

I. Characterization(I) Collection and Management of Analysis Dataa. Identification of contamination distribution- Development of analysis plan

Annual analysis
planning

Mid- and longterm planning

Planning fornext year• A draft of next year’s annual
plan has already beendeveloped.

*Analysis plans will be reviewed as necessary.

©International Research Institute for Nuclear DecommissioningNo.12(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (a) Development of analysis plan Analyses are tools to investigate methods for waste management (including storage, treatment and disposal). They provide
information needed to classify waste at each stage of the waste management process. The functions of the individual stages of
the waste management process were clarified, and needs required for analysis arising out of those functions were laid out.
 Analysis items vary with the individual stages of the waste management process. (Figure 1)
 Waste that may pose high risks during storage and treatment (waste conditioning and stabilization) have high priority.
 The procedure to determine disposal category and activity concentration acceptance criteria applied to waste bodies, requires
the activity analysis of materials, including difficult-to-measure nuclides. In addition, not only radioactive materials but alsochemically hazardous materials need to be detected by the analysis. Thus, the volume of analysis work is enormous. (Figure 2)
Characteristics required for storage• All kinds of waste must be tested before being acceptedat storage facilities. Extra attention must be paid to thesecondary wastes generated from contaminated water
treatment (such as adsorption vessels, sludge, and slurry)
since those are main sources of risks.
• Requirements for safety storage– Dose rate as an exposure index– Hydrogen and other gas generation - Concentration of
main nuclides and decomposing materials such aswater– Amount of nuclear materials as a criticality risk index

Characteristics required for treatment(waste conditioning)• All types of waste must be tested before being
accepted at treatment facilities. Extra attention
must be paid to waste having no prior results(such as secondary wastes generated fromcontaminated water treatment, incineration ash,and soil).
• Requirements for selecting treatment method– Physical and chemical properties, theconcentration of main nuclides, etc.

Volumereduction andstabilization
Waste
generation

Primary
classifica
tion

Temporary
storage

Secondary
classifica
tion

Storage

Items of higher priority ☜ Conducting analyses

Tentativeclassification of waste(Nuclide composition)
Evaluation accuracy
improvement

Tertiaryclassifica
tion

Waste bodytreatment

Waste
classification

Undergrounddisposition

Analytical inventory evaluation
Radioactive inventory

Characteristics required for treatment (stabilization)• As with the storage stage, waste that is considered a main sourceof risks (such as adsorption vessels, sludge, and slurry) areprimary targets.
• Requirements for selecting treatment method– Hydrogen gas, heat generation and exposure - concentration of
main nuclides– Gas generation - decomposing materials such as water

Characteristics required for disposal(including waste body verification*)• Identification of disposal category and determinationof waste body check method are critical• Requirements for quantification of inclusions– Concentration of radioactivity including those ofdifficult-to-measure nuclides– Concentration of chemically hazardous materials

Investigation of the content of materials
with impact on disposal and otherhazardous materials

Characteristics of waste(chemical properties)

Activity concentration
determination procedure
Consideration of
disposal strategy
Disposalstrategy

Consideration of scaling
factors

Consideration of safetyassessment methods
Safety assessment
methodDisposal
categorization

Safety assessment
for disposal
ImportantnuclidesSelection of
nuclides(*Check the disposal safety and of waste bodies of integrity)

Figure 1. Issues of Characterization Required forClassification

Figure 2. Illustrative diagram of activity concentration analyses in different
stages in the context of the whole disposal process©International Research Institute for Nuclear DecommissioningNo.13(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (a) Development of analysis plan Analyses that need to be performed with high priority in the mid-term period through 2021 were extracted
based on the consideration of the purpose of analysis. (Table 1)
 Priority was quasi-quantitatively determined by assigning scores to each purpose, and analysis sample
collection timing was presumed. Based on the preceding information, a table that summarizes the collection ofsamples and the implementation of analyses by year was created. (Table 2)
Table 1. Requirements for High-Priority Analysis in
Mid-Term Range
Classification

Storage

Treatment

PurposeAnalysis item

Nuclear material
amount(Criticality)

Fuel debris containing
waste

Total α concentration

Amount ofradioactivity(Measure forhydrogen gas
generation)

Slurry, sludge fromdecontamination systems,
waste from fuel debris
retrieval

Total α concentrationand total β, γ
concentration

Chemical form

All

Concentration ofelements (whennecessary)

Stabilization ofsludgeSludge fromdecontamination systems

Density, dry mass,
particle size
distribution, pH,chemical composition,
main nuclide
concentration, thermal
conductivity, electricalconductivity, organic
substances (TOC)

Solidification ofincineration ash
Soilclassification

DisposalTarget waste

Waste
classification

Incineration ash

Soil

Rubble, incineration ash,waste from dismantlementof structures andequipment outside thePCV, carbonate/ironcoprecipitation slurry,sludge fromdecontamination systems,
waste from debris retrieval,cesium adsorption vessel

Time limit foranalysis
implementation
To be matched withthe timing of theconsideration anddetermination ofstorage methods

Classification

Number of collected samples

Target waste
2017*

Rubble

2018

2019

2020

20

20

Already generated waste (cover soil
storage vessels, etc.)

135

20

Already generated (stored in containers)

10

10

20

Vicinity of areas where materials
extravasated out of the PCV are observed
and other areas(no contact with water)
To be matched withthe timing of theconsideration anddetermination oftreatment methods.

40

20
20

Multi-nuclide removal system (adsorbent)

2018

2019

2020

2021

195

20

20

20

20

0
0

40

60

0

20

20

3

15

5

5

5

20

20

5

20

80

10

10

40

0

20
20

5

10

10

20

21

3

3

5

10

21

25

5

5

5

5

20

Reactor building

0

Contaminated water Turbine building(stagnant water)
Centralized radiation waste treatment

Correlation between
particle size
distribution and themain concentration ofradioactivity

facility

To be matched withthe timing of theconsideration anddetermination of
disposal methods.

80

5

20
3

Total

5

80

Sludge from decontamination systems

coprecipitation slurry)

Total

20

40

Cesium adsorption apparatus (KURION
and SARRY) and adsorbent

Secondary wastes
generated from
Multi-nuclide removal system (carbonate
contaminated water slurry)treatment
Multi-nuclide removal system (ironNumber of analyzed samples

2021

Metal and concrete inside the PCV(Lower portion)
Metal and concrete outside the PCV(contacted with water)

Elemental
composition,
component analysis

Activity concentration
by nuclide(distribution)

Table 2. Annual Analysis Plan

In order to find the input amount to migrate into the said secondary waste, these analyses
are performed according to the sampling/analysis result of secondary wastes generated
from contaminated water treatment.

000Contaminated water Cesium adsorption apparatus(treated water)
Multi-nuclide removal system

10

10

40

40

0

Combustibles

Incineration ash (Tyvek, etc)

5

5

0

Soil

Soil (under atmosphere)

44

44

0

Soil (contaminated water permeated)

6

6

0

PlantsBranches, leaves, etc.

31

Total

329

31
98

40

85

70

622

0
48

58

60

* Includes existing preserved samples.

©International Research Institute for Nuclear Decommissioning

65

231

(I) Collection and Management of Analysis Data

No.14

a. Understanding of contamination distribution - (a) Development of analysis plan As a mid- and long-term analysis plan should be reviewed every few years, potential issues in development of the analysis plan wereidentified to be prepared for future plan revisions. (Table 3)
 The table includes remaining issues to be solved (sample collection, the utilization of analysis resources and data) as wells as
new issues to be addressed (analysis quality assurance).
Table 3 Existing issues in conducting analyses and those that need to be addressed in future planningClassification
Remainingissues to besolvedSample collection

Issues•Waste and samples concerned

Establishment of sample collection methods intended for areas with highradiation such as inside and outside the PCV and for highly radioactive objects
such as cesium adsorption vessels.

•••

New issues tobe addressed

Effective sample collection in cooperation with all decommissioning work andother projects.

•

Sludge from decontamination
systems, projected waste fromplanned debris removal work
Cesium adsorption vessel•

Samples collected in the project toidentify the condition inside reactor
and other projects
Projected waste from debris removal•

Establishment of a method to estimate the radioactive inventory of waste
concerned in consideration of the representativeness of analysis data obtained
from a limited amount of samples. Also with the development of a sample
collection plan that makes use of the method.

•

All types of waste•

Preparation of sample storage areas and methods before the Okuma Analysis
and Research Center starts its operation.

•

All types of waste

Utilization of resources
associated withanalysis•

Reasonable use of existing facilities until the Okuma Analysis and ResearchCenter starts its operation.

•

All types of waste

Utilization of analysis
data•

Establishment of a database accessible for researchers and developers.

•

All types of waste

Analysis quality
assurance•

Establishment of an analysis quality assurance system and the standardizationof analysis methods based thereon with the aim of utilizing existing analysis data
under a regulatory system•

All types of waste©International Research Institute for Nuclear DecommissioningNo.15(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (b) Sampling and transportation
FY

Implementation plan2017

• Regarding slurries and adsorbent produced by the operation of existing and expanded multi-nuclide removal
systems and treated water produced by the operation of cesium adsorption apparatus, secondary cesium
adsorption apparatus (SARRY), and existing and expanded multi-nuclide removal systems, sampling ofthem will be performed in sequence according to the analysis plan and depending on the status of readinessfor the sampling at target locations, which is determined by considering the reliability of the sampling method
and the estimate of exposure dose associated with the sampling.
• Available samples are collected in cooperation with on-site operation and transported to analysis facilities.

• Samples are collectedbased on the annual
analysis plan andtransported to analysis
facilities.

2018• Regarding slurries and adsorbent produced by the operation of existing and expanded multi-nuclide removal
systems and treated water produced by the operation of cesium adsorption apparatus, secondary cesium
adsorption apparatus (SARRY), and existing and expanded multi-nuclide removal systems, sampling ofthem will be performed in sequence according to the analysis plan, subject to the confirmation ofaccessibility to them.
• Available samples are collected in cooperation with on-site operation and transported to analysis facilities.

• Samples are collectedbased on the annual
analysis plan andtransported to analysis
facilities.

Implementation details

LegendBlue: Plan
Red: Result

Goal achievement index

FY2017
4

5

6

7

8

9

10

11

12

1

2

3

1. Characterization(I) Collection and Management of Analysis Dataa. Identification of contamination distribution- Sampling and transportation(Oct 4)(Nov 21)(Dec 15)(added in Nov 10)(added in Mar 29)* Sampling will be performed throughout the year. Transport dates of samples may bechanged when necessary, depending on the readiness of them.

©International Research Institute for Nuclear DecommissioningNo.16(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (b) Sampling and transportation
Items in red text are samples collected in this project.

Table 4. Summary of Sample Transport in FY2017 and Analysis Status

Classification

SampleRubbleContaminated
water

Type

Destination*1

Date

Analysis status

Cores samples obtained by the boring of concrete on the 1st to 5th
floors of Unit 4 reactor building (R/B)

8

1F

NDC

Dec 15, 2017

In process

Samples
associated withdecontamination
tests

Materials used for decontaminating floors and walls in R/Bs of Unit
1 to Unit 3 R/B (flannel cloth, strippable paint, boring core, etc.)

22

JAEA, Oarai

JAEA, NSRI

Mar 29, 2018

Scheduled

RubbleRubble of R/Bs, rubble of stored cover soil, sand gravel of turbine
buildings(radioactivity distribution)

5

1F

JAEA, Oarai

Feb 27, 2017*2

Scheduled

Stagnant water

Stagnant water in Unit 1 to Unit 3 R/Bs

8

1F

JAEA, NSRI

Oct 4, 2017

Reported (Mar 2018)

Stagnant water in the R/Bs, the HTI building, and the RW building,treated water from KURION and SARRY, etc. (uranium analysis)

19

JAEA, NCL

NFD

Nov 10, 2017

Reporting planned

Treated water from existing ALPS

11

1F

JAEA, NSRI

Oct 4, 2017
Feb 27, 2017*2

In process

Treated water from expanded ALPS

12

1F

NDC

Dec 15, 2017

In process

Cerium oxide, activated carbon and chelate resin 2 used in theexisting ALPS

3

1F

JAEA, NCL

Oct 4, 2017

In process

Chelate resin 1 and silver zeolite used in the existing ALPS

2

1F

JAEA, NCL

Nov 21, 2017

Adsorbent

To be analyzed(stored)SludgeSoil

Origin *1

Floor

Treated water

Secondarywastesgenerated fromcontaminatedwater treatment

Transport

Number ofsamples

Sludge from decontamination systems, clear supernatant liquid

2

1F

JAEA, NCL

Nov 21, 2017Partly reported (Mar2018)

Sludge from decontamination systems (elemental analysis, etc.)

1

1F

NFD

Nov 21, 2017

In process

Surface layer of areas F, H, J and K, H4 tank area (measurement
points A and B)

6

1F

NDC

Dec 15, 2017

In process

Areas K and P, H4 tank area (measurement point A) (particle sizevs activity concentration)

3

1F

JAEA, NSRI

Oct 4, 2017

In process*1. 1F: Fukushima Daiichi Nuclear Power Station, JAEA: Japan Atomic Energy Agency, JAEA, NSRI: JAEA, Nuclear Science Research Institute, JAEA, NCL: JAEA, Nuclear Fuel Cycle
Engineering Laboratories
JAEA, Oarai: JAEA, Oarai Research and Development Institute, NDC: Nuclear Development Corporation, NFD: Nippon Nuclear Fuel Development Co., Ltd.
*2 Transported during the FY2014 supplementary budgets project©International Research Institute for Nuclear DecommissioningNo.17(I) Collection and Management of Analysis Data
a. Understanding of contamination distribution - (c) Conducting analyses

FY

Implementation plan

Goal achievement index

2017
2018

• Sample analysis will be performed according to the analysis plan.
• Besides activity concentration, other characteristics necessary for the storage and managementof waste will be analyzed.

• Implementation of analysis and thereporting of analysis data in line withthe annual analysis plan.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

3

1. Characterization(I) Collection and Management of Analysis Dataa.. Identification of contamination distribution- Conducting analyses

Reporting* Analysis data will be reported at around the end of every quarter term. Reporting timing may be changed
depending of the content of analysis data.

©International Research Institute for Nuclear DecommissioningNo.18(I) Collection and Management of Analysis Data
a. Understanding of contamination distribution - (c) Conducting analyses

 The analysis of core samples obtained by the boring of structures in the Unit 4 reactor building is underway.
 To evaluate representativeness of analysis samples, local contamination distribution within samples from various types of rubble
is investigated using imaging plate (IP) autoradiography. (Figure 3 and 4)
 An analysis of rubble (concrete) showed inhomogeneous contamination distribution in general, which suggests that a series of
events that occurred in different parts of the unit during the accident had caused difference in contamination distribution.
 It is necessary to take account of the representativeness of samples for the estimation of the amount of radioactivity becauseof the possibility of inhomogeneous contamination distribution.

(1) Sample(2) IP-autoradiography

Surface (untreated)• The contamination levelof the light blue coating
is low (The gray dottedline indicates the profileof the sample).
• Contamination on theside is shown in theimage.

• The white coating is
obviously
contaminated (Thegray dotted line
indicates the profile ofthe sample).
Surface (untreated)• No contamination isobserved (andpenetrated) in theinterface between thecoating and concrete.
Low

High

Dose rate(4) Scene of sample(3) IP-image sample
surface grinding• Place a sample (1) on an imaging plate (2), andcapture an image (3).
• The imaging plate changes the color of portions
corresponding to the amount of received radiation
dose. A profile in image is generally different fromthat of the sample because the image plate also
senses radiation from part of the sample not in a
direct contact with the plate.
• Repeat the imaging process by grinding thesurface of the sample while maintaining its position
same as that on the imaging plate (4) and thenimaging a newly created cross-section surface.

Figure 3. Method of IP Analysis

After 5 times of grinding

After 10 times of grinding• No contamination wasobserved inside thesample. Similar images
were obtained from thecross-section surfaces
created by 11 or moretrials of grinding.

After 5 trials of grinding• The IPautoradiography
results after 5 trials ofgrinding infers
penetration ofcontamination into theinterface.

After 10 trials of grinding• No contamination wasobserved inside thesample. Similar
images were obtained
from the cross-section
surfaces created by
11 or more trials ofgrinding.

Rubble collected around Unit 3

Rubble collected on operation floor of Unit 3(Sample ID: 3U-01)(Sample ID: 3RB-OP-C1)

Figure 4. Local Contamination Distribution of Rubble (Left photos: appearance, Right: IP images©International Research Institute for Nuclear DecommissioningNo.19(I) Collection and Management of Analysis Data
a. Understanding of contamination distribution - (c) Conducting analyses

 The analysis of water samples collected from stagnant water is underway.
 Samples taken from stagnant water Unit 1 to Unit 3 R/Bs (reactor buildings) were analyzed (Figures 5 and 6).
 Stagnant water samples from Unit 1-3 reactor buildings show a single-digit variation in 90Sr/137Cs ratio (between around 101and 100). The ratios are nearly equal to those of stagnant water in the turbine buildings. There seems no impact of the change
in the amount of nuclear reactor water injection that was implemented around last March (2017).
 Although the 238Pu/137Cs ratio is high on the stagnant water in primary containment vessels (PCV), there is no significant
difference in this ratio among stagnant water samples from R/B, TB (turbine building) and centralized RW (centralizedradwaste building). The ratios of stagnant water in Unit 1 and Unit 2 R/Bs are larger than that in Unit 3 R/B by about one digit.
 Since different levels of activity concentration are detected depending on the type of nuclides, unit number, and thecontaminated water flow path (reactor buildings are on the upstream side and the high temperature incinerator building is on
the downstream side), it is important to investigate the migration of contamination on the upstream side continuously toestimate contamination by contact with contaminated water.
1E+09

1000

10

100

Legend

2T/B

1T/B

P

P

P

4T/B

3T/B

P

P

P

8
100000000

P

90Sr

P

P

3R/B

4R/B

P

Stagnant water was sampled from theblue pipeline for Unit 1 R/B, the red
pipeline for Unit 2 R/B, and the green
pipeline for Unit 3 R/B respectively at thehigh temperature incinerator building. *1

concentration (Bq/g)

P

P

10

1000000
6

100000

10

4
10000

1000

102

101

100

Unit 3 R/B*
Unit 1 TB

1100

Unit 2 TB*
0.1

Unit 3 R/B*
Unit 2 PCV

0.01
-2

10

Unit 3 PCV
0.001

10-1 10-2

10

101
10

100

3
1000
10

10-3

10000

5
1000001000000 10000000100000000
10
107

137Cs放射能濃度（Bq/g）
137Cs activity
concentration (Bq/g)Centralized RW

10-4

10

-4
0.0001

2
100
10

1

Figure 5. R/B Stagnant Water Sampling

238Pu activity
238Pu放射能濃度（
Bq/g）

P

90Sr activity
concentration (Bq/g)

P

90Sr放射能濃度（Bq/g）

P

2R/B

Unit 2 R/B*

10

10000000

1R/B

Unit 1 R/B*

238Pu

102

10-5 10-60.00001
1

101
10

10-7 10-8
100

3
1000
10

(White-fill symbols indicate that thevertical axis value of the data point
is lower than the detectable limit.)

