Annual Research Report
2018

International Research Institutefor Nuclear Decommissioning

Greeting
The “Mid-and-Long-Term Roadmap toward the Decommissioning of Fukushima Daiichi Nuclear Power Station(NPS), Tokyo Electric Power Company (TEPCO) Holdings Co., Inc.” was revised by the government. It is based on
the updating of the Technical Strategic Plan for 2017 Decommissioning of Fukushima Daiichi NPS, TEPCO Holdings
Inc., by Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF) in September 2017.
It indicated that retrieving fuel debris at the bottom of the primary containment vessel (PCV) shall be performed inadvance by focusing on the “Partial submersion side-access” retrieval method as “Fuel Debris Retrieval Policy”, anddecide the fuel debris retrieval method for the initial unit in fiscal year 2019.

The International Research Institute for Nuclear Decommissioning (IRID) has been engaged in the research anddevelopment (R&D) of technology required in the decommissioning of the Fukushima Daiichi NPS as an urgent issue
since being established in August 2013. This has resulted in the situation with the PCV and the nuclear reactor being
clarified upon as well identifying the technical issues to be overcome by using the development of technology forinvestigating inside the PCV and detecting the fuel debris using cosmic rays.
The Annual Research Report 2018 is intended to summarize the achievements of the R&D projects (15 subsidized
projects and 2 in-house research) undertaken by the IRID in FY2018. We would appreciate if this report helps tounderstand the R&D achievements and what IRID has been responsible for.

Eight years have now passed since the accident that occurred at the Fukushima Daiichi NPS after the Great East
Japan Earthquake. The situation has been largely improved upon when compared to just after the accident, however,the decommissioning work is about to enter a crucial phase. The IRID is committed to proceeding with our R&D on
the steady and efficient nuclear decommissioning in thereby fulfilling our responsibilities.
We sincerely appreciate your kind guidance and continued support.

March 2019

Hideo Ishibashi

President of International Research Institute for Nuclear Decommissioning

Contents
2
4

6
8
10
12

Major Research Achievements Subsidized Projects

14
16
18
20
22
24
26

Major Research Achievement In-House Research

28

29

Others - Datasheet

32

Development of Technology for Detailed Investigation inside PCVDevelopment of Technology for Detailed Investigation inside PCV(On-site demonstration of technology for detailed investigation consideringcountermeasures for deposits)(On-site demonstration of technology for detailed investigation through X-6 penetration)
Development of Technology for Investigation Inside RPVDevelopment of Sampling Technology for Retrieval of Fuel Debris and Internal Structure
Development of Technology for Fuel Debris Analysis/ Characterization

Development of Technology for Water Circulation Systems in PCVDevelopment of Technology for Water Circulation Systems in PCV (Full-scale Test)

Upgrading of Approach and Systems for Retrieval of Fuel Debris and Internal Structures
Upgrading of Approach and Systems for Retrieval of Fuel Debris and Internal Structures(Development of Technology for Establishment of Criticality Control Method)

Upgrading of Fundamental Technology for Retrieval of Fuel Debris and Internal Structure

Development of Fundamental Technology for Retrieval of Fuel Debris and Internal Structure(Development of Small Neutron Detectors)
Development of Technology for Collection, Transfer and Storage of Fuel Debris
R&D for Treatment and Disposal of Solid Radioactive Waste

R&D for Treatment and Disposal of Solid Radioactive Waste(R&D on Preceding Processing Methods and Analytical Methods)(1) Investigation Research on Deposit at the Lower Part of Pedestal in Fukushima Daiichi NPS
Unit 2(2) Test and Research for the Stable Operation of Advanced Liquid Processing System and Slurry
Stabilizing Facility
Major Research Results in FY2018

List of Contract Research in FY2018
List of Major Research Facility/ Equipment

IRID’s R&D Projects in FY2018 (Overview)

Technology for Investigation/ Analysis
inside Reactor

Subsidized Projects

R&D for Preparation of Fuel Debris Retrieval
Technology for Fuel Debris Retrieval

On-siteDemonstrationof Technology for
Detailed Investigation
Considering
Countermeasuresfor Deposits

On-siteDemonstrationof Technology
for Detailed
Investigation
through X-6
Penetration

Development ofSampling
Technology for
Retrieval of Fuel
Debris and Internal
Structure

Upgrading ofApproach and
Systemsfor Retrieval of Fuel
Debris andInternal
Structures

Developmentof Technology
for Investigation
inside RPVDevelopmentof Technology
for Fuel
Debris Analysis/
Characterization

Research andStudy on Depositat the Lower Partof Pedestal in 1F
Unit 2Development ofTechnology for
Establishment of
Criticality Control
MethodDevelopmentof Small Neutron
Detectors

Developmentof Technology for
Water Circulation
Systems in PCV

Upgrading ofFundamental
Technology
for Retrieval of
Fuel Debris
and Internal
Structures
Developmentof Technology
for Collection,
Transfer andStorage of Fuel
Debris

In-House Research

Developmentof Technology
for Detailed
Investigation
inside PCVDevelopmentof Technology
for Water
Circulation Systems
in PCV (Full-scale
Test)

R&D for Treatment and Disposal of Solid Radioactive Waste
R&D onPreceding
Processing
Methods andAnalytical
Methods

Test andResearch
for the Stable
Operation of
ALPS and Slurry
Stabilizing
Facility

DataSheet

R&D forTreatment andDisposal of Solid
Radioactive
Waste

Subsidized Projects (15 projects)
In-House Research (2 projects)
1F: Fukushima Daiichi NPS

Overview of Reactor Building and IRID’s R&D Project
Fuel Debris

Reactor Building
Spent Fuel Pool

Reactor Pressure Vessel(RPV)

R&D for Preparationof Fuel Debris
Retrieval
Dry Well (D/W)
Vent Pipe

R&D for Treatment
and Disposal ofRadioactive Waste
Torus
Room

Primary Containment
Vessel (PCV)

Suppression Chamber(S/C)
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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Detailed Investigation inside PCV
Background
Melted fuel debris are considered to have fallen into the pedestal at Fukushima Daiichi Nuclear Power Station (NPS) Units 1-3. Valuable
information was obtained from investigation of the pedestal inside and outside in Units 1-3 until so far, however, the needs for fuel debris
retrieval were not satisfied due to dimensional restrictions of the existing penetration opening.

PurposeThe project of the “Detailed Investigation inside PCV” is amed to enlarge the size of access device and to upgrade the measuring
technology for the device to meet the needs of fuel debris retrieval. Additionally, this project is intended to develop the device to perform
more detailed investigation through a newly established larger access route.

Major Approach and Results

1 Investigation and Development PlanningA submersible access device that can move a wide range of the basement floor by accessing through X-2 penetration into PCV wasselected for Unit 1 as planned based on the latest situation and investigation results of the pedestal inside and outside of Unit 1 and 3.
An arm type access device whcih can move in the air by accessing through X-6 penetration into PCV was selected for Unit 2 based on theinvestigation results of the Unit 2 pedestal . Additionally, the investigation results for Unit 1 and 2, and measuring technology for accessdevice were studied to update and to concrete the plan.

2 Development of Access and Investigation Device① Establishing Access Route into PCV via X-6 Penetration
An isolation room that will be the PCV boundary when opening the X-6 penetration hatch was designed and produced, and a
combination test of the isolation room and the remote-operated hatch opening device was conducted. In addition, the design andproduction of connecting tubes with isolation valves to be connected with X-6 Penetration by the remote operation in the isolation
room (Reference: Fig.1 “X-6 Penetration Connection Structure”) were performed to confirm various functions (connections to X-6
penetration, flange gripping, PCV boundaries (airtight), isolation functions of isolation valves and remote self-driving).

② Establishing Access Route into PCV via X-2 Penetration
The PCV access route device such as extension tubes, isoation valves and guide pipes to connect a double door opening hole ofX-2 penetration, which will be a new boundary, were designed and produced. In addition, equipment that is required for opening
holes for inside/outside of double doors and the grating in PCV, and guide pipe installation separated from PCV inside wasdesigned and produced. The performance, retrieval methods and the construction conditions were confirmed to extract issues andcountermeasures (Reference: Fig.2 “In-plant Validation Status of Opening Hole in X-2 Penetration Inner Door”).

③ Access/ Investigation Device
The investigation device for Unit 1 that is to carry in/out access device separated from PCV and 6 kinds of submersible type accessdevice were designed and produced. The performance, methods and combination operation were confirmed to extract issues andcountermeasures (Reference: Fig.3 “In-plant Validation Status of Submersible Type Access Device”).
The arm type access device (the arm and its enclosure) which has access to PCV inside from X-6 penetration was produced for Unit
2 (Reference: Fig.4 “Production Status of Arm Type Access Device”).

3 Applicability Validation of Element Technology

The design and production of the systems for the deposit 3D shape measurement, the deposit thickness, fuel debris detection andidentification of the access device were performed for Unit 1. The performance, measurement operations, measurement disturbanceof a single and combination device were confirmed to extract issues and countermeasures. (Reference: Fig.5 “Status of Compatibility
Validation for Measurement of Deposit 3D Shape”).
For the contribution to the fuel debris estimation for Unit 2, sensors were produced to gether information of the shape and dimension,and dose rate in PCV under high radiation environments, and the compatibility validation were confirmed (Reference: Fig.6 “Status ofCampatibility Validation for Laser Scanner”).

Future Development

This project developed the access route establishment, a mockup test with the device for the detailed investigation inside PCV,
operation training and the on-site verificaiton in Unit 1 and 2, which will be followed under the projects for FY2018 and FY2019 (“Projectof Development of Technology for Detailed Investigation Inside PCV (On-site demonstration of technology for detailed investigationconsidering management of deposits and on-site demonstration of technology for detailed investigation through X-6 penetration).

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Subsidized Projects

Isolation Valve

Isolation Valve

3D ModelAppearance

Fig. 1: X-6 Penetration Connection Structure
Water Jet

X-2 Penetration X-2 Penetration
Inner Door
Outer Door

Isolated Part

Rail

Cutting
Supporting

Nozzle Part

Cut Confirmation

Telescope Part

Enlarged AWJ Head

Fig. 2: In-plant Validation Status of Opening Hole in X-2 Penetration Inner Door
Guide Ring

Guide Ring 2
Simulated Interference Object(Horizontal Piping)
Guide Ring 1

Simulated
Interference Object Guide Ring 3 Guide Ring 4(Vertical Piping)Access Door for Workers

Pedestal Outside
Pedestal Inside

Jet Deflector
Pedestal

ROV
Unit 1 Jet Deflector

Jet Deflector

Access Doorfor Workers

Unit 1 Access Doorfor Workers

Fig. 3: In-plant Validation Status of Submersible Type Access Device
Tool Changer

Enclosure

Arm Parts

Fig. 4: Production Status of Arm Type Access Device

MEMS Unit

Radiation Resistance Camera

Laser Injection Window
Side Mirror

ROV for Sediment 3D Mapping

50A Piping(Half Buried)
Anchor

100A Piping(Half Buried)

Water depth 2.1m

Example of Sediment 3D Mapping

Fig. 5: Status of Compatibility Validation formeasurement of Deposits 3D Shape

Example of Laser Scanner 3D Mapping

Fig. 6: Status of Compatibility Validation
for Laser Scanner
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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Detailed Investigation inside PCV(On-site demonstration of technology for detailed investigation considering management of deposits)

Background
Fuel debris in the Fukushima Daiichi NPS Unit 1 exists under the deposits located inside/outside of the pedestal. To retrieve the fueldebris, the information on the deposits and the situations in the pedestal is necessary.

PurposeThe purpose of the project is to establish an access route through X-2 penetration into PCV inside and to perform the site verification in
Unit 1 using a submersible investigation device with measuring technology.

Major Approach and Results

1 Investigation and Development Planning

A mockup test plan and the test facility design for Unit 1 were performed. Working hours and radiation exposure were calculated
from the mockup test results of the access route establishment to review on-site verification plan. In addition, the sequence of 6 kinds ofsubmersible investigation device for the detailed investigation was reviewed.

2 Development of Access and Investigation Device① Establishing Access Routes in PCV
The mockup test for new boundary connections with X-2 penetration, cutting of inner/outer doors and interference objects, andguide pipe installation confirmed that work under the isolated state from PCV inside can be conducted. After operation training, thepreparation of the site verification was started. (Fig. 1: “Status of AWJ Mockup Test for 350A”).
② Development of Access and Investigation Device
The mockup test facility simulated a work area and the half circle of the inside-outside pedestal was produced to start the test.

Future Development

The access route establishment and the site verification for detailed investigation will be performed in FY2019. The guide ring installationand detailed visual inspection will be performed for detailed investigation, and subsequent investigation sequence will be decided. The
deposits thickness measurement, fuel debris detection, the deposits sampling and the deposit 3D mapping measurement will be performed.

Development of Technology for Detailed Investigation inside PCV(On-site demonstration of technology for detailed investigation through X-6 penetration)

BackgroundAs a result of internal investigation performed in Fukushima Daiichi NPS Unit 2, the deposits that were assumed to be fuel debris wereconfirmed at the bottom of inside the pedestal.

PurposeThis project aims to insert the access and investigation device through a large opening of X-6 penetration and to verify technology for thedetailed investigation inside PCV.

Major Approach and Results

1 Investigation and Development Planning

Based on the investigation results of Unit 2 inside PCV and the design achievements of access and investigation device, the study on theaccess from X-6 penetration and the access procedure of investigation device, obstacle removal, accessible ranges (measurable ranges),data acquisitions and schedule for investigation were studied to establish the investigation plan.

2 OnSite Verification of Access Device and Investigation Technology① Mockup Test Considering Site Situation
The test procedure for the mockup test of the arm type access and investigation device was reviewed to concrete the test plan basedon the test plan studied by the project of the “Development of Technology for Detailed Investigation inside PCV.” In addition, a part of
the test facility for the mockup test was produced.
② Operation Training
The Virtual Reality (VR) system which reflects the design results of access, investigation device (specifications, dimensions, movable
ranges of each arm joint) and the PCV internal shape were established for simulation operation training (Fig.1: VR System for Access andInvestigation device).
③ Test for Establishing Access Route to PCV and Onsite Operation Training
Operation training for the hatch opening was conducted. In addition, a combination test of the isolation room and connecting tubes
with isolation valves (X-6 penetration connection structure) that will be connected with X-6 penetration by remote operations was
conducted to test cable handling and emergency separation action in case of failure (Fig.2: Status of Combination Test).
④ On-site Verification
The specifications for the neutron detection system (detectors, the sheilding facility, neutron mederators, etc.) that is designed toestimate the range and distribution of nuclear fuel existance during on-site verification, were established. Furthermore, a validation
test plan for the neutron detection system was developed as well as the design and production were started.

Future Development

On-site applicability of the arm type access and investigation device will be confirmed and evaluated by combination and mockup tests,
and the operation training will be performed. After the completion of the mockup test and operation training, the device installation for
Unit 2 PCV and on-site validation (site survey) will be performed.