10-9
10000

10

10

5 1000000 10000000100000000
7100000137Cs放射能濃度（Bq/g）
137Cs activity
concentration (Bq/g)*Data reported this fiscal year

Figure 6. Concentration of 90Sr and 238Pu in Stagnant Water from different sources(correlation with the concentration of 137Cs)*1

This information is cited from the material 5 of the 53rd meeting of the Commission on Supervision and Evaluation of theSpecified Nuclear Facilities (on May 22, 2017) in part with some additions.

©International Research Institute for Nuclear DecommissioningNo.20(I) Collection and Management of Analysis Data
a. Understanding of contamination distribution - (c) Conducting analyses

 The analysis of secondary waste generated from contaminated water treatment is underway.
 Sludge from decontamination systems was analyzed (Figure 7).
 The physical properties of sludge that were considered important for the determination of transport andtreatment methods (such as density, ratio of solid content, particle size distribution, and fluidity) wereanalyzed as an initial action.
 Particles with a diameter of 5 μm or less amounted to about 90 percent of all particles contained. 90Sr isconsidered to be a main nuclide as the source of radiation.
 Analyzed carbonate slurry generated by multi-nuclide removal systems (Figure 8).
 The analysis results suggested the influence of the characteristics of contaminated water that was treated by
the systems, such as lower 90Sr concentration compared with that of samples analyzed in the past *1.
OthersAfter left standstill

Appearance of dispersed particles

Appearance

Chemical composition [wt%]Mg(OH)2

31.9

CaCO3

46.6

Na2CO3

5.29

SiO2

6.85

FeO(OH)·H2O

1.06

Al2O3

0.16

Ti(OH)2

0.6

B

0.06

Ce(OH)3

0.04

SrCO3

0.03

Others

7.3

Estimated chemical composition

Figure 7. Analyzed sludge from decontamination systems

Figure 8. Analysis of Carbonate Slurry from Multi-nuclide Removal Systems(Sample ID: LI-AR-SL1-3)(Sample ID: AAL-S8-6)*1

The 34th Session of Team for Countermeasures for Decommissioning and Contaminated Water
Treatment/Secretariat Meeting, held on September 29, 2016

©International Research Institute for Nuclear DecommissioningNo.21(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (d) Study of waste classification
based on analysis data

FY

Implementation plan

Goal achievement index

2017• Radionuclide compositions that determine the characteristics of waste will be identified.
• Data analysis will proceed along the correlation between the concentrations of radionuclides andthe amount of source term standardized by the composition of nuclides (transport rate) for thecomposition identification and the study of classification.

• Proposal of waste classification basedon analyses data.

2018• The results of classification will be reviewed as analysis data is accumulated.

• Proposal of waste classification basedon analyses data.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

1. Characterization(I) Collection and Management of Analysis Dataa. Identification of contamination distribution- Study of waste classification based on analysis data

Study

Summary*The study results are reflected in the study of analysis methods.

©International Research Institute for Nuclear Decommissioning3

No.22(I) Collection and Management of Analysis Data

a. Understanding of contamination distribution - (d) Study of waste classification
based on analysis data Modeling work was started based on the element groups formulated by the U.S. Nuclear Regulatory Commission (NRC). The
reconfiguration of the element groups was attempted by using accumulated analysis data to describe the state of contamination
more accurately.
 The elimination and consolidation of the groups was considered using the frequency distribution of transport rate. Streamlining
due to a decrease in groups to be examined, as well as the improvement of reliability due to an increase in the frequency of
data, are expected for the future study by consolidating nuclides that have similar data distributions. (Figure 9)
 Newly formulated element groups intended for air and water contamination will be used for inventory estimation and otheroperations from this point forward. (Table 5)
Halogens (2) group

Table 5. Expedient Element Groups Formed based on Analysis Data

Transport rate

Data frequency

Data frequency

Tellurium group

Transport rate

No
Contamination via air
1 Noble gases
Semivolatile
2
H
Elements
C, Cl, I, S, Se, Sb,
3 Volatile Elements
Te, Ag4 Alkali Metals
5

Data frequency

Probability density

Volatile elements group

6 Transition Metals

7
Transport rate

Table 9. Reformation of Element Groups based on
Analysis Data

Alkaline Earth
MetalsLanthanides,
Actinides

Na, K, Rb, Cs

Contamination via water
Same as on the left
Soluble Elements

H, C, Cl, I

Chalcogens

S, Se, Te

Same as on theleft
Alkaline Earth
MetalsBe, Ca, Sr, Cd, In,
Ba, RaMn, Fe, Co, Ni, Zn,
Mo, Tc, Ru, Rh, Pd, Transition Metals
Sn
Si, Zr, Nb, La, Nd,
Pm, Sm, Eu, Gd,
Tb, Ho, Tm, Lu, Hf,
Same as on the left
Ta, Re, Os, Ir, Pt,
Tl, Pb, Bi, Po, Am,
Cm, Cf, Ce, Ac, Th,
Pa, U, Np, Pu

Be, Ca, Sr, Ag, Cd,
In, Sn, Ba, Ra
Mn, Fe, Co, Ni, Zn,
Mo, Tc, Ru, Rh, Pd,
Sb(Elements in red text are those that belong to either contamination via air or via water, not to both.)©International Research Institute for Nuclear DecommissioningNo.23(I) Collection and Management of Analysis Datab. Development of sampling technology - (a) Sampling of secondary wastes
generated from contaminated water treatment, including sludge
FY

Implementation plan

Goal achievement index

2017i) Collection of cesium adsorbent, etc.
• The issues that were clarified by the conceptual study in FY2016 (such as the adhesion ofadsorbent) will be studied to substantiate a collection method.
ii) Collection of sludge from decontamination systems• The site survey of D-pit will be conducted to substantiate a collection method to develop thedesign and production of waste sludge collection jigs and to create a construction plan.

i) Collection of cesium adsorbent, etc.
• Show proposed absorbent sampling
methods.
ii) Collection of sludge fromdecontamination systems• Determine a sludge collection method
and produce collection jigs.

2018i) Collection of cesium adsorbent, etc.
• Based on the results of detailed discussions in FY2017, development plan of sampling collection
device will be developed to start designing a mockup device.
ii) Collection of sludge from decontamination systems• Waste sludge from the D-pit will be collected using the produced collection jig.

i) Collection of cesium adsorbent, etc.
• Formulation of a development plan,and the design of a mockup.
ii) Collection of sludge fromdecontamination systems• Collect sludge from D-pit.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(I) Collection and Management of Analysis Datab. Development of sampling technology, etc.
- Sampling of secondary wastes generated fromcontaminated water treatment, including sludge(Collection of cesium adsorbent, etc.)

Preoperational
planning

Discuss specification of specific sampling device foradsorbent from the cesium adsorption vessel(Collection of sludge from decontamination systems)

Survey planning

Survey of periphery and inside of D-pit

Summary

Production of sampling jigs and formulation of
construction plan

©International Research Institute for Nuclear Decommissioning3

(I) Collection and Management of Analysis Data

No.24

b. Development of sampling technology - (a) Sampling of secondary wastes generated
from contaminated water treatment, including sludge
 Method to sample zeolite from the cesium adsorption vessel by remote control
 The concept design of sampling system and the design of element test device to examine the performance and practicality of the sampling head were already
implemented in projects last year or earlier.
 Element test device was built (Figure 10), and tests were conducted with the combination of 9 conditions (boring speed, rotation speed, and tip shape) to evaluate theamount of collected samples. The shape of the sampling head should be reviewed for improvement because the amount of collected samples were still insufficient in theshallow part of the zeolite layer in all conditions, despite a good sampling performance shown by a flat-shaped sampling head.
 Method to sample slurry that was discharged from the multi-nuclide removal systems and stored in
the HICs*1 from those HICs
 The characteristics of slurry samples (dose rate, chlorine concentration, and the timing ofgeneration) vary largely with their origin, and the generation is also large. For these reasons,samples with adequate representativeness were selected. (Figure 11)
 A plan to focus on slurry in HICs stored in No. 2 and 3 storage facilities, and to collect 20 samples
as a target number at No. 2 storage facility was developed. Sampling is performed effectively incooperation with water draining work.
*1. HIC (high integrity container)

Number ofunits (pieces)

Dose rate
Samples are selected sothat mainly the dose rateof them has an adequate
distribution.

Dose

Number ofunits (pieces)

Upper portion of zeolite layer
Simulated column

Chlorine concentration

Sampling head
Outer dimension: ⌀27 x 200 mm
Sampling capacity: approx. 20 cm3Outer dimension: W800 × L1,000 × H2,435 mm
Stroke: 1,000 mm

Figure 10. Element test device for the sampling ofzeolite from cesium adsorption vessel

Chlorine
concentrati
on

Figure 11. Frequency distribution of carbonate slurry(upper graph: dose rate, lower graph: chlorine concentration)©International Research Institute for Nuclear Decommissioning

(I) Collection and Management of Analysis Data

No.25

b. Development of sampling technology - (a) Sampling of secondary wastes
generated from contaminated water treatment, including sludge
 Method of collecting sludge from decontamination systems and the implementation of collection
 As a site survey of the storage area (D-pit), an underwater camera was loaded for surveying the sludge sedimentation andseparation status, the turbidity of clear supernatant water, etc. (Figure 12)
 The image of the underwater camera (Figure 13) indicated that the thickness of accumulated sludge layer was about 40 cm.
The volume was estimated to be just under about 37 m3.
 An investigation tool was made and inserted into the actual sludge layer at the site to investigate the degree of sludge
consolidation as well as to sample the sludge. It is inferred that dense consolidation has not occurred in the sludge layer
because the head of the investigation tool was confirmed to have reached the bottom of the D-pit. Sludge was collected into
the vial container that was attached to the head of the investigation tool. (Figure 14)
 The collected sludge sample was sent to JAEA, Nuclear Fuel Cycle Engineering Laboratories(JAEA, NCL) and Nippon Nuclear Fuel Development Co., Ltd. (NFD), and is being analyzed now. The target was achieved
ahead of the original schedule (sampling in FY2018).

Ground floor:
Decontamination system(AREVA) installation area

Underwater
cameraapprox. 40 cm

Figure 12. Schematic illustration of theSite (D-pit) Survey *Figure 13. Condition of Sludge
Sedimentation (yellow portion)*

Figure 14. 10 mL Vial Container Setup(Right) and Collected Sludge (Left)**Material 2 of the 6th meeting of the Committee on Radioactive Waste Issues of the Specified Nuclear Facilities on July 25, 2017
©International Research Institute for Nuclear DecommissioningNo.26(I) Collection and Management of Analysis Datab. Development of sampling technology - (b) Study of sampling methods in R/Bs

FY

Implementation plan

Goal achievement index

2017• Study on sampling technologies and target sampling locations depending on objectives andpriorities will be conducted based on the outcomes of R&D activities in FY2016.
• Samples will be collected accordingly at target sampling location based on the analysis plan.

• Proposal of target sampling locations
according to objectives and priorities
as well as technologies necessary to
perform sampling• Implementation of sampling at
locations for which a sampling method
has been established2018• Study on target sampling locations and sampling technology will be continuously conducted in
the same manner as FY2017.
• Samples will be collected accordingly at target sampling location based on the analysis plan.

• Implementation of sampling at
locations for which a sampling method
has been established

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(I) Collection and Management of Analysis Datab. Development of sampling technology, etc.
- Study of sampling methods in R/Bs

Preoperational
planning

Consideration of the purpose ofsample collection in the reactor
buildings

Clarification of local
circumstances

Summary©International Research Institute for Nuclear Decommissioning3

(I) Collection and Management of Analysis Data

No.27

b. Development of sampling technology - (b) Study of sampling methods in R/Bs
 Methods for sampling in reactor buildings (R/Bs) by remote control
 The feasibility of sampling in R/Bs was established using a sampling tool that was developed inconsideration of applicability to remote operation during the last fiscal year project. (Figure 15)
 The data of conditions inside the Unit 1 to Unit 3 R/Bs accumulated through past investigations (Figure16) was scrutinized, and feasible access routes and sampling locations and exposure doses associatedwith the samplings were studied.
 A prototype of remote sampling system was designed and produced. The sampling performance andremote operation performance of the prototype is being tested by mounting it on an existing robot(Packbot).

Significant difficulties are
expected to reach these areas.
Difficulties are expected toreach these areas.
Areas of decontamination being
presumably pending (need to
examine whether access ispossible )Areas of decontamination
pending, and significant
difficulties are expected toreachRed text: Limited passage
widthLight blue text: Limited
passage heightBlue text: Name of importantequipment

Figure 15. Sampling tool designed for remote operation (left)
and a scene of concrete sampling in the Unit 4 reactor
building (southwest corridor on 2F) (right)

Purple line: importantequipment
Dose rate in each spot
(mSv/h)Figure 16. Reactor building illustration diagram with sampling locations©International Research Institute for Nuclear DecommissioningNo.28(I) Collection and Management of Analysis Data
c. Streamlining of analysis methods - (a) Study on migration behavior and thecontamination mechanism of radionuclides

FY

Implementation plan

Goal achievement index

2017• The contamination mechanism of radionuclides will be estimated in reference to information
about the positions (locations) and processes where and how waste was contaminated andinformation about the development of the situation after the occurrence of the accident, as wellas knowledge about the waste classification (composition of radionuclides that characterize thewaste) studied in (I)-a.

• Presentation of the estimatedcontamination mechanism based on
analysis data.

2018• The results of classification will be reviewed as analysis data is accumulated.

• Presentation of the estimatedcontamination mechanism based on
analysis data.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(I) Collection and Management of Analysis Data
c. Streamlining of analysis methods- Study on the migration behavior and the contamination
mechanism of radionuclides

Study on contamination
mechanisms

Summary©International Research Institute for Nuclear Decommissioning3

No.29(I) Collection and Management of Analysis Data
c. Streamlining of analysis methods - (a) Study on migration behavior and thecontamination mechanism of radionuclides

The establishment of a model to describe contamination is needed for the estimation of the inventory of waste. It is important to gain
knowledge about the types of radionuclide concentration distribution, to deal with the analysis data of radionuclide concentration.
 The phenomena of contamination can often be described using the lognormal distribution assumption in general. The same is true forradioactive waste.
 A transport rate is useful as an index because it makes it possible to deal with nuclides with similar chemical characteristics as one group.
 A good match to analysis data can often be obtained if the frequency distribution of transport rate is calculated based on a hypotheticalcontamination route and element group of similar chemical characteristics and approximated using the lognormal distribution. (Figure 17)
 There is a possibility of different contamination behavior of elements if a good match is not obtained. Conversely, this result can be used
to consider the grouping of contamination types.
Frequency distribution of transport rate
approximated using the lognormal distribution

These data points are better treatedas those belonging to other groups.

(checked by the Kolmogorov–Smirnov test)

Analysis data

Definition of transport rate T

X,sampleX,source

N std,sample N std,source=

X,samplestd,sample

Data frequency

Probability density

TX =Data frequency

Characteristics
rate
N of transportc
Nc

Probability density

Fitting• The transport rate of element X is calculated as aquotient of a ratio of X elements transported fromthe source term to the sample to the number of X
elements in the source and the same ratio for the
standard nuclide.

AX,source Astd,source• N is the number of atoms, c is a concentration(Bq/g), A is an activity (Bq), X indicates the data of
the nuclide concerned, and std indicates the data of
the standard nuclide (137Cs). A is applied after halflife period correction.
• It doesn’t provide information in transient state
during transport. The transport rate of Cs is 1 (one).

Transport rate

Volatile elements

Transport rate

Rare earths, actinide elements

Figure17. Frequency distribution of transport rate and the suitability of lognormal distribution(Graphical presentation of analysis data on reactor building’s internal contamination)©International Research Institute for Nuclear DecommissioningNo.30(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods - (b) Study on representativeness of analysis
data

FY

Implementation plan

Goal achievement index

2017
2018

• Study of assessment methods for the representativeness of analysis data will be conductedunder the restriction conditions that the amount of the data and/or sampling locations are limited.
Also, methods for estimating the distribution of waste contamination will be studied based on theassessment method established by the aforementioned study.