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Annual Research Report 2018

X-2 Penetration Inner Door
(Present Situation: Closed)

Inner Door Perforation
Range(350A)

Inner Door Perforation
Range(200A)

PCV Inside
Handrail Removing Range

Grating
Perforation Range

PCV Outside
New Boundary Part

Inner Door Perforation X-2 Penetration
Inner Door
Range(350A)

350A Isolation Valve

Inner Door Perforation
Range(200A)

Handrail
Removing Range

200A Isolation Valve

Basement Interference
Object Removing Range

Interference Object
Removing Part

Monitoring
CameraAWJ Device

AWJ Device
Grating Perforation
Range

250A Isolation Valve

X-2 Penetration
Outer DoorAWJ Device
Pedestal Removing Range(as necessary)

Subsidized Projects

Cutting Plan

Monitoring Camera(a)Work Area (Simulated AIR Lock Room)

Handrail Removing Range

AWJ Control Device

Monitoring Camera

AWJ Device(b) Instruction and Operation Room
<Simulated Enclosure, 2F>

Fig. 1: Status of AWJ MockupTest for 350A

Arm for Investigation

Display Image (Arm is accessing PCV internal.)

Input Screen

Fig. 1: VR System for Access and Investigation Device
Simulated X-6
Penetration
Isolation room
Side Wall

X-6 Penetration Connecting Structure
Isolation room (near X-6 Penetration)

Cable

X-6 Penetration
Connecting Structure

Isolation room (cable layout)

Fig. 2: Status of Combination Test

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Investigation inside RPV
Background
Fuel debris and the reactor internals remain in the Reactor Pressure Vessels (RPVs) and therefore advance information on those location,
the shape and the conditions is necessary. However, it is difficult to directly access RPV to obtain the information due to the complicatedstructures and extremely high radiation rate.

PurposeThe decommissioning work including fuel debris retrieval is required to gather information on unclear situations to ensure a safe andefficient work. This project is intended to clarify the investigative target, and to develop required investigation technology by the reactor
core drilling method of accessing from the top or side of the RPV.

Major Approach and Results

1 Investigation and Development Planning
The information on fuel debris retrieval was organized and updated, and as the result, the requested information was resulted into
the appearance and dose rate as before. Accordingly, development of device that is assumed to establish access routes and to insert theinvestigation device up to the reactor core was planned. The investigation plan was reflected from the updated step-by-step plan toacquire the information; (1) The next step of investigation inside RPV, and (2) Study on the fuel debris retrieval method.

2 Investigation Method Planning

The concepts for the safety conditions during investigation was examined to study the safety evaluation method. Furthermore, thecontamination level of the structure to be processed was estimated based on the results of the accident progress analysis. A simplified
model of radiation exposure using airflow analysis in PCV was established to estimate the impact on radioactive dusts during processing.

3 Study on Investigation Auxiliary System

Maintaining negative pressure in PCV requires gas control, nitrogen supply systems, dust monitoring, criticality control sysytems,
water treatment sysytems and the auxiliary system which will be required from the aspect of safety and investigation. These requirement
specifications were extracted.

4 Development of Access and Investigation Device① Development of Investigation Device for Top Access Method
The device specifications were established by desk analysis and partial tests based on the achievements and remained issues in
FY2017. In addition, the plan for element tests (confirmation tests for small-diameter drilling for PCV/RPV heads, the RPV heatinsulator removal method, workability and accessibility for processing the reactor internals, and accessibility for investigation device)
was concretized to confirm the validity of the device specifications.
The study on the structure changes considering effective working procedures for the device to be handled in a work cell which isestablished as the prevention system of the spreading contamination, and the simplified work in a work cell were performed fromthe aspect of radiation exposure reduction to reflect the device specifications. In addition, the appearance simulation was performed
based on the result of visuality test for investigation device which simulates fog environments.

② Development of Investigation Device for Side Access Method
The test plan for element tests required for device specifications (cutting ability test with hybrid waterjet tool, compressed seal test,collection and drainage test for treated water, and strength test of chemical resin joint parts) was concretized to prepare for the test
facility and the device for element tests.

Future Development
Accepting the progress of FY2018, the radiation exposure evaluation required for the safety evaluation and element tests by using thepartial test model of each device will be conducted in FY2019. Based on the results, the device will be designed to establish the device andfacility specifications. In addition, investigation procedures and the maintenance method will be examined to concrete the investigation
plan.

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Annual Research Report 2018

Work
cell

Airlock
cell

Tool box cart
Spare nozzle Flange opening
cuttingdeviceremoval device

Work cell

Investigation device

Drilling device for Reactor Internals

Tool box
Gate valve
Well cover
Guide pipe

Access device (Stored into tool box)
PCV head connection part
Guide pipe

PCV head
RPV head

Seal

Steam dryer

Simulated PCV head

Steam-water
separator
RPV head drilling device
AWJ tool head

Subsidized Projects

Equipment
carry-inport

Boundary function maintenance device (Guide pipe)
Superposition simulation image by 3D
model (Assuming the center of upper gridplate is melted)

Shroud headUpper grid plate
For preliminary
confirmation

For full scale
Investigation

Image of celestial sphere
Reactor internals
drilling device

Investigation device

Fig. 1: Overview of the Top Access Investigation Method

Work cellHybridConcrete test
water-jet tool
facility
Cutting ability test withhybrid water-jet tool

Reactor building
PCV shield wall

PCV
Strength test for chemicalresin connection part

RPV
Biologicalshieldingplate
Collection anddrainage test for Drainage/waste materials,
measuring device
treated water
Drain part

Applying loads
Compressed seal test

Concrete test facility

Test seal material

SleeveLoad meter
WJD head for PCV Collection Separation unit Filtrateshielding wall
unitwith solid tankMeasure load capacityof connection part

PCV mockup

Chemical resin

PCV guide
tube
mockupFig. 2: Overview of the Side Access Investigation Method

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Sampling Technology for Retrieval of Fuel Debris and Internal Structure
Background
Identifying the component and mechanical properties of fuel debris is important to ensure the appropriate safety management and thefacility design during retrieving fuel debris which will be collected as a sample from the lower part of Primary Containment Vessel (PCV)
and the reactor. Therefore, the sampling method that collected debris sample is taken out to the analysis facility has been studied to obtain
information that cannot be taken by a camera.

PurposeThis project aims at obtaining the debris information in a safe and prompt manner. The sampling scenario is established to promote thedesign and test production of sampling device, safety studies and study on sampling device while utilizing the investigation results of inside
PCV and development results of the arm for detailed investigation inside PCV.

Major Approach and Results

1 Development of Fuel Debris Collection and Sampling Scenario
The total scenario and development plan of collecting fuel debris were reviewed and updated based on the following points.

① The phased sampling steps were reviewd considering the balance between analytical needs and analysis facility capability.
② A feasible sampling method of retrieving small amount of fuel debris was established without using special facilities during

investigation inside PCV.

③ The countermeasures for troubles with having access dificulities by PCV interference objects were extracted.

2 Fuel Debris Sampling System for PCV, and Design and Test Production of Device
The design and trial production of specific technologies required for fuel debris sampling were promoted.
① Basic Design of Fuel Debris Sampling System
The structure and specifications for the remote-operated debris sampling system for transporting from the high-dose sampling site,
and neutron monitors to detect re-criticality symptoms due to crack occurrence when cutting debris were established. In addition, thefeasibility of the remote-operated transport system for debris sampling was confirmed.

② Design and Trial Production of Fuel Debris Accessing Device
The performance enhanced measures were established to apply the arm type access device for detailed investigation inside PCV
to collect debris sample. Specifically, the enclosure structure considering the radioactive materials confinement was reviewed, andthe additional joints which can avoid the pedestal structures that is difficult to handle with the above arm type access device were
designed.

③ Design and Trial Production of Fuel Debris Sample Collection Device
The device specifications for the sample collection device structure developed last fiscal year were concretized based on theelement test to improve handling ability considering transporting debris sampling to the analysis facility, and accessing into PCV withthe arm type access device as well as easy-to-retrieve fuel debris.

3 Conceptual Study of Sampling System for Fuel Debris in Reactor Pressure Vessel (RPV)

The device specifications for the sample collection device structure developed last fiscal year were concretized based on the element
test to improve handling ability considering sampling transport to the analysis facility, and accessing into PCV with the arm type accessdevice as well as easy-to-retrieve fuel debris.

4 Trial Production of Small-Amount Sampling Device for Detailed Investigation inside PCV

Two kinds of collection devices which can be connected to the arm type access device for detailed investigation inside PCV and carriedout sealed samples from the sampling site were produced toward the completion of small-amount sampling.

Future Development⃝ Focusing on technology required for fuel debris retrieval, validation items, design specifications and development schedule were

examined. The device development will be promoted to complete early-stage fuel debris sampling since next fiscal year.

⃝ This project will develop to complete small-amount sampling in cooperation with on-going project of “Development of Technology for

Detailed Investigation inside PCV.”

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Subsidized Projects

EnclosureX-6 penetration
Connection Structure

Fig. 2: Enclosure and X-6 Penetration
Connection Structure
Considering Radioactive
Material Confinement

X-6
penetration

Enclosure

Enclosure
Horizontal Offset
Mechanism

PCV
Arm type access device
Carry-out debris sample fromhigh-radiation sampling site

Fig. 1: Concept of Fuel Debris Sampling in PCV

Debris transport
container
Gripper forsucking andgripping
Bucketfor fallingprevention

Remote-operated
transport curt

Fig. 4: Concept of Pebble-like and Sand
Shape Debris Collection Device

Fig. 3: Debris Remote Transportation System

Actuator for press
CameraActuator for brush press
CameraCap for brush
containment

Vacuum bloodcollection tube
for debris
Suction port

Vacuum vessel type method

Rotation actuator forbrush containment
Brush for debris collection

Ultrafine metal wire brush methd

Fig. 5: Example of Small-Amount Debris Collection Device

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Fuel Debris Analysis/ Characterization
BackgroundThe nuclear decommissioning requires retrieval, collection, transfer and storage of fuel debris in safe and effective manners. Therefore, itis necessary to understand fuel debris properties for decommissioning work. In addition, technology development is necessary to analyze
fuel debris sampling in the future.

PurposeThis project aims to estimate fuel debris properties through deposit analysis and a basic test with simulated fuel debris materials, and toprovide the information to the project for the decommissioning work. In addition, the development of element technology for fuel debris
analysis and investigation of transportation technology will be performed for the preparation of the future fuel debris sampling.

Major Approach and Results

1 Estimation of Fuel Debris Characterization
As part of fuel debris characterization, the surface dose rate was continuously estimated, and besides, the behavior of radioactive
airborne particles was studied and the deposits (the reactor internals deposits) during investigation inside PCV were analyzed.
Firstly, the evaluation formula for the surface dose rate established in FY2017 was modified considering the apparent burnup andradio-activation source to improve the applicability of various fuel debris while representative cases were evaluated.
Data for the radioactive airborne particles generated by dismantling the Globe Box (GB) at the existing facility was gained andevaluated (Fig.1) to understand the behaviors of radioactive airborne particles that may be an issue during fuel debris retrieval. In
addition, powder having various diameters and non-organic materials were selected as test samples, and the property data for thetransportation and migration behavior in air, in water and air-liquid interface was obtained from the test.
As for the deposit samples collected from the internal investigations for each Unit including investigation inside the Unit 3 pedestal(Fig.2), the surface of the deposits was observed, and elements and nuclides were analyzed to confirm the consistency of the estimationon the situation of the reactor internals.
These findings are summarized as the “List of Fuel Debris Property.”

2 Property Evaluation by Using Simulated Debris

Emission Behavior Evaluation of Fission Products (FP) during Dry Heat Treatment
The dry treatment is studied as one of the pretreatments for storage of the retrieved fuel debris, and therefore, the volatile FP was
evaluated as a basic data for the off-gas treatment of facility. Focusing on the FPs among severely evaluated nuclides that were selectedbased on environmental emission evaluation in FY2017 and considered that the vulnerability may change due to the treatment
conditions, their chemical compounds were experimentally evaluated for temperature andemission start temperature/ emission
temperature as emission behaviors. In addition, the change of surface area and the impact of air pressure for the migration of volatile FP
off-gas during the dry treatment of fuel debris were confirmed.

3 Development of Element Technology for Fuel Debris Analysis

For the preparation of fuel debris sample analysis, the analysis instruction for all of twenty two items was prepared since FY2017. A
series of a work flow from the acceptance of test samples up to the disposal and return was reviewed by considering the consistency
and efficiency of the analysis results.
As for development of the analysis method that multi-nuclides are analyzed with the new type ICP-MS (Inductively Coupled Plasma
Mass Spectrometry)simultaneously and promptly, the types and flow rates of the reaction gas that sufficiently reduces nuclides impact
hindering during the nuclides measurement were studied. As the result, it was confirmed that original nuclides to be measured andhindering nuclides can be sufficiently separated excluding a part of nuclides. Excluded nuclides which are insufficient to separate from
hindering nuclides are found out to be separeted from hindering nuclides before the system installation.
Hearing investigation was conducted at the existing facilities in Ibaraki Prefecture where will accept a candidate facility for transport
containers. Prior to acceptance, issues were clarified whether the containers can be accepted without a large scale of construction.

Future Development
Preparing analysis for a small amount of fuel debris sample will be continuously conducted for the plan of transporting collected sample
to Ibaraki area by using containers. The fuel debris properties will be obtained from actual fuel debris data. Furthermore, the behaviorof radioactive airborne particles will continue to be studied for collecting the behavior data of fine articles generated during fuel debris
retrieval work in cooperation with domestic and international research institutes.

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Annual Research Report 2018

Dust Monitor Count Rate in Work Area (Integrated Value)(cpm)A) Leaving from work areaA) Cutting end, Clean up start

Dust monitor filer
paper replacement

Increase of atmosphericconcentration due to assumed
airborne (re-floating)radioactive particles during andafter work.

A) Shred of small diameter pipeA) Shred of large
diameter pipeA) Pipe shred starting
Workerentering into
work area(A, B)

Subsidized Projects

A) Main worker (cutting)B) Assistant worker (Packing)B) Packing ofB) Leaving from work area
dismantled object

Time (hh, mm, ss)Fig. 1: Example of GB Demolition Work at Existing
Hot Facility
Radioactive particles will be dispersed during various
work at contaminated areas. This project studies
dispersed radioactive particles generated by fuel debris
retrieval based on the data of particles generated fromthe GB demolition work.

SEI

10μm

U

10μm

Cr

10μm

Fe

10μm

Zn

10μm

Zr

10μm

ペデスタル
Unit 機3 Pedestal

Fig. 2: Wiping Deposits (photo in the center) andAnalysis Result by Field Emission Type Electron Scanning Microscope (Figure in right)
The deposits attached to the investigation robot that was returned from the pedestal inside of Unit 3 was wiped
off with smear filter paper and cotton-tipped stick. Element analysis for the surface of the wipped deposits which
was observed. The data showed that the reactor internal condition was consistent with the analysis results.