• Proposal of the assessment method
for the representativeness of analysis
data and the method for estimating
contamination distribution.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(I) Collection and Management of Analysis Data
c. Streamlining of analysis methods- Study on the representativeness of analysis data

Study of data
representativeness

Summary©International Research Institute for Nuclear Decommissioning3

(I) Collection and Management of Analysis Data

No.31

c. Streamlining of analysis methods - (b) Study on representativeness of analysis data
 Waste is continuously generated, and sampling cannot cover all generated wasted due to the volume. It is conceivable thatthese facts make the establishment of clear and reliable waste classification, and the implementation of sampling inconsideration of the representativeness of analysis data, difficult. (Figure 18)
 Based on the assumption that the destructive analysis of difficult-to-measure nuclides does not provide enough information
about representativeness, inventory needs to be estimated.
 The non-destructive analysis and dose rate measurement of waste stored in storage containers are relatively easy and are
expected to provide analysis and measurement data that assumes representativeness. For these reasons, a goal was set toestablish a method that combines these two perspectives. (Figure 19)
 A mid- and long-term analysis plan was considered based on the above viewpoint.
Waste that cannot be sampled
Analysis sample

Radioactive inventory of individual waste

Influence of the continuous
generation of waste anddifficulty in sampling

Information obtained from thestorage and management of waste

Model of contamination
Migration rate specified for eachelement group based on analysis
data (transport rate) (fingerprint; UK)

Uncertainty of classification
Partial analysis
Partial unavailability of samples

Destructive analysis(difficult-to-measure nuclide analysis)
Scaling factor(radionuclide vector; UK)

Sampling to assure the representativenessof analysis data is difficult.

Waste of a certain classification

Figure 18. Illustration to explain difficulties in
obtaining samples with representativenessNon-destructive analysis(analysis of γ-ray nuclides such as Cs)
Conversion factor (Bq/Gy)

Analysis and measurement that Dose rate measurement
assume representativeness(γ/βγ-ray dose rate measurement)

Figure 19 Relationship between the representativeness ofanalysis samples and inventory estimate©International Research Institute for Nuclear DecommissioningNo.32(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods - (c) Reselection of nuclides to be analyzed

FY

Implementation plan

Goal achievement index

2017• Radionuclides to be analyzed will be selected in consideration of the accumulated analysis data,
limited capabilities of the analysis methods, and the importance in waste management. The resultof the selection will be reflected in the analysis plan.

• Presentation of selected radionuclides
to be analyzed.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods- Reselection of nuclides to be analyzed

Selection of nuclides*Conduct a review and the analysis plan to reflect the result to the plan.

©International Research Institute for Nuclear Decommissioning3

No.33(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods - (c) Reselection of nuclides to be analyzed
 The method of difficult-to-measure nuclides analysis needs to be streamlined
because it requires more resources (manpower, facility, and time) than that for
typical nuclides. The list of nuclides to be analyzed was reviewed based on theaccumulated analysis data to date and contamination behaviors (element groups)that were inferred by the analysis data.
 The concentration of some nuclides that weren’t detected in the past (Table 6) canbe estimated by the fuel consumption based-calculation, from the viewpoints ofisotope and chemical similarity, not relying on analysis. In addition, the necessity ofanalysis and calculation was examined by taking into account the importance indisposal safety (Figure 20).
 Twenty-three (23) nuclides were listed as those whose concentration would be
estimated by calculation, while the number of nuclides to be analyzed was reduced
from 38 to 30. (Table 7)
Table6. List of undetected nuclides
Element
H
C
Cl
Ca
Ni
Tc
Cs
Eu
U
Np
Pu
Am
Cm

Detected

Not detected

3H
14C
36Cl
41Ca
63Ni

59Ni

99Tc
137Cs

135Cs

154Eu

152Eu

234, 235, 236, 238U
237Np
238, 239+240 Pu

241, 242Pu

241Am

242m, 243Am

244Cm

245, 246Cm

Radiation protection
to be selected• Refer to ICRP Pub.107

1252 nuclides

Exclusion by halflife• T1/2 ≥1.5 y

134 nuclides
Importance in disposal safety(1 )51 nuclides
Importance in disposal
safety(2 )30nuclidesReflection of analysis results
and streamlining
30 nuclides*
Nuclides to be analyzed
in the future• Nuclides listed as important in
the past evaluation of disposal
safety in Japan• Reflection of the results of investigations
conducted until last fiscal year on waste
generated by the accident• Necessity of analysis by considering
contamination mechanisms, etc.
• Detection record• Possibility of detection in the future• Rationality of calculation-based estimate*23 other nuclides are listed as those
whose concentration is estimated by
calculation.

Figure20. Procedure to select nuclides to be analyzed
considering the safety in disposal

Table 7. Nuclides Subject toAnalysis and Calculation
Element
H
C
Cl
Ca
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Ru
Pd
Ag
Sn
Sb
I
Cs
Sm
Eu
Pb
Po
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm
NumberofnuclidesNuclides to beanalyzed
3
14
36
41
60
63
79
90
93
94
93
99
106
107
108m
126
125
129
137

Nuclide subject tocalculation

59

93m

154

234, 235, 236, 238
237
238, 239, 240
241
244
30

135
151
152
210
210
226, 228
227
228, 229, 230, 232
231, 233
233
241, 242
242m, 243
245, 246
23©International Research Institute for Nuclear DecommissioningNo.34(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods - (d) Study on how to find more streamlined andreasonable analysis methods

FY

Implementation plan

Goal achievement index

2017• Analysis methods that can be streamlined and improved in efficiency will be extracted based on
the analysis plan.

• Listing analysis methods to be
streamlined and improved in efficiency.

2018• Applications of technologies expected to contribute to streamlining and improvement in efficiency
will be studied.

• Presentation of the application scopeof the listed analysis methods

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

3

I. Characterization(I) Collection and Management of Analysis Data
c. Streamlining of analysis methods- Study on how to find more streamlined and rational
analysis methods

Viewpoint of streamlining

Study of streamlining analysis
methodTechnical development plan
discussed©International Research Institute for Nuclear DecommissioningNo.35(I) Collection and Management of Analysis Data

c. Streamlining of analysis methods - (d) Study on more streamlined and reasonable
analysis methods
 For the necessary high-level implementation of complex analysis, technologies expected to contribute to working out more
efficient and rational analysis methods were listed from the viewpoint of cost reduction, worker exposure reduction and data
quality assurance. (Figure 21)
 Promising R&D items were listed including the followings: change in measurement principle (from activity measurement to ICPMS analysis), automation by introducing robots to automate chemical separation operation, and introducing a quality assurance
system (Table 8). The realization of these themes was considered.
Table8. Evaluation of R&D Items Subject to Improvement in
Efficiency and Reasonability
PurposeMeasures
Operation timeshortening

Cost
reduction

R&D issues
Direct analysis of solid samples(for example, introduction ofthermal decomposition of solids
Sophisticated discrimination in
detection(for example, wave height
discrimination in β-ray measurement)

Reduction of workers’
radiation exposure

Reduction ofsampleamountOperation
labor-saving
Data quality
assurance

Standardization ofanalysis methods

Change in measurement principle(for example, from radiation analysis
to mass analysis)
Concentration increase by chemical
separation(for example, development of highly
selective adsorbent)
Integration of a series of steps(for example, application of microchip
analysis)
Chemical separation operation
automated by robots
Introduction of quality
assurance system

Figure 21. Required Improvements inAnalysis and Required R&D Tasks

R&D items

R&D resources

Degree ofeffectCost

Time

Degree ofpriority

Classificationof theme

Direct analysis ofsolid samples

MediumHigh

Long

MediumFundamental

Sophisticateddiscrimination in
detection

N/A

Low

MediumLow

N/A

Change inmeasurement
principle

Large

Medium

Short

High

Application

Concentration
increase by chemical
separation

Small

MediumLong

Low

Fundamental

Integration of aseries of steps

MediumMediumMediumMediumFundamental

Chemical separation
operation automatedby robots

Large

High

MediumHigh

Application

Introduction ofquality assurance
systemLarge

MediumMediumHigh

Application©International Research Institute for Nuclear DecommissioningNo.36(I) Collection and Management of Analysis Datad. Analysis data management
FY

Implementation plan

2017i) Establishment of analysis database• A database through which all persons involved in analysis can share data will beestablished and used to streamline analysis work.
• A database to be opened to the public will be created to utilize accumulated analysis datafor decommissioning effectively.
ii) Organize and update waste data
• The latest waste analysis data will be collected and the data obtained in the FY2016
project will be re-organized. Additional data will be collected and updated accordingly.

i) Establishment of analysis database• Creation and utilization of the databasei) Establishment of analysis database• Issues for operation will be extracted to improve and upgrade functions.

i) Establishement of analysis database• Improvement of the database to solve issues
identified during useii) Organize and update waste data
• Update of the waste list with the latest waste
data and analysis results

2018

Goal achievement indexii) Organize and update waste data
• Update of the waste list with the latest waste
data and analysis resultsii) Organize and update waste data
• The latest waste analysis data will be collected and updated the data obtained in the
FY2016 project. Additional data will be collected and updated accordingly.

Implementation details

LegendBlue: PlanRed:
Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

3

I. Characterization(I) Collection and Management of Analysis Datad. Analysis data management- Building of analysis database

Specification
review- Organize and update waste data

Preoperational
planning

Creation of database
Consideration of data management and update
policy, publication of updated data set

Operation andidentification of issues
Summary©International Research Institute for Nuclear Decommissioning

(I) Collection and Management of Analysis Data

No.37

d. Analysis data management - (a) Building of analysis database In the past, publicly released analysis data was included in a
Data about samples, analysis
published report*1. Methods to publish analysis data on theimplementation, quantitative values, etc.
internet were studied to improve the convenience of uses. In
Analysis data to be public
Partially copied data
addition, methods to share data related to the management of
Database
Database< Subset >samples and analysis work among relevant people werestudied. (Figure 22)
Public serverInternal database To secure the database, public data were extracted andreleased as public contents. (Figure 23)
Sample inventory, analysis data,
Sample collection status,characteristics, and inventory methods, data, and quality assurance The contents built is currently being checked for vulnerability
and will be released on the Internet with the announcement as
Analysis A Analysis B Analysis C
Data viewers
Sampling andsoon as the check is complete*2.
transportation
Internet
Intranet
 The expansion of data coverage and English translation will be
Figure 22. Analysis Database Structure
promoted.

Top page

Graph view (change with time)

Table view (data downloadable)

Figure 23. Examples of Database Contents*1 Analytical data on contaminated water, rubble and other debris collected at the Fukushima Daiichi Nuclear Power Station, JAEA-Data/Code 2017-001 (2017)*2 https://frandli-db.jaea.go.jp/FRAnDLi/©International Research Institute for Nuclear Decommissioning

(I) Collection and Management of Analysis Data

No.38

d. Analysis data management - (b) Organize and update waste data
Information regarding stored waste is essential to the study of waste management. A control sheet has been created for every single waste
generated and stored until now. However, paper-based management is not suitable for the utilization of waste information contained in apaper medium.. To cope with this issue, data items were considered and the information of the selected data items was computerized by
input work, with the aim of making such information electronically accessible in a database.
 Multiple types of control sheets have been created depending on the classification of waste and the time of use. A new database format
was designed by carefully checking data items that form the basis of a new database in order not to miss any important items.
 All the information of the selected items contained in about 47 thousand control sheets was input into the computer system. (Table9)
 The basic structure for the electronic data utilization was established by consolidating all waste volume and dose rate data contained in thecontrol sheets. The following utilization methods are expected:
 Reference for waste storage planning based on a macroscopic view over the characteristics of waste storage amount (volume) and dose
rate. (Figures 24 and 25)
 Identification of materials with impact on disposal
 Utilization in reasonable storage and management planning.


1. Metal rubble

Table9. List of computerized control sheets

Rubble andfelled trees

Waste fromcontaminatedwater treatment
facilities

Period of paper-based
control
From Apr to Jul 2012

Waste type

Number ofcreatedcontrol
sheets80,000Equipment & control
panels <1% Earth and
sand

12%
Metal rubble
15%

No category374

From Jul 2012 to Jan 2014

8

5,282

From Jan 1024 to Sep 201521

17,144

Concrete andasphalt rubble

From Sep 2015 to Nov 201719

23,83018%

From Nov 2017

22

208

From Apr 2012 to Nov 2017

8241

70,000Combustible
andincombustible
materials

60,00050,00040,000Volume (m3)Classification

90,000

Others 5.4%

2. Concrete and asphalt
rubble30,00026%

20,00010,000Felled trees

23%

Figure 24. Percent of Waste
Storage Amount (Volume) byWaste Type0

4. Combustible and incombustible materials

Surface dose rate (μSv/h)

Figure 25. Relationship between Dose Rate andStorage Amount of Waste (Example)©International Research Institute for Nuclear DecommissioningNo.39(II) Accuracy Improvement of Analytical Evaluation Methods

a. Study of accuracy improvement methods, b. Study of data collection methods to improveaccuracy, c. Establishment of analytical evaluation methods
FY

Implementation plan

Goal achievement index

2017
2018

i) Study of accuracy improvement methods• Methods necessary to improve the accuracy of inventory evaluation will be studied basedon the results described in Section (I), including the classification of waste (composition ofradionuclides that characterize the waste), the contamination mechanism of radionuclides,
and the study results on the representativeness of analysis data.
ii) Data collection for accuracy improvement• Necessary data will be collected based on the knowledge accumulated through studies
and the knowledge obtained through the activities listed in Section (I).
iii) Establishment of analytical evaluation methods• An environment for the analytical evaluation will be established which incorporates
methods required for the improvement of inventory evaluation accuracy (including
calculation tools).

i) Study of accuracy improvement methods• Indicate the measures for improving
analytical evaluation accuracy.

Implementation details

LegendBlue: Plan
Red: Resultii) Data collection for accuracy improvement• Collection of data that contributes toaccuracy improvement.
iii) Establishment of analytical evaluation
methods• Preparation of analytical evaluation tools.

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(II) Accuracy Improvement of Analytical Evaluation Methods- Study of accuracy improvement methods andestablishment of analytical evaluation methods

Planning

Accuracy improvement study andthe establishment of methods

Summary- Collection of data for accuracy improvement

Planning

Collection of basic experimental
data

Summary©International Research Institute for Nuclear Decommissioning3

No.40(II) Accuracy Improvement of Analytical Evaluation Methods
 While the classification of waste is determined for the handling of them in storage, treatment, and disposal separately, the safetyof them in disposal is closely related to the composition of nuclides. It was thought to be necessary to assess disposal safety
and compare the result with a concentration equivalent to the standard dose rate for classification. (Figure 26)
 Although the contamination level of waste from dismantlement is not known, the setting of the parameters of nuclide migration to
the waste was updated, considering the classification of waste generated by ordinary nuclear reactor decommissioning,
contamination via air, and the process of secondary contamination caused by the contact with stagnant water and theadherence of fuel debris. (Figure 27)
物
炉内構造
Reactorinternals

炉内燃料
In-corefuel

Emission
to air and sea
大気・海洋放出Inside
R/B
建屋内

Soil
土壌
In-building
airborne emission
建屋内気中放出Activity concentration of waste

Disposal Category A(deeper undergrounddisposal is required)

燃料デブリ
Fuel debris

Activity concentrationequivalent to thestandard dose rate

Waste 6

Waste 5

Waste 4

Waste 3

Waste 2

Waste 1

Disposal Category B(shallow undergrounddisposal isacceptable)

Rubble(outdoor))
瓦 礫(建屋外
Felledtrees
伐採木

Stagnant
滞留water
水

汚染水処理Contaminatedwater treatment

吸着塔tower
Adsorption
Decontamination
除染装置
systemMulti-nuclide
removal
多核種除去設備
system

処理 済水Treatedwater

Figure 26. Comparison of the activity concentration of waste with aconcentration equivalent to the standard dose rate of disposal
safety for the classification of waste (concept illustration)

RPV 物
waste[L1][事故前: L1相当]
圧力容器廃棄(燃料デブリ付着)
留水による汚染(Fueldebris ++滞
stagnant
water)

RPV 物
waste[L2][事故前: L2相当]
圧力容器廃棄(燃料デブリ付着)
留水による汚染(Fueldebris ++滞
stagnant
water)

RPV物waste[L1][事故前: L1相当]
圧力容器廃棄(滞留水による汚染)(Stagnantwater)[事故前: L2相当]
圧力容器廃棄RPV 物
waste[L2](滞留水による汚染)(Stagnantwater)

Waste from
解体廃棄物
dismantlement) [事故前: L2相当]
格納容器内廃棄PCV waste(metal)[L2]物(金属(燃料デブリ付着)
滞留水による汚染(Fueldebris ++stagnant
water)) [事故前: L2相当]
格納容器内廃棄PCV waste(concrete)[L2]
物 (コンクリート(燃料デブリ付着)
留水による汚染(Fueldebris ++滞
stagnant
water)

PCV waste(metal)[L2]) [事故前: L2相当]
格納容器内廃棄物(金属(燃料デブリ付着
滞留水による汚染)(Fueldebris ++stagnant
water)

PCV waste(concrete)[L2]) [事故前: L2相当]
格納容器内廃棄物 (コンクリート
留水による汚染
(燃料デブリ付着)(Fueldebris ++滞
stagnant
water)

PCV waste(metal)[L3]) [事故前: L3相当]
格納容器内廃棄物(金属(滞留水による汚染)(Stagnantwater)

PCV waste(concrete)[L3]) [事故前: L3相当]
格納容器内廃棄物 (コンクリート(滞留水による汚染)(Stagnantwater)

PCV waste(metal)[L3]) [事故前: L3相当]
格納容器内廃棄物(金属(滞留水による汚染)(Stagnantwater)

PCV waste(concrete)[L3]) [事故前: L3相当]
格納容器内廃棄物 (コンクリート(滞留水による汚染)(Stagnantwater)

R/B waste
建屋内廃棄) [事故前: L3相当]
物(金属(metal)[L3](滞留水による汚染)(Stagnantwater)) [事故前: L3相当]
物(金属(metal)[L3]
建屋内廃棄
R/B waste

R/B
waste 物
(metal)[non-radioactive]
建屋内廃棄(金属) [事故前: 非放射性](滞留水による汚染)(Stagnantwater)(金属) [事故前: 非放射性]
建屋内廃棄
R/B
waste 物
(metal)[non-radioactive]

R/B 物
waste
(concrete)[L3](コンクリート) [事故前: L3相当]
建屋内廃棄(滞留水による汚染)(Stagnantwater)(コンクリート) [事故前: L3相当]
建屋内廃棄
R/B 物
waste
(concrete)[L3]

R/B
waste物(concrete)[non-radioactive](コンクリート) [事故前: 非放射性]
建屋内廃棄(滞留水による汚染)(Stagnantwater)
建屋内廃棄
コンクリート) [事故前: 非放射性]物((metal)[non-radioactive]
R/B waste

Figure 27. Contamination routes and classifications for theestimation of the inventory of waste from dismantlement©International Research Institute for Nuclear Decommissioning

51

No.41(III) Measures for Materials with Impact on Disposal
Implementation plan and goal achievement index
FY

Implementation plan

Goal achievement index

2017• Case studies on acceptance criteria of waste inside and outside Japan will be surveyed,regarding not only radionuclides but also materials that influence treatment and disposal
methods.