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Water Circulation Systems in PCVDevelopment of Technology for Water Circulation Systems in PCV (Full-scale Test)
Background
The project of “Upgrading of Approach and Systems for Retrieval of Fuel Debris and Internal Structures” is also promoted to improve
environment for the risk reduction and to ensure safety of fuel debris retrieving work toward the decommissioning of the Fukushima
Daiichi NPS. Therefore, the establishment of the water intake structure is needed as the water circulation system of Primary Containment
Vessel (PCV).

PurposeWater intake parts of the water circulation system has challenging issues to be addressed for the confinement function of gas phase andliquid phase, achieving a long-term integrity and on-site remote workability during installation and operation under high radiation
environments. This project aims to develop applicable technologies for accessing and connecting to PCV to intake water in the water
circulation system.

Major Approach and Results

1 Clarification of Technological Specifications, Work Plan Study and Development Plan for

Upgrading Water Circulation System in PCV① Study on Water Circulation System and Technology Using Drywell (D/W)⃝ Based on various investigation results so far, access route establishment technology for the D/W and the suppression chamber(S/C) water intake was studied and clarified for environmental dose rates on the first floor of the reactor building (R/B) inside/outside of of PCV, estimation results of the reactor internal situations, deposits in PCV and water levels in PCV (current situations,
during construction and operating the water circulation systems) for each Unit.
⃝ In addition to these results, the technological specifications (in common with D/W and S/C) for the access route establishment in
PCV and the D/W water intake design were examined based on the study results of the water circulation system developed by the
project of “Upgrading Approach and System for Retrieval of Fuel Debris and Internal Structures,”and the candidate penetration of
the D/W water intake port was selected considering the site environment.
⃝ As a result of study on existing technologies, technology for establishment of the access route from outside to inside PCV, provided
a perspective on applicability of existing technologies including access methods into PCV which was proved by another project, andthereby, issues of the access route establishment, a route up to a water intake point, were clarified that; 1) The remoteoperated
pump (piping) suspension method, 2) The remote operated connection and replacement method for piping.

② Study on Water Circulation System and Technology Using S/C⃝ Design specifications of the S/C water intake part structure were clarified based on the common study results of (1). In addition,
the candidate location of the S/C water intake part structure considering the site environment was selected for each Unit.
⃝ The function requirements of the S/C water intake part were examined based on the study results by the project of “Upgrading
Approach and System for Retravel of Fuel Debris and Internal Structures.” The S/C water intake structure and maintenanceprogram were established to satisfy the requirements.
⃝ Existing technologies were studied based on establishment and maintenance of the S/C water intake part. The study result
showed that a main development issue is a connection method of extended pipes, and therefore, a policy was determined that theapplicability was confirmed by element test.
⃝ The project development items were summarized among applicable device during construction and maintenance.

2 Development and Validation of Element Technology for Access and Connection to PCV⃝ The access and connection in the D/W and S/C will be verified by eleement tests in FY2019.

3 Full-scale Validation of Technologies for PCV Access and Connection⃝ As part of cooling water management flowed out from the D/W of to the Unit 1 S/C inner side of torus room, the validation test planof the water circulation system boundary at the inside of torus room was developed, which utilized the full-scale test facility of Naraha
Center for Remote Technology Development, Japan Atomic Energy Agency (JAEA Naraha Development Center). This test is to confirm thepossibility of utilizing the space of the S/C inner side of torus room as the boundary of contaminated water and fuel debris powder.
⃝ As the test preparation of the JAEA Naraha Development Center, operation and maintenance of the heating and water supply facilities,
and the turbid water discharge facility were performed.

Future Development

A development plan of the project will be established. The element tests will be conducted to verify a remote-operated pump (piping)suspension required for the D/W water intake part establishment, pipe connection methods, replacement methods during maintenance, andconnection methods of extension pipes required for the S/C water intake part establishement. In addition, tests will be conducted to confirm
workability for the S/C water intake part establishment by full-scale remote operation and to verify the water circulation system boundary.

12

Annual Research Report 2018

Cooling facility

Corrosion
control tank

D/W insidewater intake

Water level monitoring

To watertreatment system

Adsorption building/RO Unit
(Nuclide/ salt content removal)

Filtering facility(solid content removal)

S/C insidewater intake

Criticality monitoring
Water level monitoring

Buffer Tank

Subsidized Projects

Boronic acidwater tank① D/W water

intake part

Note: Image description is shown Water level monitoring② S/C water intake partTo water treatment system

Scope of Project Development: Location of Water
Intake Candidate

Fig. 2: Selection Result of the D/W Intake
Candidate Penetration (Example: Unit 3)

Fig. 1: Water Circulation System during Fuel Debris Retrieval
PCV outside (R/B1F)

PCV inside

Personal airlock (X-2)

Extension pipe
D/W water
intake pump
A used penetration part will be selected as the D/W water
intake port depending on requirements of the water
circulation system layout. This figure shows the example of the
airlock for staffs (X-2) as a water intake port.

Fig. 3: Image of Water Circulation System Using D/W
Suspending an
extension pipe fromthe 1st floor of R/B

Fig. 4: Selection Result of the S/C Water Intake
Structure Candidate Locations (Example: Unit 3)

Torus room

Underground
water
Adding displacement
absorption mechanism
between R/B1F and the S/C

SUS for corrosion
prevention
Connection structure
avoiding dissimilar
metal connection at
the siteFig. 5: S/C Water Intake Structure

Strengthen support columns→ Separation of underground water andcontaminated water① Injection of repairing materials② Injection of contaminated water and simulated
fuel debris powder③ Confirmation for the leaking situation ofcontaminated water and simulated fuel debris
powder④ Confirmation for the immersion state ofcontaminated water and fuel debris powder
simulation body due to the S/C cutting

Fig. 6: Overview of Validity Test for Water
Circulation System Boundary13

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

 pgrading of Approach and Systems for Retrieval of Fuel Debris
U
and Internal Structures
Background
Fuel debris in the Reactor Pressure Vessel (RPV) and Primary Containment Vessel (PCV) of Fukushima Daiichi Nuclear Power Station(NPS) has been currently secured a cooling condition. However, the reactor building, RPV and PCV were damaged by the accident and theplant itself is in an unstable situation. This project aims to retrieve fuel debris from the unstable state and to remain the stable condition
without spreading radioactive materials.

Purpose

The technology for upgrading the approach and the system for retrieving the fuel debris and internal structures has safety issues;ensuring the confinement functions, capturing and removing the dust generated from fuel debris, and the monitoring of alphanuclides(collective terms for the radioactive nuclides that release alpha rays). This project aims at developing technology to solve these issues
and to optimize the methods and systems for ensuring safety.

Major Approach and Results

1 R&D for Confinement Functions⃝ As the element technology development for securing the confinement function by differential pressure management, analytical methods

that can predict airflow distribution , dust spreading prediction (Fig. 1), and hydrogen local accumulation were almost established.
Additionally, the differential pressure conditions which can secure the confinement of radioactive dust were set by element test.
⃝ A connection method for the Reactor Building (R/B) outside a work cell and PCV with the Reactor Building (R/B) access tunnel outside
a work cell and PCV with the access tunnel having airtight and shielding functions by the partial submersion side accesse method (Fig.2)
was studied, and element test (Fig.3) that simulated the connection to PCV with the access tunnel delivered from the R/B outside wasconducted to confirm the feasibility.

2 Technology Development for Collection/ Removal of Dust Derived from Fuel Debris⃝ Element tests for t echnology development of dust collection and removal in the gas phase and liquid phase systems (Fig. 4 and Fig. 5)

were performed for the collection and removal methods which have items to be confirmed, and thereby, comparability data that fluid
properties containing dust will be available was obtained.
⃝ Assuming the possibility of alpha-nuclides dissolving into water, element tests for technology development of solubility nuclides
collection and removal in the liquid phase system were conducted to confirm collection and removal methods. As a result, camparability
data that alpha-nuclide behavior of inorganice wate quality will be available was obtained.

3 Study on Alpha-Nuclide Monitoring System Associated with Fuel Debris Retrieval⃝ The necessity and the purpose of alpha-nuclides monitoring technology during fuel debris retrieval were updated since last fiscal year, as

well as the required measurement range for the gas phase and liquid phase systems were clarified.

⃝ Investigation for the existing technologies of the gas phase and liquid phase systems alpha-nuclides monitoring was performed, andissues of feasible applications were clarified.

4 Optimization Study on Ensuring Safety of Methods and Systems⃝ The methods was updated focusing on the partial submersion side access method since last fiscal year.
⃝ The safety and function requirements for fuel debris retrieval based on the defense in depth was reclarified, and the monitoring

parameters required for the installation was clarified.

⃝ The radiation exposure evaluation method for publics and workers during fuel debris retrieval were studied and the trial test calculation

was performed to expand the requirement items for the equipment design.

⃝ Prerequisites for insufficient information shortage of the safety system required for fuel debris retrieval were set and the required

system configurations were established under the conservative conditions.

Future Development

This project will promote effective and reasonable work for fuel debris retrieval and the reactor internals and concretize the best retrieval
method. It will be necessary to undertake a detailed study of the safety system established in FY2018 for feasible applications based on
on-site situations, and therefore, the operator of the NPS will take over the responsibility of engineering.
As for part of the confinement functions, and collection and removal technology of the liquid phase system, development of applicabletechnology is continuously planned.

14

Annual Research Report 2018

Subsidized Projects

Dust injection
Termination

LegendFracture Nitrogencross section injection

Dust
concentration

Dust amount [Kg]Aerosol injection

Integrated injected dust amountIntegrated exhausted dust amountDust
injection
start

Dust amount in PCV
Integrated leaking dust amount

Exhaustport

Analysis hours [hr]t = 5 hours

t = 10 hours
Equiv. Unit 3: Dust
concentration distributiont = 15 hours
4 hours aftertermination of cutting

Fig. 1: Example of Technology Development of Confinement Functions(Dust spreading prediction)
R/B WallAccess Tunnel
R/B Building
Temporary
Building

Let-off
Biological Shielding WallＢＢＳＷＳＷ(BSW)
Opening

PCV
Opening

R/B Opening
口Access口Tunnel

R/B WallFig. 2: Access Tunnel Installation Method

Fig. 3: Access Tunnel Element Test

Back washingpressure direction

HEPA Filter
Installation Area

Air Flow
Air Flow

Filter Connecting
Flector
Filter Moving
CylinderAir Flow
Filter
Installation Area

Dust Collection
Area

Access Tunnel

Air Flow

Filter Moving
CylinderCylinder Operation
CylinderFig. 5: Example of Collection and Removal Test
Testing Device for Remote replacement HEPA
Fig. 4: Example of Collection and Removal Test
Back washing Filter Testing Device

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Upgrading of Approach and Systems for Retrieval of Fuel Debris and Internal Structures(Development of Technology for Establishment of Criticality Control Method)
Background
It is assumed that fuel debris has not currently reached criticality, while a criticality control method has been developed to ensure thecriticality prevention despite change of the shape and water amount of fuel debris during fuel debris retrieval work in future, and to safely
terminate criticality even if criticality occures.

PurposeThis project aims to confirm the feasibility of each element technology based on the site situation, study the application of retrieval
device and systems, and optimize the criticality control method from the aspect of ensuring safey for the entire retrieval work based on thesite situation.

Major Approach and Results

1 Technology Development for Subcriticality Measurement and Criticality Approach Monitoring
The development of the subcriticality estimation approach method based on neutron measurement is promoted. In the current fiscal
year, various experimental systems composed of two reactor cores were configured by the Kyoto University Critical Assembly (KUCA) to
confirm the feasibility of subcriticality measurement technology in case of assuming a widely spread fuel debris system based on images
obtained from investigation inside primary containment vessel (PCV). (Fig. 1). As the result, it was confirmed that the change of local
subcriticality approach can be captured by neutron signal analysis (noise method).
In addition, for the application study of criticality monitoring removal device, the concept of nuetron detector to be installed on fueldebris (Fig.2) and cable handling methods were studied, and the basic plan and issues for a practical application of the device size,
weight and handling performance were clarified.

2 Development of Recriticality Detection Technology

Technology development for early detection of recriticality occurrence is promoted by upgrading of the radioactive gas control
system. In the current fiscal year, criticality event evaluation based on the operation condition of the negative pressure control system
during fuel debris retrieval was performed. It was confirmed that early criticality occurrence can be detected by monitoring Kr-88 which
responds faster than currently monitoerd Xe-135 (Fig. 3).
In addition, radioactive gases such as Kr-88 was generated by using neutron irradiation piles (Fig. 4) in the National PhysicalLaboratory (NPL) to reduce the radioactive concentration measurement errors of a gas radiation detector and to improve the criticality
detection accuracy, and calibration technology for a gas radiation detector was established.

3 Development of Criticality Prevention Technology

Development of water-soluble/non-soluble type of neutron absorbing materials to prevent criticality occurance. In this fiscal year, , along-term irradiation test was performed to evaluate the impact of insoluble absorbing materials (B4C metal sintered material, Gd2O3
particle) during fuel debris retrieval, and it was confirmed that the impact on hydrogen generation and water quality (Fig. 5) was small.
In addition, the fuel debris injection and crushing test (Fig. 6) was conducted to study the workability of insoluble absorbing materials
and fuel debris, and the conceptual study of absorbing material injection device was performed for on-site application.

4 Optimization Study of Ensuring Safety in Methods and Systems (for Criticality Control)

Based on the safety and the site applicability, the most applicable criticality control method have been studied. As the result ofinvestigation inside PCV and a practical criticality risk assessment by statistical assessment, there is no significant change in criticality
risks, and the tendency of criticality risk increase was not seen from the past assessment.
In addition, fuel debris retrieval plan including the criticality control method for the fuel debris retrieval expansion phase and thedefense in depth of criticality control that is consistent with the safety concepts for the entire debris retrieval work were developed.

Future Development

On the basis of the feasibility study of criticality control technology and the study results of practical application of the retrieval device
and systems, study on the validation test and the site applicability will be performed, aiming at the completion of element technology forcriticality prevention and monitoring.

16

Annual Research Report 2018

Reactor core-2

Reflector
regionVoidregionWeak nuclear bond system

Effective multiplication
factor measured value

Reactor core-1

Polyethylene
Lead shield

0.9 Strong nuclear bondsystem0.8

Weak nuclear bond system

0.7

KUCA Test System Series

Subsidized Projects

1.0

Controlrod

Neutron detector

0.7
0.8
0.9
1.0
Reference solution of effective
multiplication factor

Neutron detector
Polyethylene

Fig. 1: Subcriticality Measurement Test and Measured Resultsof Kyoto University Critical Assembly (KUCA)

Fig. 2: Design Concept of
Neutron Detection Unit

In case of largely expanded fuel debris, there is a possibility of having strong
and weak bound atomic nucleus in the regions. Even in these cases, it was
confirmed that the degree of local subcriticality can be estimated.