• Presentation of cases that address theacceptance standard of waste in terms ofthe content of influential materials to
disposal concepts.
• Proposal of quantitative evaluationindexes for the evaluation of barrier
performance at the time of disposal andfor the impact on nuclide migration.

2018• Case studies on acceptance criteria of waste inside and outside Japan will be surveyed.
• Based on the survey result, concepts of the acceptable concentration and content inmaterials with impact on disposal will be examined, which have risks of being contaminatedin predisposal management and disposal facilities.
• Methods of the quantitative evaluation will be studied for the impact of components that are
known to have an impact on barrier performance (structural and nuclide migration-related
chemical properties transformation) at the time of disposal and nuclide migration (nuclide
migration-related chemical properties transformation).

• Indication of concepts about theacceptable concentration of materials
exerting influence on disposal strategy
and content of such materials in waste in
Fukushima Daiichi Nuclear Power
Station.
• Proposal of quantitative evaluationindexes for the evaluation of barrier
performance at the time of disposal andfor the impact on nuclide migration.

Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

I. Characterization(III) Measures for Materials with Impact on Disposal- Case study inside and outside Japan

Planning- Investigation of impact on disposal safety

Planning

Research on cases

Study of indexes that can express impacts on
performance and quantitative evaluation methods

Summary

Summary©International Research Institute for Nuclear Decommissioning3

(III) Measures for Materials with Impact on Disposal




No.42

The waste generated by the accidents may contain materials that are supposed to be separated and removed before disposal because of
the restriction on the discharge of such materials into the environment. The same is true for materials with a potential adverse impact on thesafety of treatment and disposal. Measures taken in Japan, the UK and the USA for dealing with such problems were surveyed. (Table 10)
Typical materials with a potential adverse impact on the disposal and the solidification of waste were identified and listed. (Table 11)When examining the impact of a material on disposal, attention must be paid on the direct impact of the material on the migration of
nuclides (deterioration of sorption performance due to complex formation, etc.) and indirect impact to cause change in the characteristicsof the medium.

It is necessary to examine whether there’s anything to be reflected to the decommissioning process in parallel to the study on the impacts ofsuch materials and their permissible concentrations.

Table10. Example of Waste Acceptance Standards in the UK (LLWR*1)
Category

Category 1

Category 2

Category 3

Definition

Typical materials

Requirement of control only
without specific restrictions or sorting
requirements
The amount of the materials concerned
needs to be recorded in the inventory.

Halogenatedplastics, asphalt,
copper, stainlesssteel, etc.

Sorting required.
Waste with the content of the materials
concerned equal to or lower thancriteria for acceptance can be
accepted. The amount of the materials
needs to be recorded in the inventory.
Materials classified as a hazardous
material.
Special management is required forthese materials until disposal.

Table11. List of materials with a potential adverse impact on thedisposal and the solidification of waste
Classification

Arsenic, lead,
mercury, electronicparts, etc.

Asbestos, oil,solution, etc.

Environmental
toxinsMaterials with
impact on
disposal
Materials with
impact on
solidification ofwaste

Definition
Materials with a potential
adverse impact on theenvironment
Materials with a potential
adverse impact on nuclide
confinement performance
Materials that generate gases
by corrosion during solidification

Typical materials
Lead (shielding
material), boron
Organic material,boron, oxidizing
material(nitric acid, etc.)
Aluminum, zinc(staging material,etc.)*1. LLWR: a low-level radioactive waste disposal facility in the UK
operated by Low Level Waste Repository Ltd.

©International Research Institute for Nuclear DecommissioningNo.43

3. Study on Predisposal Management©International Research Institute for Nuclear DecommissioningNo.44

Contents of Report(I) Applicability assessment based on characteristics of solid waste(II) Study and Evaluation of Waste Storage and Management Methods
Appropriate for the Characteristics of Solid Waste
a. Study of safety measures required for the storage of highly radioactivewaste(a) Measures for hydrogen gas generation(b) Measures for projected wastes generated by fuel debris retrieval
b. Evaluation of technology to stabilize the secondary wastes generated fromcontaminated water treatment(a) Applicability assessment of in-drum glass solidification technology(b) How to stabilize sludge from decontamination systems(III)Research on Technologies for Reducing Waste Production©International Research Institute for Nuclear DecommissioningNo.45(I) Applicability Assessment Based on Characteristics of Solid Waste
FY

Implementation plan

Goal achievement index

2017• Characteristics wastes generated by the accident will be extracted for solidification technologies that have
proven in the treatment of radioactive waste (such as cement solidification technology and high-temperature
treatment technology). The applicability of the solidification technologies will be evaluated in consideration of the
impact of dose rate and heat generation to identify issues to be addressed continuously.

• Clarification of issues incase of the application of
conventional technologies
to highly radioactive waste.

2018• New solidification technologies will be investigated to evaluate the possibility of the solution for the identified
issues. Expected effects by conducting tests will be evaluated and test the actual applicability when necessary.

• Proposal of a solidification
technology that can solve
the identified issues.

Progress status
The applicability of solidification technologies to waste that possess
unique characteristics attributable to the accident was examined bycomparing those wastes with the existing classification of radioactive The figure below illustrates theclassification of six
waste.
Waste
Stabilization
technologies with respect to Six technologies that were considered in the fundamental tests
conditioning
storage,performed until the end of FY2016 were grouped into the following
A possibility ofFunctionalization
Application of dewatering
two groups: technologies that have proven track records inreworking at thetechnology Simple
process and effective instage of the pending
pretreatment before disposal (waste conditioning) and technologies
dispersion prevention,
disposal strategy.
thus useful.
that can improve safety in storage (stabilization).
 From the results of the applicability assessment of the conventional
Well-established
Pellet solidification
Sintering solidification
technologies to highly radioactive waste, the following issues were
waste conditioning
A room-temperature setting material with low
identified: the understanding of the relationship between thetechnology for lowmoisture content. Application to the storage of highly
Cementand mid-level waste
radioactive waste and use as an emergency repair
solidification
radioactivity amount of nuclides contained in the waste and decaymaterial are expected. It can be used for the material
heat that is generated in the waste and makes an impact on theof waste conditioning because it has a glass-likealuminum-silicon network framework. As it is amaterial properties of solidified bodies, the development andrelatively new material, additional properties data
adoption of an assessment index for the comparative assessment ofneeds to be obtained.
AAM solidification(geopolymer)
applicable waste treatment technologies, and the collection and
The performance of water removal, organic content removal, andarrangement of data (data range) related to the developed index.
volume reduction, as well as the applicability to a wide range of
waste make this technology applicable to various processes from The necessity of considering approaches for refinement technologies
storage to waste conditioning.
It can be an option for technology used to process ordinary waste
applicable to pretreatment was also presented from the identified
generated every day.
This technology requires a melting furnace and other facilities andissues.


Future plan
 Develop an investigation and test plan toward solving identified
issues, propose and implement it.

equipment along with labor.
Property control tests are required for stable operation.
The composition and leaching characteristics of molten materials
need to be examined.

Glasssolidification

Melt solidification

Figure 1. Classification of Six Technologies Subjected to
Fundamental Tests Performed by FY2016©International Research Institute for Nuclear DecommissioningNo.46(II) Study and Evaluation of Waste Storage and Management Methods Appropriate
for the Characteristics of Solid Waste
a. Study of safety measures required for the storage of highly radioactive waste(a) Measures for hydrogen gas generation
FY

Implementation plan

which Goal achievement index

2017• Expertise from overseas (the UK, the USA and France) and Japan will be surveyed about
the hydrogen gas generation method during the storage of solid waste containing highradiation and water, and requirements for the storage container such as vents that hydrogengas may generate..

• Presentation of knowledge concerninghydrogen gas generation assessment methods
and requirements for the storage container suchas vents

2018• Based on the results of research on hydrogen gas generation conducted in FY 2017 whichwere collected from overseas and Japan, the applicability to 1F will be studied to clarify
potential issues.

• Presentation of the 1F site applicability andissues

The survey on concepts of generated hydrogen gas, hydrogen gas generation assessment
methods, container specifications, and measures for hydrogen gas generation will be
conducted, focusing on differences from Japan and including reasons for differences.

[Survey items]
1.
Survey on concepts of hydrogen gas generation in eachcountry(regulations and technical requirements)

UK

Concepts of hydrogen gas generation in eachcountry(regulations and technical requirements) were surveyed
in storage, treatment, disposal and transport phases.

France

2.

Survey of evaluation methods for hydrogen gas
generation


3.

4.

Hull and end piece wastes (France)
Ferric hydroxide floc (UK)
Barium carbonate slurry (UK)
Fuel cladding and activated metals(UK)
NPS KURION adsorption vessel on
1F (Japan)
SARRY adsorption vessel on 1F (Japan)
Sludge from AREVA decontamination
system on 1F (Japan)
Iron coprecipitation slurry on 1F (Japan)

USA

Waste on
1F

Carbonate coprecipitation slurry on
1F (Japan)

Types and materials of containers were broadly investigated,
mainly measures for hydrogen gas generation.

Measures for hydrogen gas generation


TRU waste (USA)G-value settings and concepts were mainly investigated.

Investigation of required functions for a solid waste
container

Solid wastes generated from the 1F site are requiredhydrogen gas management. Wastes were selected frominventory in each country which are similar to the wastes in
1F site.

Measures for reducing hydrogen gas generation were
investigated.

Rubble in R/B of Units 1 to 3

Inventory of waste (Bq/waste 1kg)
Source of 1F waste inventory data:
Appendix 4.4.3-1 titled "Inventory and nuclide migration parameters," of the research report (interim report) for “R&D for Treatment and Disposalof Solid Radioactive Waste” under Subsidy Project of Decommissioning and Contaminated Water Management in the FY2016 Supplementary
Budgets
Note that data of rubble in R/Bs of Unit 1 to Unit 3 are cited from the International Research Institute for Nuclear Decommissioning web site
(2018).

©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.47
Appropriate for the Characteristics of Solid Waste
a. Study of safety measures required for storage of highly radioactive waste(a) Measures for hydrogen gas generation - Result overview
By the comparative investigation on cases in Japan and other countries, useful information on measures for hydrogen gas
was obtained concerning:(1) Reduction in free water volume, (2) Use of adequate G-values depending on materials and the type of rays, (3)Ventilated container.
Japan
Waste to beinvestigatedAppearance ofcontainer(example)

USA

Waste generated by power stations
and reprocessing facilities, which canbe accompanied by hydrogen gas
generation• Activated metals• L1 waste, such as polymers• Waste from reprocessing facilities
Waste from active facilities
Lid

UK

FranceILW waste which hydrogen gas
generation needs to be considered• Ferric hydroxide, barium carbonate• Oxide fuel cladding• Plutonium-contaminated material• Waste from Magnox fuelreprocessing• Mixed legacy waste

Waste with a risk of hydrogen gas
generation and/or waste similar to thaton 1F
• Hull and end piece wastes (CSD-C)• Dry sludge from liquid waste
treatment facilities (DSC)

Solid waste with a moisture content of 1
vol% or lessBarium carbonate slurry waste

CSD-C: Hull and end piece wastes• Radiolytic water splitting
G-value (H 2):
α-ray: 1.6
βγ-ray: 0.5• G-value for α-ray is determined
depending on the characteristics ofwaste.

• Radiolytic water splitting
G-value (H 2):βγ-ray: 0.4• G-value of cement solidified material
G-value (H2):βγ-ray: 0.15

• Radiolytic water splitting
G-value (H2):α-ray: 0.044*
γ-ray: 0.0021*• Storing gas in a ventilated container(Stored in a sealed container duringtransport)• Moisture content ≤ 1%

• Storing gas in a ventilated container(Stored in a sealed container duringtransport)• Storing gas in a ventilated container(Stored in a sealed container duringtransport)• Maintaining a hydrogen gas
concentration of 1 vol% or less byventilation

Waste stored in disposal facility
considering experience of waste
acceptance• TRU waste(WIPP disposal facility)• Class C waste(WCS disposal facility)

Weld

Bolt ring
Additional shield
Filter

Waste

Ring belt

Bolt

Screw
Top lid, gasket
Top plate, gasket
HandleWaste
Filter port

Main
body
bottomplate
Guide handle

Measures forhydrogen gas
generation

Offset tube

Main body

Activated metal waste

Hydrogen gas
generationassessment

Main body• Radiolytic water splitting
G-value (H2):
γ-ray: 0.45

• Storing in a sealed container(Gas leakage prevention from thecontainer)• Dewatering and drying*A value of the crystallization water of MoZr precipitate• Corrosive gas generation wasassessed.

©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods
Appropriate for the Characteristics of Solid Waste
a. Study of safety measures required for storage of highly radioactive waste(a) Measures for hydrogen gas generation - Result summary

No.48

Measures for hydrogen gas generation obtained from investigations in this fiscal year are summarized below
from the aspect of reduction in free water volume, use of adequate G-values depending on materials and thetype of rays, and storage using containers with vents.
Operators in all three countries determined the reasonable methods of treatment, storage, and disposal, andcoordinated the methods to be approved by the regulatory authorities.

(1) Reduction in free water volume• All three countries consider the reduction of free water volume by the drying or cement-solidification of waste
because the decomposition of free water is basically dominant to estimate volume of hydrogen gas
generation.
• The handling of free water and other water (ex. cement crystallization water) varies by country.
Example)USA and UK:Volume of hydrogen gas generation is estimated depending on the volume decrease or state change of free waterby cement solidification.
France:G-value of crystallization water in MoZr precipitate was set to estimate volume of hydrogen gas generation.

(2) Use of adequate G-values depending on materials and the type of rays
• All three countries use different G-values in hydrogen gas generation volume estimation depending on thecharacteristics of waste (such as water state, the type of organic materials, and the type of radioactive rays).

(3) Storage in containers with vents• All waste investigated in this research (ex. TRU waste in USA, ferric hydroxide waste in UK, and CSD-C
waste in France) were stored in containers with vents, which were implemented hydrogen gas measures.
• Beside hydrogen gas, public radiation exposure from fission products (H-3, Kr-85, etc.)was also considered.

©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.49
Appropriate for the Characteristics of Solid Waste
a. Study of measures for storing highly radioactive waste(b) Measures for wastes generated by fuel debris retrieval (1/3)
FY

Implementation plan

Goal achievement index

2017• The latest information on wastes generated by fuel debris retrieval work will be collected
and organized the collected information based on study results from other projects
conducted for fuel debris retrieval (such as the project of fuel debris retrieval, andcontainment, transport and storage of fuel debris). The proposed waste storage andmanagement methods will be studied based on collected and organized information.

• Presentation of the latest information of projected
wastes generated by fuel debris retrieval work
based on study results from other projects
conducted in relation to debris removal (such as
the fuel debris retrieval project and thecontainment, transport and storage project).

2018• The latest information will be collected and organized the collected information incooperation with other projects conducted for fuel debris retrieval (such as the project offuel debris retrieval and collecting, transferring and storing of fuel debris). All ideas ofreasonable methods to store and manage wastes generated by fuel debris retrieval work
will be reviewed and proposed recommendable methods in consideration of fuel debris
retrieval process as well as the collected and organized information.

• Examine all ideas of reasonable methods to store
and manage wastes generated by fuel debris
retrieval work and propose recommendable
methods among the ideas in consideration of fueldebris retrieval process, as well as the collected
and organized information.

 Progress status
 Types and volume of wastes generated by fuel debris retrieval were
estimated and the first version list (draft) of said waste was createdbased on the latest investigation results provided by the fuel debris
retrieval project and the containment, transport and storage project.
(Figure 2)
 A list of the safety function requirements that will be required in eachstep for the storage and management of wastes generated by fueldebris retrieval was created. Several feasible process flows ofstorage and management (including a storage container case withfiltered vent case) were developed.
 Future plan All reasonable and feasible methods to store and manage wastes
generated by fuel debris retrieval work will be summarized andproposed recommendable methods among the methods.