A neutron shielding plan is being studied toshield Gamma-ray with lead and neutron in
the other area with wrapping polyethylene
for around a neutron detector.

Kr-88
detected time

1.E+03 (concentration:

100 times)

1.E+02

Xe-135detected time(concentration:
1Bq/cm3)

1.E+01NBR of nuclear fission[fission/sec]<Evaluation Result>• Increase of Kr is faster thanXe. Detection of criticality
sign is approximately 5
hours, in half the time ofnormal case.

Nuclear fission amount

1.E+04[-]
Initial concentration[-]<Assumed Events>
The event being criticality
due to accumulated fine
powder when retrieving fueldebris.
• Injection Reactivity: 5 x 10-3$/h• Outer Flow: 21.8m3/h

1.E+00

0

2

4

6

8

10

12

Time [h]
[h]

pH
pH

Fig. 3: Evaluation Result of Criticality Detection
Performance in case of Criticality Event (Example of
Minor Criticality due to Fine Powder Accumulation)

13
12
11
10
9
8
7
6
5
4
3
2
1

Solid

Solidified
substance

Fig. 4: Neutron Irradiation Pilein the NPL, UK

Fuel debris working
tool: Chisel

B4C sintered metal material

B·Gd contained Glass
B4C reagent(comparison material)
Gd2O3 particle
Water glass/ Gd2O3
Granulated powder body

B4D/ SUS
B and CD
Sinteredcontainedmaterial glass material

pH region with less corrosion
influence: 5 ~ 9
0

10

100

1000
10000
Irradiance level(kGy)
kGy

100000Fig. 5: Test Result for Water Quality Impact
Evaluation by Long-term Irradiation of
Insoluble Absorbing Materials

Surfacebefore test
CrushedAbsorbent
simulatedmaterial
fuel debris

Simulatedfuel debris(a)Underwater chisel work test
(Lower part of pedestal: simulated)

Absorbent material
dispersion after work(b) Upper surface
appearance before/
after test

Fig. 6: Insoluble Absorbing Material
Injection and Crushing Test for
Fuel Debris

This is an experiment test to investigate how fueldebris and insoluble neutron absorbing materials
are mixed or unevenly distributed during fuel debris
processing.

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

Upgrading of Fundamental Technology for Retrieval of Fuel Debris and Internal Structure
BackgroundA fuel debris retrieval policy was determined and thereby the partial-submersion side access method is focused on. This project aims toacquire the necessary data and information by performing element tests and a conceptual study to evaluate the feasibility of the retrieval
method.

PurposeThe subsidized project of the “Upgrading of Approach and System for Retrieval of Fuel Debris and Internal Structures” has studied
theoretical retrieval methods. This project aims at increasing the feasibility of the project not only through theoretical study but also throughobtaining the relevant data through element tests. In addition, the development plan will be formulated after identifying issues.

Major Approach and Results

1 Development of Technology for Prevention of Fuel Debris Spreading① Development of Fuel Debris Collection System⃝ Assuming the fuel debris collection process for the bottom of Primary Containment Vessel (PCV) and the technology information on thecollection system of powder-shaped fuel debris was clarified.
⃝ As to processed fuel debris and pebble-like fuel debris, the technology information on an effective collection work tool was clarified.
② Development of Fuel Debris Cutting/ Dust Collection System⃝ Fuel debris to be cut and collected was clarified with respect to where and what shape of the fuel debris would be, and the most effective
processing method and the collection method.
⃝ A workability test for chisel processing and ultrasonic core boring against the MCCI (Molten Core Concrete Interaction) products was
conducted to confirm processing properties and their speed.
⃝ The components of the MCCI products and the trial production method as simulated fuel debris were studied. The trial product wasprocessed by a processing test to analyze processed waste liquid, and thereby the grain size data was obtained.
③ Development of Prevention Method of Fuel Debris Spreading⃝ To prevent fuel debris at the PCV bottom spreading into the vent pipes, the suppression chamber (S/C) and other area associated with
fuel debris retrieval work, element tests were conducted assuming the weir installation in PCV, which provided the feasible prospect of
the remote installation.

2 Development of Element Technology for Retrieval Device Installation① Development of Element Technology for Work Cells⃝ Technology for the cell confinement and connecting to the PCVs were compared and clarified.
⃝ The element test for inflate seals which can seal for cells and PCV connections was conducted, thereby confirming the feasibility of work
steps and extracting issues. (Fig. 1)② Development of Interference Object Removal during Fuel Debris Retrieval⃝ Interference objects are required to be removed until fuel debris at the PCV bottom is reached by the partial submersion side accessmethod, and thefore processing methods were clarified.
⃝ The element test for the interfering object removal was performed based on the clarification results of the interfering objects, and thefeasibility of the work steps was confirmed and issues were extracted. (Fig. 2)⃝ The element test for removing interfering objects at the reactor bottom by the partial submersion side access method was performed,
and basic cutting means of the CRD (Control Rod Driving Mechanism) housing and the fragment collection method were confirmed.
(Fig. 3)
⃝ The element test of interfering objects removal for inside and outside the pedestal by the partial submersion side access method was
performed, and the feasible prospect was obtained for a basic cutting and collection work at narrow parts. (Fig. 4)⃝ The combination element test of the robot arm and the access rail was conducted to extract issues, and the feasibility for basic actionof the removal of fuel debris and interfering objects were confirmed. (Fig. 5)

3 Development of Remote Maintenance Technology for Fuel Debris Retrieval Device⃝ Basic concepts of remote maintenance, for example equipment in the cell of the partial submersion side access method, were studied toclarify the maintenance classification and facility.

Future Development⃝ Based on issues extracted by the element tests including a simulated fuel debris processing test and interfering object removal, adevelopment plan will be established to contribute the feasibility study of the method.

⃝ Fuel debris retrieval work will be promoted in a reasonable manner which is necessary for effective methods, and the study of the mostfeasible method will be promoted.

⃝ Considering the transportation and storage method for the retrieved fuel debris and wastes, the feasible carry-out process of fuel debris andwastes will be proceeded in the cooperation with relevant other projects.

18

Annual Research Report 2018

Subsidized Projects

Inflate seals

Collection Device
CRD housing

Manipulator
Fig. 3: Partial Submersion
Side Access Method:
Element Test of Interfering
Object Removal

Fig. 1: Partial Submersion
Side Access Method:
Inflate Seal Element Test

Core Boring

BSW

Pedestal Opening

Device Arm Part

Fig. 2: Partial Submersion Side Access Method:
Element Test of Biological Shielding Wall
(BSW) Removal
Access RailFig. 4: Partial Submersion
Side Access Process Method:
Element Test of Interfering
Object Removal
Robot Arm

Fig. 5: Partial Submersion Side Access Method:
Element Test for Combination of Robot Arm andAccess Rail19

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

 evelopment of Fundamental Technology for Retrieval of FuelD
Debris and Internal Structure(Development of Small Neutron Detectors)
Background
Retrieving fuel debris necessitates identifying the location and amount of fuel debris. One of the method for identifying fuel debris is theneutron measurement method that is to measure neutrons generated by spontaneous nuclear fissions in the fuel debris. However, existing
sensors may not be capable of being used due to the dimensional restrictions, thus making a smaller sensor necessary.