Type

Weight (t)Waste density(t/m3)DSP slot plug

Concrete

Unknown

2.5(>100)

Heat-retention
material

Metal

Unknown

7.8

–

Waste

Generated
as waste

Generated
as wastecontaining
fuel debris

Volume (m3)Dose rate (mSv/h)
4.0E+3–

PCV head

Metal

50

7.8

6.4

4.0E+3

Shield plug

Concrete

600

2.5

240.0

4.0E+33.0E+4

RPV head

Metal

70

7.8

9.0

Dryer

Metal

28.0

7.8

3.6

4.0E+5

Separator

Metal

17.2

7.8

2.2

2.0E+5Top guide

Metal

3.4

7.8

0.4

1.0E+6

Shroud

Metal

25.1

7.8

3.2

3.5E+5

Core plate

Metal

4.0

7.8

0.5

3.5E+5
3.5E+5

CR guide tube

Metal

11.2

7.8

1.4

CRD

Metal

28.9

7.8

3.7

3.5E+5

RPV bottom mirror

Metal

50.0

7.8

6.4

3.5E+5

–

887.8

Total

–

276.9

–

Figure 2. First version list (draft) of projected wastes generated byfuel debris retrieval (at Unit 1)©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.50
Appropriate for the Characteristics of Solid Waste
a. Study of measures for storing highly radioactive waste(b) Measures for wastes generated by fuel debris retrieval (2/3)"List of safety function requirements" List of potential risks and safety function requirements during transport (partial listing for transport only)
Potential serious accidents that could occur in each step of
the handling of projected wastes generated by fuel debris
retrieval, such as containment, transport, and storage, wereconsidered along with risks posed in association with theaccidents, as well as required safety functions to prevent
accidents and risks.
Assumed serious
accidents

Required safetyfunctions

Spread ofcontamination

Confinement
Corrosion
resistanceInternal
exposure

Duringtransport

Seriousaccidents

Hypothetical accident
scenarios

Spread ofcontamination• A transport cask falls and isdamaged during transport,
and the content spills out of
the cask.

High• The impact becomes serious
when contamination spreads
in the premise.

• Make the transport cask strong enough towithstand the impact of a fall and devise
a secure sealing mechanism to prevent
spillage or leakage even in case of a fall.

Internal
exposure• α nuclides leak out of acontainer and a worker is
exposed to internalradiation by inhaling them.

High• There is a high risk of internal
exposer if dusts that contains
α nuclides leaks• Make the transport cask strong anddevise a high quality sealing mechanism.

External
exposure• A worker near the transport
cask is exposed to radiationfrom the cask directly.

High• There is a high risk ofexternal exposure unless thetransport cask possess
sufficient shielding
performance.

• Make the shielding of the transport cask
thicker to improve its shielding
performance.

Fire• The inside of the transport
container is filled withhydrogen gas and catches
fire.

High

• There is a risk high of theoccurrence of hydrogenconcentration when moistureand α nuclides coexist inside
the sealed container.

• Exercise time control to maintainhydrogen concentration below lower
explosion limit.

Work accident• A worker is injured by afalling transport cask.

Low• There is a similar risk of work
accident to the transport ofordinary goods.

• Use fall prevention measures fortransport casks.

Structural
strength

External
exposure

ShieldFire

Hydrogen gas
control
Operability

Work accident

Primary factor:
Secondaryfactor:Clarification of safety function requirements inindividual work steps
 Each of the confinement, transport, andstorage steps

Risk

The most important challenge is to satisfy both of
the following requirements.
• Ensure confinement performance• Ensure measures for hydrogen gas

Measures according to required
safety functions

Conflicting safety
requirements in case ofhydrogen gas release

Development of a process flow that satisfied the above
conflicting safety requirements and also meets all other safetyfunction requirements.
©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods
Appropriate for the Characteristics of Solid Waste

No.51

a. Study of measures for storing highly radioactive waste(b) Measures for wastes generated by fuel debris retrieval (3/3)
Allocation of safety functions that satisfies the prevention of nuclide dispersion (confinement) andmeasures for hydrogen gas (idea)
Roles of Safety functions for hydrogen gas confinement and reduction(idea)
Formulation of ideas forreasonable storage andmanagement process flow
• Use of transport casks• Use of containers with filter• Use of cask transporters
with sealing function• Adoption of dispersion
prevention methods• Drying pretreatment:::

The ideas to be narrowed
down in FY2018

Roles of safety functions
Pretreatment

N/A

During transport

Storage

Confinement

Hydrogen gas control

Confinement

Hydrogen gas control

Sealed transport cask

Time management

Storage building

Storage building

Time management

Storage building

Storage building

Tunnel ventilation

Storage building

Storage building

Intra-premise cask transporter
with sealing function
Simple structure tunnel
Sealing bag

Time management

Sealed storage container ventilated with high-performance filter

Drying treatment

Sealed transport cask
Intra-premise cask transporter
with sealing function
Simple structure tunnel

Sealing bag
Periodic implementation
Sealed storage container ventilated with high-performance
filter+ Storage building
Storage building

Not required(due to drying treatment)

Storage building
Storage building

Sealing bag

Sealing bag

Sealed storage container

Sealed storage container

Anti-scattering agent
application/facing bag

Unnecessary (due toapplication of anti-scattering
agent/packaging in facing bag)

Ventilated storage container

Drying treatment + Antiscattering agent
application/facing bag

Unnecessary (due toapplication of anti-scattering
agent/packaging in facing bag)Not required(due to drying treatment)

Not required(due to drying treatment)

Unnecessary (due to
Ventilated storage container +application of anti-scattering
Storage building
agent/packaging in facingbag)
Unnecessary (due toapplication of anti-scattering
agent/packaging in facingbag)

Not required(due to drying treatment)

11 storage and management flows were designed to evaluate the safety functions
in each step.
©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.52
Appropriate for the Characteristics of Solid Waste
b. Evaluation of stabilization technology for the secondary wastes generated from contaminated water treatment(a) Applicability assessment of in-drum glass solidification technology
FY

Implementation plan

Goal achievement index

2017• With the aim of clarifying glass solidification conditions, glass composition will be
studied and crucible melting tests will be performed to obtain the data, such astreatment conditions, volume reduction ratio, and the physical properties of glass.

• Presentation of glass solidification conditions by studying
glass composition and performing crucible melting tests.

2018• Engineering-scale tests using the glass solidification conditions derived from thetest results in FY2017 will be performed to investigate conditions for glass melting
by joule heating, and to evaluate the impact of nuclides on the characteristics ofsolidified materials and off-gas systems. Additionally, the durability andmaintenability of device will be confirmed.

• Presentation of the results of study on the impact of
nuclides on the characteristics of solidified materials andoff-gas systems and the results of the assessment on the
durability and maintainability of equipment by performing
engineering-scale joule heated glass melting tests.

 Purpose In-drum glass solidification technology (GeoMelt ICV) is focused on as a
candidate stabilization technology for the pretreatment of secondary
wastes generated from contaminated water treatment facilities from the
aspect of the prevention of contamination spread and equipment size.
The applicability of treatment which zeolite is used as the main
component of the glass solidification agent to simultaneously melt slurries Is studied and evaluated.
 Progress status Basic tests (glass composition examination and crucible melting test)• Crucible melting test (Figure 3)
 Simulated waste: 20 types selected from the waste generated by
the water processing facilities of Fukushima Daiichi NPS (Table 1)
 Composition: Zeolite (main component), simulated waste,
additives, and tracer were mixed with compositions determinedbased on analysis results (Total 40 kinds of mixing ratios).
 Melting procedure: Remove volatile materials (such as water) at(700°C) first, heat the mixture to1250 C while confirming the
progress of melting, and anneal it at (500C).

Table 1 Simulated waste and tracer
Water treatment facility

Simulated waste*1)

TracerDecontamination system(AREVA)

Sludge (1 type)Sr, CsCesium adsorption apparatus(KURION)

Filter material (1 type)
Adsorbent (4 types) *2)

Sr, CsSecondary cesium adsorption Filter material (2 types)*3)
apparatus (SARRY)
Adsorbent (1 type) *4)
Multi-nuclide removal system

Sr, CsSlurry (2 types)

Sr, CsAdsorbent (9 types)

Depending on adsorbent*5)*1) Additives (such as B2O3, CaO, Na2O) are mixed in simulated waste*2) Types of absorbent include zeolite (two types), silver zeolite (one type), and silica
titanate (one type)*3) One of 2 types of filter materials contains zeolite*4) Zeolite is used as adsorbent*5) Sr, Cs, Co or Sb is used depending on the elements the adsorbent tries to remove.
Zeolite, sludge, and additive(such as B2O3) aremixed

Figure 3. Example of Glass Melted
in Crucible©International Research Institute for Nuclear Decommissioning

No.53• Results of glass observation and analysis

 Fluidity
The fluidity of the mixture was visually observed at 1,150°C or higher temperature in all compositions.
 Volume reduction ratio, waste filling rate
Volume reduction ratio (percentage of the volume of glass to the volume of simulated waste) is between 14% and 26%. Waste filling rate is
between 70% and 80%.
 Visual observation
The glass solidification was confirmed on both zeolite and other waste. Precipitates were observed in some solidified glass samples.
 Analysis
Solidified glass samples with precipitates were sent to elemental analysis (7 samples to EDS) and crystallographic analysis (7 samples to
XRD) to identify precipitated rutile from silicon titanate. As a result, it was found that homogeneous glass can be produced from the mixture ofcarbonate precipitates when the mixing ratio is between 16 to 21 wt%, and also from the mixture of iron coprecipitation with a mixing ratio of
11 to 12 wt%, both of which are kept a lot on 1F. In addition, the successful glass solidification of the mixture of carbonate precipitates(approx. 16 wt%) and iron coprecipitation (approx. 2 wt%) was confirmed when approx. 65 wt% zeolite was used.
As to titanate, which is also kept a lot, the production of homogeneous glass from its mixture was confirmed when its mixing ratio was
between 8 wt% and 18 wt%.
Sludge may produce precipitates during glass solidification. However, glass without the layer of sulfate was produced from the mixture ofsludge with a mixing ratio of 3-6 wet%.
 Leach test (Figure 4)
Simulated waste samples were grouped into 7 groups and15 types of mixtures each of which contains a simulated
waste picked up from the 7 groups, with 1 to 3 different
mixing ratios prepared. Glass samples were produced fromthose mixtures and subjected to leach tests (MCC-1, 28days). As a result of the leach tests, the leaching amount
NL(B) from the glass sample was found to be nearly equal tothat of the P0798 glass developed by former Power Reactor
and Nuclear Fuel Development Corporation (data is shownin Figure 4, indicated by ←) for all glass samples, and thus
the high chemical durability of all glass was proved.
 Tracer retention rate
The tracer retention rate in the glass samples was asfollows regardless of the type and mixing ratio of simulated
waste samples: Sr retention rate, 72 to 128%*; and Cs
retention rate, 59 to 95%. Cs shows small volatilization
suppressing effect in crucible melting.
* The analysis result of retention rate exceeded 100% due to analysis error.

Glass produced by
crucible melting tests

International
standard glass

ZeoliteTitanate-containing

Carbonate slurry

Iron slurry

Sludge←Carbon-containing

OthersDue date

Figure 4. Comparison of leaching amount with international
standard glass©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.54
Appropriate for the Characteristics of Solid Waste
b. Evaluation of stabilization technology for the secondary wastes generated from contaminatedwater treatment(b) Study on sludge stabilization of decontamination systems (1/2)
FY

Implementation plan

Goal achievement index

2017• Fluidity evaluation tests using simulated sludge samples will be performed to obtain basic dataof fluidity properties for sludge concentration as a test parameter. In addition, a demonstration
testing plan will be developed for FY2018.

• Collection of the basic data of fluidity andrelated properties and the formulation ofa demonstration testing plan.

2018• Sludge discharge methods will be verified by performing engineering-scale sludge collection,cleaning, and transferring tests that use simulated sludge samples and the simulated D-pit
internal structure.

• Presentation of the feasibility of a sludge
discharge method from the D-pit.

Implementation details
Simulated sludge samples were produced according to the actual
process and systems. A test plan was developed for the collection ofbasic data.
 3 types of simulated sludge samples were produced considering
time deterioration of actual sludge (especially polymer sludge).
• Sludge A: containing polymers• Sludge B: no polymer• Sludge C: irradiated sludge A
 Collection of basic fluidity data using simulated sludge samples
 Analysis results of actual sludge were compared and a verification
testing plan was developed.
 Outcomes A method for determining sludge discharge technology wasestablished by producing simulated sludge considering time
deterioration of sludge under actual conditions and understanding of
fluidity properties necessary to design sludge collection and transfer
methods in the combination of the results of the investigation of
inside the D-pit.


Test sample and concentration
(g/L)

Item

(1)

Density(2)

Particle size
distribution

Test items (analysis items)

60

Wet density of sludge

60

Particle size distribution ofsludgeFe, Ni, Ba, Na, Si(3)

Chemicalcomposition

SludgeSludge A
Sludge B
Sludge C

Cl
Total C, Total S

60

Fe, Ni, Ba, Na, Si
Clear
Free CN
supernatant
Cl, SO4
water
TOC

(4)

Sludge A
Sedimentatio
Sludge B
n property
Sludge C(5)

Viscosity(6)

Effectivenes
s of stirring

60

Sludge A

75 to 300

Sludge B

2 concentrations

Sludge C

1 concentration

2 types60

Sedimentation behavior inwater

Viscosity variation with sludge
concentration
Effectiveness of bubblingstirring©International Research Institute for Nuclear Decommissioning

(II) Study and Evaluation of Waste Storage and Management Methods No.55
Appropriate for the Characteristics of Solid Waste
b. Evaluation of stabilization technology for the secondary wastes generated from contaminated water treatment(b) Study on sludge stabilization of decontamination systems (2/2)
Physical properties of simulated sludge (density, particle size, and chemical composition: Figure 5)

If all Ba atoms are assumed to exist in a form of barium sulfate, about 66% of sludge in weight is barium sulfate.
The particle size of sludge A is about 20 times larger than that of sludge B due to the aggregating effect of polymers.
Wet density(g/cm3)

Average particle size
(volume basis)
(μm)

C

Na

Si

Chemical composition of sludge (g/kg)
S

K

Fe

Ni

Ba
391

Sludge A

1.15

120

36.4

1.41

6.24

105

22.3

48.8

32.2

Sludge B

1.14

6.45

33.9

1.03

4.40

107

19.4

44.0

28.4

400

Sludge C

1.09

10.8

36.8

1.35

3.63

98

22.6

56.6

30.4

396

Actual sludge

1.18

8.89

－

20.5

13.2

108

16.2

107

32.6

461



Sedimentation property (Figure 6)



Sample sludge was stirred in a measuring cylinder and left standing to observe theoccurrence of sedimentation (change in the height of sludge phase boundary).
Actual sludge showed sedimentation behavior that is like the even mixture of those ofsludge A and B. Sludge C (irradiated sample) showed a similar behavior to that of actual
sludge.



Viscosity (Figure 7)



The viscosity of the sludge was measured with different concentrations to investigate theinfluence of the concentration.
The viscosity of the sludge decreased with an increase in the shear velocity in a highconcentration range on both sludge samples A and B. Meanwhile, the sludge samples
showed a near-constant viscosity in a low concentration range regardless of the shear
velocity.
■Sludge A◆Sludge B▲Sludge C[D] Actual
sludgeSludge A

Sludge B

Figure 6. Sedimentation Property Test Result

Sludge C

Effectiveness of stirring (Figure 8)

Sludge was put in a glass cylinder (Φ40 mm×600 mm) and left
standing for a week to cause sedimentation. Then air was injectedfrom the bottom for stirring and the concentration of the sludge wasmeasured at three different heights.
The measured sludge concentration was nearly equal to the initialvalue (60 g/L), which suggested that the homogeneous re-stirringand mixing of the sludge could be done.



600 mmSampling◆■217 g/L◆■82 g/L◆■253 g/L◆■68 g/L

Sampling Sludge A

0 mm

Sampling

Air

Time (min)Sludge B

Figure 5. Chemical Composition of
Simulated Sludge Samples♢ Data during shear velocity increase□ Data during shear velocity decrease

Viscosity (mPa･s)

Ratio of sludge phase boundary
height to initial liquid surface height

Sludge A

Element concentration (g/kg)

Sludge B

Sampling height(mm)Concentration
(g/L)

600

55

300

56

0

58

600

48

300

49

0

50

Figure 8. Stirring Verification Test result
Shear velocity (1/s)

Figure 7. Result of viscosity measurement

Sludge A contains polymers, Sludge B contains no polymer and Sludge C is an irradiated sample©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production
FY
2017

Implementation plan

No.56

Goal achievement index• Technologies applicable to the measurement and assessment of α-contamination in highly radioactive
environments will be investigated and studied with the aim of applying them to waste sorting methods thatcan reduce waste amount and help realize waste categorization.

• Indicate technologies
applicable to 1F.

 Progress status
 The purpose of measurement, waste to be measured, and environmental conditions for measurement were studied, all of which wereoriginally planned to be conducted in FY2018 but decided to do in this fiscal year by moving up the schedule by half a year.
 Future plan
 The purpose of measurement, waste to be measured, and environmental conditions for measurement will be studied in FY2017.
Technologies applicable to the measurement and assessment of α-contamination will be investigated and studied by measuring
principle, and promising technologies will be proposed in FY2018.
Original and Revised Schedule
Process

1Q 2017

2Q 2017

3Q 2017

4Q 2017

Original process

2Q 2018

Study of conditions

Research on
Technologies for
Reducing Waste
Production

3Q 2018

Investigation andcomparison of technologies

Study of conditions
Reporting

4Q 2018

Study ofapplicability to 1F

Investigation andcomparison of technologies

Revised process
Research on
Technologies for
Reducing Waste
Production

1Q 2018

Reporting

Study of applicability to 1F

Reporting

To be reviewed
based on the
results of actions
ahead ofschedule©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production

No.57- Technical requirements for α nuclides measurement at 1F site  Technical requirements for α nuclides measurement were clarified from the following three viewpoints:
 Measurement environment on 1F site (workspace, on-site dose rate, and influence of mixed βγ nuclides)
 Operation of α nuclides measurement technology in the process of waste management (what and when
to measure)
The following timings were assumed to be feasible since the measurement of α nuclides contained in waste is difficult
once they are stored in containers.
• Before waste is detached from the structure/equipment (on-site measurement)• After waste is detached and before it is put into bags (such as vinyl bag)
• After waste is detached and put into bags
(Measurement from the outside of the bag or cutting a part of the bag)

 Measurement performance (sensitivity, measurement time requirement, and ability to measure
distribution)
Examples of requirements for α nuclides measurement
Work space
Width of corridors:
approx. 1.5 m in
R/Bs with possible
presence ofobstacles
Assumedmeasurement Measurement
should be
conditions
possible even if
andobstacles exist
requirementsin the corridor.
(Contact
measurement is
not required)On-site dose rate

Influence of
mixed βγ nuclides

R/B 1F corridor:
1.9-36 mSv/h
α/βγ =
Unit 2 operation
1/10 6-8
floor: max. 800
mSv/hMeasurement
should be donetoo highly
radioactive for a
without theworker to
influence of βγ
conduct
nuclide.
measurements
directly

MeasurementtargetProtection by vinyl
sheet covers is
assumed.
Measurement
should be done
over protection
covers.

Sensitivity

Measurement
time period

Maximum
Shorter time
permissible surface
requirement
contamination
fordensity in themeasuremen
controlled area:t and setup
4 Bq/cm 2 (α)
is desirable
Material transferfrom thestandard: 0.4
viewpoint of2
Bq/cm (α)
workerSet the above
exposure
criteria as
reduction.
tentative targets.

Contamination
distribution
measurement
From the viewpoints ofcontamination spread
prevention and waste
amount reduction
The identificationof hot spots andthe measurementof contamination
distribution should
be possible.

A comparison study on α nuclides measurement technologies’ principle, in terms of theirapplicability, was conducted and applicable technologies were selected in light of themeasurement environment and operation methods on 1F site.
©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production

No.58

Reference information: Consideration of measurement environment and device operation
method on 1F siteEstimated measurement conditions
Direct measurement by workers is difficult due to highradiation.
Access by remote operation device is difficult due to thepresence of obstacles
The shapes of measurement targets are very different, suchas wall, pipes, and equipment.

[1] Quoted from the material for an IRID design review on Feb 14, 2017 of the “Development of
Repair Technology for Leakage Points inside PCV” under Subsidy Project of Decommissioning
and Contaminated Water Management in the FY2015 Supplementary Budgets

Various shapes (pipes/plane floor)

Difficult to approachwith a detector and acurt

Environmental image inside R/B on 1F[1]Idea of operation taking into consideration themeasurement environment
The following two equipment using methods were considered based on the assumption that measuring equipment ismounted on a remote controlled transporter to realize measurement in a highly radioactive environment: Device for a wide area: Traveling through corridors to collect information (see concept illustration in Figure 9
below)
 Device for narrow areas: Reaching difficult-to-access locations to collect information (see concept illustration in
Figure 10 below)


Attachment

Detector
Figure 9. Detector Mounted on Relay Transporter(concept illustration)

Figure 10. Detector Mounted on Kobra(concept illustration)©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production No.59
Reference information: Identification of measurement environment on 1F site (α nuclides concentration)• Stagnant water in basement: (1) < 7×10^-4, (7) 3.6×10^-3• Stagnant water in basement: (7)<5×10^-4• Stagnant water in basement: (7)4.5×10^-4• Sludge: (1) 0.7 to 1.1×10^0, (6) <0.3 to 1.7^0[Unit: Bq/g]• - Stagnant water in basement: (2) <3<4×10^-4• - Stagnant water in basement: (5) 6.2×10^-4• Slurry from expanded ALPS: (6) 2.1 to 3.6×10^-1α-contamination level on 1F was clarified.
• Concentration ratio of α nuclides to βγ
nuclides is about 1/106-8.

Legend>10^010^-1 to 10^-0
10^-3–10^-1<10^-3

Reference cited(1) Analysis result of waste samples (stagnant
water in Unit 2 & 3 PCVs, stagnant water
including sludge in Unit 1 turbine building, coresample bored from the operating floor of Unit 1
reactor building) Feb. 23, 2017 JAEA
(2) Radioactivity analysis of stagnant water andtreated water, Nov 28, 2013, JAEA
(3) Analysis of rubble collected in the premise of
Fukushima Daiichi NPS, Apr 28, 2016, JAEA
(4) Radioactivity analysis of rubble collected in
the reactor buildings of Fukushima Daiichi NPS,
Oct 1, 2015, JAEA
(5) Analysis of supply water/discharged waterto/from water treatment facilities for theradioactivity assessment of secondary waste
generated from contaminated water treatment,
Mar. 31, 2016, JAEA
(6) Analysis of waste samples (rubble on Unit 1
reactor building 5th floor, sludge in the 1st
basement of Unit 1 turbine building, slurry frommulti-nuclide removal equipment HIC) Feb 23,
2017, JAEA
(7) Analysis result of waste samples (stagnant
water, treated water from water treatment
facilities, etc.) Jul 27, 2017, JAEA

• Concrete around reactor well on operation floor: (6)6.0×10^-1• Paint of deck plate on the ceiling of the operation floor: (1)9.4×10^-1• - Heat-retention material on 1F: (3) 0.6 to 1.3×10^-1• Floor coating around reactor well on operation floor: (4)6.1×10^-1• Concrete surface coating on 1F: (3) 8.2×10^-2• Deck plate of the operation floor: (3) 1.4 to 2.3×10^-3• - Stagnant water in PCV: (1) 5.8 to 9.4×10^-1• - Concrete on 1st floor: (3) 1.7×10^-1• - Heat-retention material on 1F: (3) 1.9×10^-1• Stagnant water in basement: (2) <0.3 to
2.4×10^-3• Stagnant water in basement: (5) 0.8 to
2.4×10^-3• Stagnant water in PCV: (1) 2.2 to 2.4×10^-1• - Concrete on 1F: (4) 0.3 to 2.7×10^-2• Heat-retention material on 1F: (4) 8.2×10^-3• Concrete on 1F: (4) 1.7 to 4.6^-3©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production No.60- Summary of on-site applicability of α nuclides measurement technologies  Applicability of existing measurement technologies was assessed based on the results of study
on requirements in light of measurement environment and operation method on 1F site.
 Remote measurement using excitation of nitrogen was selected as a highly applicable method
because it does not require contact measurement and has a potential to locate hot spots andalso enables contamination distribution measurement.
 The detail study of test device details will be necessary for the applicability of the 1F site.
Total α-ray measurement
Measurement
technology

Direct measurement (automatic)
Contact scanning

Remotemeasurement

OthersRemotemeasurement(automatic
smearing )γ-ray spectrum
measurement

Passive neutron
measurement

Mass analysis（LA-ICP-MS）

Spectrometry(LIBS)

Measurement of the

Principle

Width of work space is a
few dozen cm or more

On-site dose rate
(Max. 800 mSv/h)Direct measurement of luminescence of α-ray excited Smear sampling by Measurement of γ-ray from
nitrogenU-235 and nuclides
α-ray from α nuclides
filter paper
contained in debris○
Limited access by remote
operation device

◎:

◎:

◎:

Influence of mixed βγ
Normally 〇nuclides(To be assessed(Ratio of α to βγ nuclides is separately if concentration
is high)the order of 10^-6 to -8)
×
Measurement target(applicable only to special
Covered by vinyl sheet, etc.
shapes)Sensitivity
0.4 Bq/cm2 : Transfer
standard
4Bq/cm2: Density limit◎:

Measurement time period○Contact scanning isrequiredContamination distribution
measurement○
Scanning is needed

Others◎:(α-ray range spacerequired)○
Limited access by
remote operation
deviceLimited access by
remote operation device

Limited access by
remote operation
device◎:◎:◎:Impact on
lower detection limit

◎:

◎:

◎:

◎:

◎:

◎:

(To be assessed
separately if
concentration is high)○Not protectable○
Shielding required

○

○
◎:

◎:

Wiping is needed○

Dark environment isrequired○○
Limited access by
remote operation
device

○

◎:

○

○

(Possibility ofsensitivity improvement)

2D distribution
measurement
is possible

Mass analysis of isotopes
Elemental masssuctioned after evaporated by spectroscopy of surfaces
laserfrom laser-induced plasmas

Limited access by
remote operation
deviceNormally 〇◎:Estimation based on theneutron generation ratefrom Cm-244 and the like○
Area segmentation isneededInfluence ofmeasurement distanceand γ-ray BG○○
Scanning
measurement
is required○○
Not protectable×
Low emission rate,
low efficiency

Dependent onenvironmental dose
rate
Collimator andscanning are needed○Not protectable

◎:

×

×
Insufficient sensitivity

×

Dependent on
Dependent on
positioning accuracy,
positioning accuracy, etc.
etc.

○
Collimator andscanning are needed○○

Scanning is needed

Scanning is needed◎:

Applicable○: Limitedly applicable
×: Not applicable©International Research Institute for Nuclear Decommissioning

(III) Research on Technologies for Reducing Waste Production No.61
Reference : Principle of selected α-contamination measurement technology
 Measurement principle(1) An α particle reacts with nitrogen in its flying distance (a few centimeters) andgenerates a few hundred photons of ultra-violet light.
(2) The strength of α-ray is measured by collecting generated ultra-violet lightthrough a lens and counting the number of photons using a photon counter.
Visible light filter
Detector(PMTs)

Mirrorγ-ray shielding member

Light emission in
ultraviolet range

Focus lensUltra-violet light

α-rayLight emission

Uranium, etc.

Emitted light spectrum
Cited from J.Sand, Remote Optical Detection of
Alfa Radiation, IAEA－CN-184/23©International Research Institute for Nuclear DecommissioningNo.62

4. Study of Disposal Strategy and Safety Assessment
Methods Applicable for Characteristics of Solid Waste©International Research Institute for Nuclear DecommissioningNo.63

Contents of Report Characteristics of 1F solid waste Study of disposal strategy and safety assessment methods applicablefor characteristics of solid waste on Fukushima Daiichi Nuclear Power
Station (1F) Survey on disposal concepts and safety assessment methods
 Identification of important reference cases➣ List of important reference cases➣ Focused important reference cases and examples reflected toneeds for Fukushima Daiichi NPS
 Investigation of background of important reference cases
 Summary of Survey Results Investigation and study plan until the end of FY2018©International Research Institute for Nuclear Decommissioning

Characteristics of Solid Waste
at Fukushima Daiichi Nuclear Power Station (1F)
■ Characteristics ofthe 1 F solid waste

No.64

Dispersion
and diffusion

Water treatment
systemRubble, felled trees, etc.
RubbleFelledtrees

Contaminated
water

Soil

Very large volume, spreading
into wide areas
Projected waste from debris
Lack of previous experience withretrieval
the treatment and disposal oftrees and soil
 Very large amount with much volume of
Mainly surface contamination byhighly radioactive waste Currently some waste is difficult to sample
dispersion and diffusion with
due to difficulty in accessing themsome penetrating contamination
by way oforiginatedstagnantfrom
water*: Contamination
activated materials or reactor

Secondary wastes generated fromcontaminated water treatment
Secondary waste
fromcontaminatedwater treatment

Replaced pipes,
storage tanks, etc.

 Lack of performance record in treatment
and disposal Some waste at the site is difficult tosample.
 The amount of produced waste andcontained nuclides can be partiallyestimated based on the system features

operation waste may be included.

Increase thecapacity ofdisposalfacilities■ Needs required for safe and reasonablesolid waste disposal, considering thecharacteristics of 1F solid waste

Investigations were conducted from three
aspects shown on the right considering
the characteristics of 1F solid waste

Acceptabilityof diverse
waste

Response touncertain
risks©International Research Institute for Nuclear DecommissioningNo.65

Study of solid waste disposal strategy suitable for the characteristics of
1F solid waste and methods to assess the safety of the strategy
Survey on disposal strategies and safety assessment methods in other countries• Survey on establishment method of disposal concepts• Survey on disposal concepts and safety assessment methods at disposal facilities

Detailed Information ofinternational cases

Research
cycle

Investigation on backgroundand grounds for importantreference case

Study based on survey results from Japan and other countries• Organizing of information on waste disposal strategies and safety assessment methods applied in Japan• Study on survey issues• Extracting of important reference cases based on the survey results from other countries

Important reference cases- Disposal concepts- Safety assessment method- Predisposal management

The latest R&D outcomes related tothe characterization and predisposal
management of solid waste at the 1F

Study on safe and reasonable disposal concepts and safety assessment methods• Organizing of information on characteristics of 1F solid waste• Study on applicability of important overseas cases in othercountries• Study on safe and reasonable disposal strategies andsafety assessment methods©International Research Institute for Nuclear Decommissioning

Survey on disposal concepts and safety assessment methods

No.66

Research on the waste disposal facilities in overseas which are available for various kinds of wastes andthe detailed public data was conducted. *1

Drigg LLWR*2 (UK)
■ Contents of research
(1)
(2)
(3)
(4)

Brief survey on disposal facilities
Research on waste characteristics
Research on waste pretreatment
Research on waste containers

SFR*3 (Sweden)(5) Research on waste treatment(6) Research on waste storage(7) Research on acceptability of waste at
disposal facilities(8) Research on disposal strategies

WCS*4 (USA)(9) Survey on safety assessment methods(10) Research on safety cases(11) Research on cost optimization processes

Exhaustive research was
conducted■ Study on research result• Comparison with previous cases in Japan• Importance determined based on needs of 1F solid waste

Listing of important reference cases*1. Quick research was conducted on cases at disposal facilities in Russia*2. A low-level radioactive waste disposal facility in Drigg, UK operated by LLWR repository Ltd.
*3. A short-lived low- and intermediate-level radioactive waste disposal facility in Östhammar Municipality, Sweden, operated by SKB Sweden AB*4. A low-level radioactive waste disposal facility in Texas, USA, operated by WCS©International Research Institute for Nuclear DecommissioningIdentification of important reference cases

No.67<List of important reference cases>

Additional survey was
conducted on the importantreference cases from thefollowing view points:• Background and evidences• Advantages• Disadvantages andrestrictions

 Study of important reference cases from the viewpointsof outlines, advantages, and the assumption of effects
for the applicability of the 1F site.

Sorted listed cases to find
how to reflect needs of solid
waste at Fukushima Daiichi
Nuclear Power Station (1F)©International Research Institute for Nuclear Decommissioning

List of Important Reference Cases at LLWR and SFR (1/3)

No.68

No.

Importantreference cases

AppliedinGeneral description

1

Inventory records andmanagement system

UK

An inventory management system developed by the UK government and NDA (UKRWI). Updated
every three years. It serves as a reference for waste management and disposal planning.

UK

A systematic approach to decision making based on factors such as safety, environmental
protection, technology feasibility, regulatory approval and financial costs It is intended to promotewaste management optimization while balancing safety with technology and costs.

2

BPEO/BAT

3

Systematic process toachieve clearanceUK

A practical method (code of practice) for the statistical characterization of difficult to measurewaste and a systematic approach for achieving clearance using this method. The same method
enables waste disposal without prior approval from regulatory bodies.

4

LLW management
strategy

UK

The UK government's waste management strategy to reduce the load on LLWR by making mostof the waste that was conventionally disposed of by LLWR very low level waste (VLLW) andapplying the same disposal procedures as general industrial waste to them.

5

Disposability
Assessment (DA)UK

Evaluation to confirm whether waste and waste bodies meet waste acceptance criteria It can be
an indicator to determine the treatment requirements for waste from operation and waste fromdecommissioning.

6

Limitation of shortlived nuclides

7

Waste categorization considering half-life in addition to total radioactivity Based on this
LLWR
categorization, some cases outside Japan show short half-life ILW shallow land disposal on theand SFRpremise of long-term institutional control.

Development of WAC
LLWR Waste acceptance criteria that serve as criteria for the restriction of radioactive materials as wellincluding for hazardous
and SFR as chemicals that may affect the environment or disposal..
materials©International Research Institute for Nuclear Decommissioning

List of Important Reference Cases at LLWR and SFR (2/3)
No.

8

9

Important reference Applied
casesinLarge size container
Overpacking of supercompacted waste
drums

No.69

General description

LLWR

At LLWR, ISO containers are used to store waste, and waste bodies are produced by applying
grouting solidification to the containers that store waste. The use of ISO containers is superior interms of transport efficiency, handling at the disposal facility, and cost.

LLWR

At LLWR, drums compressed by a super compressor are stored in ISO containers. The use of asuper compressor can significantly reduce waste volume.

10

Using waste as a
backfill material

LLWR

The use of extremely low level radioactive waste as backfill materials for disposal facilities isrecommended. Although reduction in the final waste volume is expected, there is no cases ofsuch use in the UK due to the variation of waste characteristics and the complexity of the process.

11

Multiple Container
sizes

SFR

Waste containers with different shapes used at SFR for the purpose of enhancing the volumetricutilization efficiency of cylindrical silos.

12

Disposal of large items

LLWR A method to bury large-size waste at disposal facilities without using a container by directly
and SFR solidifying them on site when the waste is too large to put into a container.

13

Large sized disposal
vaults

LLWR

A large-scale vault type disposal facility with a multilayer barrier structure.. An increase in disposal
capacity and an improvement in disposal efficiency are expected.

14

Prevention of human
intrusion by capping

LLWR

Multi-layered barrier cap with a bio-intrusion layer on the bed of rocks. Reduction of intruder’s
exposure to radiation is expected, which needs to be assessed according to the human intrusion
scenario in the safety assessment protocol.

15

Prevention of human
intrusion by location
undersea

SFR

Reduction in the risk of accidental human intrusion by constructing a disposal facility under thesea floor.

©International Research Institute for Nuclear Decommissioning

List of Important Reference Cases at LLWR and SFR (3/3)

No.70

No.

ImportantAppliedreference casesin

16

Co-location of
disposal concepts

17

Institutional
control

18

Process byagreement

19

Generic disposal
safety case20

Insight modelling21

The assessment of the impact of long-term climate change, including the quantification of coastal
Long-term climate LLWR anderosion impacts on shallow land disposal, which is used for the safety assessment of sites wherechange modelling
SFR
significant climate change is expected..

22

Post-closure
degradation ofengineered
barriers

SFR

General description
Waste disposal strategy to dispose of VLLW and short half-life ILW at a single site by constructing
both a land disposal facility and a vault type facility together.. Construction costs can be reduced bysharing the same site and infrastructure.

Institutional control consisting of the prevention of intrusion into the facility and performance
LLWR and guarantee by monitoring the disposal system. The prevention of human intrusion over a long periodSFRof time and reduction in the activity of short half-life nuclides are expected by long-term institutional
control.
UKThe phased process of consensus-based regulatory approval in the UK , which includes early phase
dialogue with regulatory agencies prior to a formal disposition application

UK

Comprehensive and general safety cases before the identification by the site , which are referred torepository for the extraction of R&D themes and the evaluation of the possibility of disposal.
geologicalUK

A predictive and simplified safety assessment model for sensitivity analysis and bounding analysis,
which is used to identify important parameters for disposal.

Safety assessment using time-dependent parameters, taking into account the deterioration of the
LLWR and engineered barrier, which is expected to be effective for the prevention of an overly conservative
SFR
forecast by suggesting a risk of instantaneous loss of barrier performance after the end of theinstitutional management period, for example.

©International Research Institute for Nuclear Decommissioning

No.71

Identification of important reference cases<Focused important reference cases and examples of those reflected to 1F needs>
Items highlighted byblue fill are detailed on
the following slides.

WCS
LLWR

Enhance containment performanceof facilities and upsize disposal
facilities

Use of modular concrete canisters
LLWR

WCS

SFR

Production of waste bodies using large
containers

LLWR

Use of very low-radiation waste as backfill
materials for disposal facilities

WCS

SFR

Increase thecapacity ofdisposalfacilities

Use of absorbent

Use of other standard containers according
to the waste type

LLWR

SFR

Issues associatedwith solid waste on
1F

Setting of short-half-life
medium-level waste
category

Assessment of uncertainty by
probabilistic safety assessment

LLWR
SFR

WCS

LLWR

Waste in different categories
are disposed in the samefacility

Acceptabilityof diverse
waste

Response touncertainrisks

Systematically organized
clearance procedure

Packaging various types ofwaste in the same container
LLWR

Practice of institutional control
LLWR

SFR

WCS

Facility design to prevent human intrusion

Control and publication of
information about the activity
and content of waste
LLWR

Requirements for acceptance ofwaste bodies that containchemically hazardous materials
LLWR

SFR

WCS

LLWR

WCS

Degradation of engineered barriers after
closure of disposal facility©International Research Institute for Nuclear Decommissioning

Investigation of background  Large sized disposal vaults
of important reference cases Enhance containment performance of facilities
Cap system structure

No.72
LLWR

and upsize disposal facilities

Overall structure• Cap (prevention of water permeation from the upper area andhuman intrusion)• Wall (prevention of lateral water permeation within the facility)• Bottom plate (prevention of upward water permeation fromunder the facility)*

Appearance ofgeomembrane(asphalt sheet)*HI: Number of stacked level of ISO containers©International Research Institute for Nuclear Decommissioning

Investigation of
background of importantreference cases

 Prevention of human intrusion
Facility design to prevent human
intrusion

No.73
LLWR

SFR

WCS

LLWR in UK•

A cap with a bio-intrusion barrier built on a bed of roundrocks with a diameter of 600 mm is employed as ameasure to enhance protection against human intrusion by
excavation.
WCS in USA••

Class C waste is buried at a depth of 5m or more from thesurface to protect unintended intruders according to the
Texas State Code.
The modular concrete canister (MCC) used at WCS has astructural strength that prevents human intrusion to Class C
waste.
SFR in Sweden••

The coastal submarine location prevents human
intrusion.
Assuming an uplift over a 1,000-year period, the disposal
facility was built at a depth of 60 m from the seafloor sothat it will remain below the seafloor after the uplift.
©International Research Institute for Nuclear Decommissioning

Background study  Large size container
Production of waste bodies using largeof important
containers
reference cases

LLWR

SFR

WCS

No.74

Improvement of waste disposal efficiency by the scale-up of disposal facilities and the use of largesized containers
LLWR

SFR

WCS

Appearance

ISO container or half height ISO(HHISO) container• Drum• Mold• Concrete tank
• ISO container

Modular concrete canisters(MCC)

Dimensions

H1.32 m × W2.5 m × L6.06 m(HHISO)

Various dimensions

Inner diameter 2 m × inner height
2.8 m

Strength

9-level stack is assumed

Up to 42-level stack is assumed

34.5 MPa (up to 6-level stack)Solidification
agent

Superplasticizer-containing PFA+ Portland cement

Asphalt
Concrete

High-strength grout(compressive strength after 28
days ≥ 2,000 psi)

Notes

A HHISO container is used more
often because of its handleability.

Containers are put in a silo andcement-grouted.

A layer of sand or soil is added in
between every stacked MCCs.

©International Research Institute for Nuclear Decommissioning

Background studyof importantreference cases

 Large size container
Production of waste bodies using large
containers

LLWR• The homogeneity of grouted waste in acontainer is secured by using a grout injection
port with a flow straightener disposed at its
lower part and inclining the container in thecontainer solidifying process.

SFR

WCS

No.75

Grout injection port

• Voids (small-bores) may be generated in theinside of pipes because grout may not go into
the pipes.
• Waste bodies are made of a mixture of wastewith different characteristics but their inventory
is known as each waste is managed under
UKRWI (UK Radioactive Waste Inventory)
from the initial stage of generation.

Cutting a solidified container

Flow straightener(baffle)• When using a large-sized container, difficulty in homogeneous
solidification and technical issues to overcome it are expected incomparison to the use of a small-sized container.
• In addition, measurement using a measuring device for a small waste
body that is mainly used domestically at present and the method ofdetermining the inventory based on such measurement cannot be applied
to a large-sized waste directly.
The use of a large-sized container requires the following
technology and method:• Technology to realize the homogeneous
solidification of waste in the container• Method to determine inventory of waste bodyCut cross-section of a solidified
container containing pipes©International Research Institute for Nuclear Decommissioning

Background studyof importantreference cases

 Post-closure degradation of engineered barriers
Degradation of engineered barriers after closure of
disposal facility

No.76
LLWR

WCS

• An overly conservative assessment of uncertainty is avoided by considering degradation.
• In many safety cases, the degradation of engineered barrier’s confinement performance after the end of theinstitutional control period is considered in the safety assessment.
(Those safety cases presume that engineered barrier’s confinement performance will be maintained during theinstitutional control period because serious defects or deterioration of the barrier will be repaired during institutional
control.)• Note that such assessment method varies greatly depending on the country.
• LLWR and SFR estimate gradual degradation in engineered barrier’s confinement performance.

Comparison of assessment for the degradation of engineered barrier’s confinement
performance between Japan and other countries
Japan (actual case)

UK

SwedenUSA• The assessment is basedon the hypothesis thatthe water permeation
prevention performanceof the existing facility
deteriorates to a level ofthat of sand immediately
after burying.

• The assessment isperformed throughprobabilistic evaluation
taking into accountchange over time after
defining the probability
distribution function (pdf)of hydraulic parameters.

• The aim of assessment isto ensure a 1,000 year
long confinement after
facility closure.
• A deterministic
assessment is performed
using different
parameters taking into
account change overtime.

• Although the structure of theconcrete container is expected
to be maintained for 300
years, it is still shorter than therequirement from the longterm (50,000-year) safetyassessment. Therefore, thelong-term safety assessment
does not rely on the concreteconfinement performance.

©International Research Institute for Nuclear Decommissioning

Background studyof importantreference cases

 Post-closure degradation of engineered barriers
Degradation of engineered barriers after closure of
disposal facility

No.77
LLWR

WCS

Probabilistic hydraulic parameters of the engineered barrier used at LLWR in UK (2011ESC)
Hydraulic parameter

Point of time
(A.D.)

Hydraulic conductivity ofcontainer (m/s)

Duringinstallation

Hydraulic conductivity of a
gap between containers(m/s)
Hydraulic conductivity ofvault (m/s)
Hydraulic conductivity ofcap (mm/yr)

Percentile

Duringinstallation
Year 7000
Duringconstruction
Year 2180
Year 3180

Hydraulic conductivity ofimpermeable wall (m/s)

Duringconstruction andyear 2180

Year 3180
Year 7000• These parameters were derived from expert elicitation in UK.
• Many barriers are comprised of multiple components.
For example, the base of a vault is constructed with geomembrane, bentonite, and concrete.
©International Research Institute for Nuclear Decommissioning

Background studyof importantreference cases

 Hazardous materials
Requirements for acceptance of waste bodies
that contain chemically hazardous materials

No.78
LLWR

SFR

WCS

Materials with impact on disposal inside and outside Japan
Issues

Japan

UK

USA

(1) Substances that
affect nuclide
migration• Chelating agent• Organic materials(cellulose)• Rubber• Inorganic phosphate• Organic phosphate• Carboxylic acid andpolycarboxylic acid
• Amino polycarboxylic
acid (EDTA, NTA, DTPA)• Liquid(Limited to 1% ofcontainer capacity)(2) Materials affecting
barrier
performance• Soluble salts
NaCl, Na2SO4, NaNO3• Under investigation• Organic materials(Total organic material
content is limited to 5%.
There is no restriction
when putting in MCC. )Notes

Sulfates: React withcement or otherhydration products andreduce its confinement
performance.
Salts: Increase thehydraulic conductivity ofbentonite.

According to the latest
findings, it is concluded
that the influence of isosaccharic acid, etc., issmall.
(The soil environment of
LLWR is pH 11 so thatorganic materials aredecomposed by microbialactivity.)

Solidification or adding
absorbent 2 times asmuch as the liquid involume in the container isrequired for liquid.

©International Research Institute for Nuclear Decommissioning

No.79<Case study> - Case of Disposal facility in Russia (NZK-type disposal facility*1) •A new-type near-surface disposal facility for lowlevel solid waste and short-lived intermediatelevel solid waste, currently planned in Russia.

•

The facility is divided into 20 independent
modules that can store waste for a long time
(50 years) with the ability to retrieve the buried
waste.

•

Waste is put in long-durability reinforced
concrete containers (NZK-150-1.5P) andsolidified airtightly with a special non-shrink
concrete mixture.

•

Multilayer protective coatings used for the final
cap include biological barriers and anti-filtration
screens made of concrete, crushed stone, clay,
geotextile, geomembrane, silt, sand, andtopsoil.

Plan of a new-type near-surface disposal facility (NZK-type)

Disposal facilities in Russia will be investigated in detail in FY2018.

*1. A new-type near-surface disposal facility currently under planning©International Research Institute for Nuclear Decommissioning

Summary of Survey Results

No.80

 Case studies were conducted on three international disposal
facilities
 Exhaustive research was done on 11 items.
 Important reference cases were listed through comparison withactual cases in Japan and in consideration of needs associatedwith waste at Fukushima Daiichi Nuclear Power Station (1F).
 A list of 22 cases was created.
 Backgrounds of important reference cases are surveyed alongwith consideration of how to reflect these case studies to theneeds of the1F Applicability of important reference cases to the 1F will be
examined in FY2018©International Research Institute for Nuclear Decommissioning

Investigation and study plan until the end of FY2018
FY2017

No.81

FY2018

Item
1Q

2Q

3Q

4Q

1Q

2Q

3Q

4Q

1. Survey on waste disposal methods and safetyassessment methods applied in other countries
Requirement survey on generic safety cases(1) Survey on establishment of disposal concepts(2) Survey on disposal concepts and safetyassessment methods
2. Study based on survey results covering Japan andother countries(1) Organizing of information on waste disposal
concepts and safety assessment methods
applied in Japan

Exhaustive research + backgroundinvestigation of important reference cases

Research ofcases in Japan

In-depth research

Survey feedback
Survey feedback(2) Identification of important reference cases(3) Research subject review
Survey feedback

Survey feedback

3. Study of safe and reasonable disposal concepts andsafety assessment methods(1) Organization of the characteristics of solid
waste in Fukushima Daiichi NPS(2) Study of the applicability of important overseas
cases from other countries(3) Study of safe and reasonable disposal concepts
and safety assessment methods©International Research Institute for Nuclear Decommissioning

No.82

5. Integration of R&D Results(Study on Waste Stream)©International Research Institute for Nuclear Decommissioning

No.83

Contents of Report Summary of Implementation Plan and Results Reflection of Research Results➣ Example and Effect of Waste-Specific Input Information
Management Sheet➣ Example of Refined Treatment Flow Result of Integration into Waste Stream Summary of Waste Stream Study©International Research Institute for Nuclear Decommissioning

No.84

Summary of Implementation Plan and Results
FY

Implementation plan

Goal achievement index

2017• The promising waste streams presented in FY2016 will be comprehensively
evaluated with respect to the progress, consistency of the outcomes, andremaining issues by reflecting the latest results obtained in previous research.

• Presentation of the progress, the consistency of theoutcomes, and remaining issues

2018• The waste streams will be repeatedly examined by reflecting issues and research
results obtained in FY2017, and present evaluation results based on it.

• Establishment of comprehensive methods to evaluate
progress, consistency, and issues and the presentationof evaluation results based on them

 Summary of results Methods to comprehensively manage the progress andissues in R&D were designed and tried (Figure).
• In order to ensure the comprehensiveness of target
waste, necessary information and issues were identified
using the classification of treatment flows created, basedon the list that enumerates all types of waste on 1F(waste list), and a waste-specific input information
management sheet and a list to verify matching between
issues and outcomes were created.
• In order to introduce schedule management,
consideration time schedules were set for each step ofwaste management process for individual waste basedon the Mid-and-Long-Term Roadmap and the storageand management plan, and a time schedule arrangement
table was created.
• In order to assign priorities and organize methods, thetotal image of waste management is clarified from theresult of organization as listed above and flows areintegrated into the form of a waste stream.
 The progress, the consistency of the outcomes, andremaining issues were clarified using the establishedmethod.

Input

Process

OutputAssurance ofcomprehensiveness(Treatment flow)

Clarification of necessary
information and issues

Waste-specific input
information management
sheet

List to verify matching between
issues and outcomes forindividual waste

Information necessary inconsidering each stepand issue

Research results
Mid-and-Long-Term
Roadmap
TEPCO’s
waste storage andmanagement plan

Introduction of schedulemanagement
Consideration scheduling foreach step for individual waste

Assign priorities andorganize methods

Time schedulearrangement table

Total image of waste management

Clarification of total image of
waste management

Total Image of Waste Stream Development©International Research Institute for Nuclear DecommissioningReflection of Research Results

No.85< Example and Effect of Waste-Specific Input Information Management Sheet >R&D progress status is indicated by colored text, and the consistency of results
and remaining issues can be clarified.
各ステップへのインプットする情報
廃棄物

核種組成
放射能濃度

－

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－処分対象廃棄物の物量処分対象廃棄物仕様(浸出性、含有物、化学組成、核種組成、放射能濃度、発熱性、強度、線量率、ガス発生程度)
容器の仕様(材質、形状、表面線量率)－

廃棄体化物の物量
廃棄体化後の廃棄物仕様(浸出性、含有物、化学組成、核種組成、放射能濃度、発熱性、強度、線量率、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

廃棄前保管物の物量
廃棄前保管後の廃棄物仕様(浸出性、含有物、化学組成、核種組成、放射能濃度、発熱性、強度、線量率、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

処分・再利用（将来検討）
・処分区分
・廃棄体技術基準－

廃棄前保管対象廃棄物の物量
処理物(固化体)の仕様(含有物、化学組
成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度、強度)容器の仕様(材質、形状、表面線量率)－

廃棄体化物の物量
廃棄体化物の仕様(含有物、化学組成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度、強度)容器の仕様(材質、形状、表面線量率)

廃棄前保管（将来検討）
・保管容量
・保管方法
・廃棄確認方法

原廃棄物の発生量
分別程度
原廃棄物の仕様(形状、含有物、化学組
成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

一時保管物の物量容器仕様(耐用年数、材質、形状、表面線
量率)
分別程度
一時保管後の廃棄物仕様(形状、含有物、
化学組成、核種濃度、放射能濃度、ガス発
生程度、飛散性、発熱性)

前処理・処理・再加工物の物量
前処理・処理・再加工後の廃棄物仕様(含有物、化学組成、核種濃度、放射能濃度、ガス発生程度、飛散性、発熱性)容器仕様(材質、形状、表面線量率)

保管物の物量容器仕様(耐用年数、材質、形状、表面線
量率)
保管後の廃棄物仕様(含有物、化学組成、
核種濃度、放射能濃度、ガス発生程度、飛散性、発熱性)廃棄体化（将来検討）
・処理容量
・処理方法

－

－

前処理・処理・再加工対象廃棄物の物量
処理物の仕様(形状、含有物、化学組成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

保管（2020年）
・保管容量
・保管方法

廃棄体化までの保管期間(廃棄体化の処理
期間)

原廃棄物の発生量
原廃棄物の仕様(形状、含有物、化学組
成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

一時保管物の物量容器仕様(耐用年数、材質、形状、表面線
量率)
分別程度
一時保管後の廃棄物仕様(形状、含有物、
化学組成、核種濃度、放射能濃度、ガス発
生程度、飛散性、発熱性)

前処理・処理・再加工（2020年）
・処理容量
・処理方法

原廃棄物の発生量
分別程度
原廃棄物の仕様(形状、含有物、化学組
成、核種組成、放射能濃度、含水量、発熱量、腐食性、飛散性、ガス発生程度)
容器の仕様(材質、形状、表面線量率)

一時保管（保管継続）
・保管容量
・保管方法

廃棄物（発生継続）
・廃棄物の物性把握

原廃棄物を保管するのに必要な性状調査
要求

廃棄物

一時保管

処理物を保管するのに必要な性状調査
要求

前処理・処理・再加工物の受け入れ方法
廃棄体化処理方法

廃棄体を健全に保つ観点での保管への要求

レベル区分の必要性の有無
核種濃度制約
発熱量制約

廃棄確認項目
廃棄確認方法

－－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

廃棄確認項目
廃棄確認方法前処理・処理・再加工までの保管期間(再処理・処理・再加工の処理期間)－

廃棄体化までの一時保管期間(廃棄体化の
処理期間)

－

原廃棄物を前処理・処理・再加工するのに必要な性状調査要求－

原廃棄物を廃棄体化するのに必要な性状
調査要求－

前処理・処理・再加工

保管

廃棄体化

Progress can be understood visuallyfor each waste type, such as good
progress or still in the planning stage

廃棄前保管

－

処分に必要な性状調査要求

処分・再利用各ステップへインプットする課題

作成:2017年度
国プロ「固体廃棄物の処理・処分に関する研究開発」での実施内容から以下の通り色分けした。
黒字:国プロの研究計画外
青字:国プロにて着手予定、一部検討済み
緑字:検討完了

Stream S10 Slurry from multi-nuclide removal systems
shows good progress in the collection of information mainly
on waste (characteristics), pretreatment, treatment, andreprocessing with the aim of achieving the start of storage in2020 (mostly blue and green text).

Stream S1 for RPV waste, which isscheduled to be considered in the future in
the time schedule arrangement table, this isin the stage of planning R&D activities(mostly black text)©International Research Institute for Nuclear Decommissioning

Green text: Consideration completed; Blue text: Planned under a subsidized program, or consideration partly completed;Black text: Not included in R&D plan under subsidized programs (Color coding will be further reviewed)
Stream S1 RPV shows the flow from installation of RPVs in each Unit of Fukushima Daiichi NPS to disposal of such RPVs.
Stream S10 Slurry from multi-nuclide removal systems shows the flow from generation of radioactive sludge waste in the pretreatment process of multi-nuclide removal systems to disposal of such waste.

No.86

Reflection of Research Results< Example of Refined Treatment Flow >To clarify necessary information and issues, the progress of refinement is confirmed based on the

input information management sheet.
Waste
generation

Temporarystorage

Contamination
category: XI-4
Multi-nuclides
storage HIC(slurry)

HIC storage

Pretreatment, treatment
and reprocessing(stabilization)
Continue withcurrent storage

Promising options applicable to waste that can betreated at existing treatment facilities and waste forwhich refinement is underway through R&D
activities were presented.
HIC storage

Leakage expansion
prevention

Sorting

Waste
conditioning

Advanced leakage
prevention storage
Storage in anothercontainer

StorageFinal
confirmation of
disposal

Storage incontainer

Disposal

Basic flow
Dewatering

Solidification[Legend]

HIC residuals

Storage

Metal(SUS reinforcing
members)

Stream S6

Combustibles(PE container, etc.)

Stream S8

Decontamination

Dismantling andsorting

Promising options and options that go with existing facilities
or plannedoptions
facilitiesthat await further refinement
Promising
Other options
Not adopted options

Example of Stream S10 Slurry from multinuclide removal systems©International Research Institute for Nuclear Decommissioning

No.87

Result of Integration into Waste Stream
In order to clarify the total image of waste management, refined treatment process flows areintegrated into waste streams
Temporary storage

Secondary wastes generated fromcontaminated water treatment

Temporarystorage(Stream S13)

Concentrated waste liquid slurry(waste liquid concentrated byevaporation)

Restoration measures, such as stagnant water treatment

Areas of policies established in the solid waste storage and management plan and the like

Areas of policies underconsideration

Waste generation

Adsorption vessels (KURION,SARRY, mobile purifier), multinuclide removal systems (slurry,
adsorbent, adsorption vessel),filters (high-performance multinuclide filter)

Adsorption vessels, HIC
storage facility(Streams S9, S10 and S12)

Large-size waste storage
repository

Pretreatment,
treatment andreprocessing

Waste
conditioning

Storage

Equipment and facilities to treat, store, and condition
the secondary waste of water treatment(Streams S9, S10, S11, S12, S13 and S14)(Streams S9, S10 and S12)

Treatment and storage of slurry from multi-nuclide removal systems (S10)P/MB D-pit(Stream S11)

Sludge from decontamination

Temporary storage(Streams S6 and S7)
Tent, cover soil, outsidecollection, outsidecontainer storage, solid
waste repository

RubbleMiscellaneous combustible
solid waste(Protective wear, combustible
materials)

Combustible materials
temporary storage(Stream S8)Felled trees

Temporary storage(Stream S8)Contaminated soil,
etc.

Temporarystorage

Temporary treatment systems
and tanks

Treatment and storage of sludge from decontamination (S11)

Expanded solidradioactive waste
storage facility(Streams S6 and S7)Volume reduction
treatment equipment

Expanded solid radioactive
waste storage facility(Streams S6 and S7)

Miscellaneous solid
waste incineration
system(Stream S8)(Streams S6 and S7)

Incineration ash
storage(Stream S8)

In FY2017, the pretreatment,
treatment, reprocessing, andstorage processes of thestreams for slurry from multinuclide removal systems andfor sludge fromdecontamination were updated

Equipment
and facilities
for metal andconcreteconditioning

Pre-incineration treatment equipment
Expanded miscellaneous solid waste
incineration system

Predispos
al storage(All
streams)Dispo
sal/recycle(All streams)

Contaminated soil
temporary storage
facility
Decontaminati
on systems

Temporary storage

Waste from the dismantling of reactors
and reactor buildings

Storedwaste

Areas in need of policy establishment after deep
deliberation

Secondary waste, etc.
Secondary waste, etc.

Waste potentiallygenerated inassociation withdismantling

Secondary waste, etc.

Equipment and facilities for the treatment and storage of metallic waste
and concrete(Streams S1, S2, S4 and S14)
Secondary waste, etc.

Secondary waste, etc.

Stream S14
A combination handled in the same way as the following process flows

Waste generated before
debris retrieval work

RPV (Stream S1), PCV metal (S2), PCV concrete (S3), metal in R/B (S4), concrete inFuel debris-adhered
R/B (S5), multi-nuclide removal system (S10), and filter (S14)
waste

Metal
(SFP, wastes in on-sitebunkers, etc.)

Plastics

Storage

Handling of stored
waste needs to beconsidered fromnow.

This chart will be reviewed and may
be revised based on the progress of
R&D activities.

Fuel debris and fueldebris-adhered waste
storage facility

In operation
In planning©International Research Institute for Nuclear Decommissioning

Summary of Waste Stream Study
Waste
Stream S1

Reactorpressure vessel(RPV)

Stream S2

Metal of primarycontainment
vessel (PCV
metal)
Concrete of primarycontainment vessel (PCV)

Stream S3

No.88

The waste-specific R&D status andrefinement were confirmed.
Study on waste streams will be repeatedby reflecting issues identified andresearch results obtained through R&D
activities, and the evaluation results
based on such reviews will be presented

Stream S4

Metal in reactor
building

Pretreatment, treatment and reprocessing methods (decontamination, volume reduction, andreprocessing methods) need to be considered based on the characteristics of nuclear waste(such as content, radiation concentration, and the shape of waste dependent on the dismantling
method) and the specifications of temporary storage containers.

Stream S5

Concrete in reactor
building

Treatment methods (decontamination, volume reduction, and reprocessing methods) need to beconsidered based on the characteristics of waste at the site (such as content, radiation
concentration, and the shape of waste dependent on the dismantling method) and thespecifications of temporary storage containers.

Stream S6

Metal rubble

Volume reduction by shredding is planned.

Stream S7Concrete rubble

Volume reduction by crushing is planned.

Stream S8

Combustibles

Felled trees undergo a volume reduction process. Protective wear is incinerated.

©International Research Institute for Nuclear Decommissioning

No.89

6. Schedule and Project Organization©International Research Institute for Nuclear Decommissioning

No.90Planned and Actual Schedule (1/3)
Schedule of FY2017 “R&D for Treatment and Disposal of Solid Radioactive Waste”
Implementation details

LegendBlue: Plan
Red: Result

FY2017
4

5

6

7

8

9

10

11

12

1

2

3

1. Characterization(1) Collection and management of analysis data
(1) Identification of contamination distribution

Annual analysis
planning

Planning for nextyear

Mid- and longterm planning- Development of analysis plan
*Analysis plans will be reviewed as necessary.
(Oct 4)- Sampling and transportation(Nov 21)(Dec 15)(added on Nov 10)(added on Mar. 29)* Sampling will be performed throughout the year. Transport dates ofsamples may be changed when necessary, depending on theReportingreadiness of them.

- Conducting analysis* Analysis data will be reported at around the end of every quarter term. Reporting timing may bechanged depending of the content of analysis data.

- Study of waste classification based on analysis data

Consideration Summary*The study results are reflected in the study of analysis methods.

(2) Development of sampling technology- Sampling of secondary wastes generated fromcontaminated water treatment, including sludge(Collection of cesium adsorbent, etc.)

Preoperational planning

Discuss specification of specific sampling device foradsorbent from the cesium adsorption vessel(Sampling of adsorbent from the Cesium adsorption
vessel No. 2)

Preoperational planning

Choose specific sampling device for adsorbent from Cesium
adsorption vessel No. 2(Collection of sludge from decontamination systems)

Survey planning

Survey of periphery and inside of D-pit

Preoperational planning

Consideration of the purpose ofsample collection in the reactor
buildings- Study of sampling methods in R/Bs

Summary

Summary

Production of sampling jigs and formulation of
construction plan

Clarification of local
circumstances

Summary©International Research Institute for Nuclear Decommissioning

No.91

Planned and Actual Schedule (2/3)
Implementation details

LegendBlue: Plan
Red: Result

FY20174

56(3) Streamlining of analysis methods

7

8

9

11

12

Consideration ofcontamination mechanisms- Study on the migration behavior and the contamination
mechanism of radionuclides

Selection of
nuclides2

Viewpoint of streamlining

Viewpoint of streamlining
analysis method(4) Analysis data management- Building of analysis database

Specification review- Organize and update waste data

Preoperational
planning

Summary

Database creation
Consideration of data management and update policy,
publication of updated data set

Technical development plan
discussed

Operation andidentification of issues
Summary- Analytical evaluation methods

Planning

Accuracy improvement study andthe establishment of methods

Summary- Collection of basic data

Planning

Collection of basic experimental
data

Summaryiii) Summary of comprehensive inventory evaluation

3*Conduct a review in line with the analysis plan and reflect the result to the plan.

- Streamlined analysis methodii) Improvement of the accuracy of analytical evaluation methods

1
Summary

Study of data
representativeness- Study on the representativeness of analysis data
- Reselection of nuclides to be analyzed

10

Not included in FY2017 R&D planiv) Response to materials with impact on disposal, etc.
- Case study inside and outside Japan

Planning- Investigation of impact on disposal safety

Planning

Research on cases
Study of indexes that can express impacts on
performance and quantitative evaluation methods

Summary

Summary©International Research Institute for Nuclear Decommissioning

No.92

Planned and Actual Schedule (3/3)
LegendImplementation details

FY2017

Blue: Plan
Red: Result

4

5

6

7

8

9

10

11

12

1

2

3

2. Study on Predisposal Management
Issues in applying to waste generated
by the accident

Planning

Consideration of applicability ofexisting technologies- Measures for hydrogen gas generation (case study outside Japan)

Planning

Research on the international know-how ledge of hydrogengas control in storage

Summary- Measures for projected wastes generated by fuel debris retrieval

Planning

Research on projected rubble from planned fuel debris
retrievalSummary- Management method of waste originated from cesium adsorption
vessels (JAEA)

Test condition considerationand preparationi) Assessment of applicability depending on the characteristics of solid waste

Summaryii) Consideration and evaluation of waste storage and management methods taking
into account the characteristics of solid waste(1) Study of measures for storing highly radioactive waste(2) Evaluation of technology to stabilize the secondary wastes generated fromcontaminated water treatment- Applicability assessment of in-drum glass solidification technology

Testplanning- How to stabilize sludge from decontamination systems

Testplanning

Tests with full-scale simulated
waste samples

Fundamental test on glass
solidification

Summary

Fundamental test on fluidity

Planning of demonstration tests

Research on α-contamination measurement andassessment technologies*1iii) Research on Technologies for Reducing Waste Production

3. Study of Disposal Strategy and Safety Assessment Methods Suitable for
Characteristics of Solid Waste

Preoperational planning

4. Integration of R&D Results (Waste Stream)

Preoperational planning

Consideration of management
methods, conclusion

Searching for target disposal
facilities

Note 1)

Investigation and evaluation of
disposal regulations

Comprehensive evaluation of progress, consistency of outcomes,
and remaining issues

Summary

Summary

Summary

Note 1) This was originally scheduled in FY2018 but completed ahead of schedule.

©International Research Institute for Nuclear Decommissioning

Tokyo Electric Power Company Holdings, Inc.
(TEPCO)Development coordination (1) to (4)R&D liaison department of TEPCO• Waste Management Group, Project Planning Department,
Fukushima Daiichi Decontamination and Decommissioning (D&D)
Engineering Company

Japan Atomic Energy Agency(JAEA)
Summary of technical development(1) to (3)Development (1) to (4)Operational assistance (5)Hitachi, Ltd. (1)*Inspection and calibration of radiation monitoring
instruments
Chiyoda Technol Corporation (1)*
APD inspection and calibration
Hitachi, Ltd. (1)*Inspection and calibration of GM counters
Hitachi Transport System, Ltd. (1)*Transport of analysis samples from Fukushima
Daiichi NPS to analysis laboratories
Hitachi Transport System, Ltd. (1)*Transport of analysis samples collected at
Fukushima Daiichi NPSAsima Co., Ltd. (1)*Commissioned operation and maintenance ofequipment and facilities related to backend
technology development buildings
Mitsubishi Materials Techno Corporation (1)*Inspection and maintenance of power receiving
equipment in backend technology developmentbuildings
Mitsubishi Materials Techno Corporation (1)*Inspection and maintenance of instrumentationequipment in backend technology developmentbuildings
NISSIN GIKEN CO., LTD (1)*General maintenance/repair and efficiency
measurement of high-performance air filtration
systems in backend technology developmentbuildings
Ascend Co., Ltd. (1)*Commissioned research on analysis and tests
associated with the characterization of thesecondary waste from contaminated water
treatment systems in Fukushima Daiichi NPS

No.93

International Research Institute for Nuclear
Decommissioning (IRID)

Project Organization
Chart (FY2017)o Formulation of overall plan and technical integration (1) to (5)o Technical management such as technical development progress
management (1) to (5)o Coordination concerning technical development (1) to (5)Hitachi-GE Nuclear Energy, Ltd.

Toshiba Energy Systems &Solutions Corporation

 Development (1) to (4) Development (1) to (4)Tokyo Power Technology Ltd. (1)*Support of information handling
operations for waste list update
JGC Corporation (1) and (3)*Case study in other countries concerning
acceptance criteria of waste including
influential materials to disposal strategy• Support of survey on reasonable disposal
methods applied in other countries*Participation by a public invitation

Mitsubishi Heavy
Industries, Ltd.
 Development (1), (2) and (4)AECOM E&C UK Limited (2) and (3)*• Survey on hydrogen gas evaluation methods
and storage in other countries (UK., USA,etc.)
• Survey on international know-how of waste
disposal strategies and safety assessment
methods (in UK, USA, etc.)

KOKUYO Co.,Ltd. (5)Organization of documents and data

ATOX Co., Ltd.
 Development (1), (2) and (3)Kanden Engineering Co. (1)*Sampling tools design and manufacturing

MHI NS Engineering Co., Ltd. (1)Mock-up test support

TAIHEI DENGYO KAISHA, LTD. (1)On-site investigation support

ANADEC (2)Sludge fluidity test
Inspection Development Company Ltd. (1)*Commissioned research on the analysis of βnuclides in contaminated water and secondary
waste from contaminated water treatment systems
in Fukushima Daiichi NPS as well as subsequentorganizing and interpretation of the analysis data
Art Kagaku Co., Ltd. (1)*Anemometer calibration
Nippon Nuclear Fuel Development Co., Ltd. (1)*Analysis of contaminated water and secondary
wastes generated from contaminated water
treatment
E&E Techno Service Co, Ltd. (1)*Tests concerning the contamination distribution
analysis of waste samples from Fukushima Daiichi
NPS
SynchroSoft Co.,Ltd. (1)*
Design and prototyping of database for the analysis
of waste in Fukushima Daiichi NPS
Central Research Institute of Electric Power
Industry (1)Study on accuracy improvement of analyticalevaluation methods

Mitsubishi Materials Corporation (1)*Collection of data concerning nuclide sorption
behavior to support the inventory estimation ofwaste zeoliteTokyo Nuclear Services Co., Ltd. (1)*Collection of basic data concerning nuclide
sorption on adsorbent in liquids with different
characteristics

Mitsubishi Materials Corporation (1)*Research on influencing factors and influence
propagation of waste generated by the accident
as well as nuclide migration parameter settings(II)

National Nuclear Laboratory, UK (2)Study on waste management (hydrogen gas
generation)

Ascend Co., Ltd. (2)*Full-scale tests to examine salt concentration
behavior using SARRY test samples

Orano ATOX D&D SOLUTIONS Co., Ltd. (1)Study of methods for sampling secondary waste
generated from high radiation water treatment

Orano ATOX D&D SOLUTIONS Co., Ltd. (2)Detailed study on methods to estimate hydrogengas generation and requirements for hydrogengas venting system

Kurion Japan, KK (2)Fundamental test on glass solidification[Types of operations](1) Characterization(2) Study on predisposal management(3) Study of disposal strategy and safety assessment
methods suitable for characteristics of solid waste(4) Integration of R&D results(5) R&D management, etc.

©International Research Institute for Nuclear Decommissioning