PurposeFor fuel debris detection, it is necessary to install a detector in the vicinity of fuel debris and measure weak neutrons under highgamma ray dose. The access route to the vicinity of fuel debris is narrow, therefore a compact type sensor is required. This project aims at
developing a small neutron detector to satisfy those requirements.

Major Approach and Results

1 Specifying the Feasible Neutron Detection Technology for Fuel Debris Retrieval and Results ofElement Tests for Feasibility Investigation (Phase-1) for Fukushima Daiichi NPS

Element test of the CMOS type neutron detector was performed to confirm the neutron flux detection at a single neutron location,
gamma-ray detection at a single gamma-ray location, and detection performance under the complex irradiation environment of
gamma-ray and neutron ray. The identification of cluster patterns using alpha-rays derived from neutrons in the combined location
was confirmed (The Fig. 1 principle was confirmed.). It was also confirmed that there is no false detection of neutrons until gamma-ray
cumulative dose reaches up to the degree of 1000Gy.

2 Development of Neutron Detector (Phase-2)

A prototype of the small neutron detector was produced based on the study results of Phase-1. The study of the sensor layout, thedesign of the sensor unit, and the detection software production and the performance validation were performed.
The sensor unit was designed considering the sensitivity secured by stacking three sensors and heat radiation performance (Fig. 2).
In addition, the neutron detection software was installed into the PC, which is transmitted images and comprised the display of neutron
counts in real-time.
As the result of the evaluation test, the following performance were confirmed (Fig. 3 and 4). 1. The neutron detection sensitivity wasable to detect 2.3 neutrons per hour in average after 7 hour measurement against the requirements that can count neutrons within 1
hour for the neutron flux on 0.1n/ (cm2·s). 2. It was confirmed that the measuring range of the neutron flux can measure more than
0.1n/ (cm2·s) and less than 1,000n/ (cm2·s). 3. Neutrons can be discriminated up to the accumulated dose of 1000Gy as radiation
resistance. 4. The detection capability was the same as normal temperature even under the environment of 40°C as heat resistance. 5.
The water resistance capability was confirmed as it was able to operate even under water without water immersion.
In addition, the conclusions below were obtained from the test assumed practical application; 1. The study on the shielding structureusing neutron absorbing materials is necessary if the neutron source is located around the area of use (within 200 ~ 300mm) inunderwater environment. 2. A neutron moderator installed around the CMOS sensor is necessary for the promotion of thermalization
because neutrons are not thermalized in the aerial environment. For the effective nuetron detection, the change of the collimator shapeand the moderator size in the collimator depending on the purpose of use are assumed to enhance the feasibility.
Requirement specifications of this sensor differ depending on the application destination, therefore, the sensor requires to meet thedesign and production.

Future Development

A small neutron detector completed in this study is smaller than a conventional neutron detector,therefore, this project aims at promoting
the application to utilize the profits. Specifically, the application of the detector will be promoted for on-going projects of decommissioningof the Fukushima Daiichi NPS including the project of investigation inside PCV and the monitoring technology for fuel debris retrieval. In
addition, a wide range of matching needs will be promoted because these technologies are possible to be utilized in general industries andmedical industries other than nuclear decommissioning.

20

Annual Research Report 2018

Subsidized Projects

CasingNeutron rays
Reaction
LayersElectric charge

Gamma rays

Flexible PCB
CMOS
sensorAlpha rays

Data collection/transmission

CMOS layers

Connection cable

Sensor unit

When neutrons enter into the reaction layers appliedon the CMOS layers, heavy-charged particles are
generated and charged on the CMOS layers as charge,
and bright patterns are displayed. On the other hand,since gamma-ray is displayed relatively small patterns,
this is discriminated from neutron entering.

Coaxial cable

Fig. 1: Principle of CMOS Type Neutron Detector

PC
Software,etc.

Sensorappearance

Receiving/ Signal
processing partPower
supplyInsert ray
sources

HDPE(High-Density
Polyethylene)
Block(Pu/Beor Cf-252)
Directionof Co-60

Acrylic plate

Co-60
Neutron-ray source (Pu/Be) + moderator,(Gamma-ray source)

NEUTRON COUNTING RATE (CPS)

Fig. 2: Overview of CMOS Type Neutron
Detection System[Test Conditions]• Neutron Flux: 0.1 ~ 737 n/(cm2·S)• Gamma-ray dosage rate: 0 ~1028Gy/h• Irradiation time: 10 ~ 60 minutes• Ambient temperature: 20°C

NEUTRON FLUX (N/(CM2·S))

CMOS sensor

Fig. 4: Neutron Flux Detection Performance(under radiation environment)
Table

Cable

Fig. 3: Mockup System of Element Test

21

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Preparation of Fuel Debris Retrieval

 evelopment of Technology for Collection, Transfer and Storage ofD
Fuel Debris
Background
According to the Mid-and-Long-Term Roadmap plan toward the Decommissioning of Fukushima Daiichi Nuclear Power Station (NPS)of Tokyo Electric Power Company (TEPCO) Holdings Co., Inc., fuel debris will be retrieved from the primary containment vessel (PCV) andthen stored in the storage building of the Fukushima Daiichi NPS until the treatment and disposal methods are determined. Therefore, it isnecessary to establish the fuel debris collection, transfer and storage system.

PurposeBased on the experience of the Three Mile Island Nuclear Power Generation Station Unit 2 (TMI-2) in the United States, and proventechnologies for transport and storage of spent fuel, this project aims at development of fuel debris canister (hereinafter referred toas “canister”) and the canister handling device for safe and efficient collection, transfer, and storage. The project mainly develped theappropriate specifications, and the shape of the canister and its handling device for the fuel debris retrieval method in FY2018.

Major Approach and Results

1 Investigation and Research Planning for Transfer and Storage
As for the safety evaluation and validation of the canister for Fukushima Daiichi NPS, the project team exchanged information withoverseas engineers having decommissioning experiences and knowledge, and thereby, confirmed the applicability of basic specifications
for the canister.

2 Study on Safety Requirements, Specifications and Storage System for Transfer and Storage

for Fuel Debris Canister

A process flow plan and the throughput (including the required amount of wastes, the number of canisters, and the storage area) for
fuel debris retrieval, collection, transfer and storage (including drying treatment) were updated, while requirements from the projectsof the fuel debris retrieval, and waste disposal and treatment, were clarified to revise the scenario. In addition, based on the scope ofthe safety function of canister and the building facility, the function scope was refined. Furthermore, , the specification plan for drying
facility design was provided from the aspect of hydrogen gas generation reduction.

3 Development of Safety Evaluation Method and Safety Validation

Unlike the TMI-2 case, fuel debris and the Molten Core Concrete Interaction (MCCI) possibly contain sea water components in the
Fukushima Daiichi NPS. For safety collection, transfer and storage of fuel debris and the MCCI products, it is required to verify the safety
for criticality, the structures, material deteriorations and hydrogen generation.
Various safety validation study and tests for the fuel debris canister were performed. For example, the canister lid structure(feasibility verification test considering a handling flow and the safety requirements (Fig. 1 and 2), and hydrogen gas management(measurement test for hydrogen generation amount using spent fuel (Fig. 3 and 4), and efficiency evaluation of hydrogen recombination
catalyzer) were studied, and thereby, necessary knowledge was acquired to establish the safety evaluation method.

4 Study on Fuel Debris Storing Type

Based on study of the safety conditions and the safety validation results for the basic specifications of fuel debris canister indicatedby above Item 2 and 3, the storage methods were studied considering the properties and the shape of various type fuel debris; lumplike, granular and powderly shaped fuel debris (Fig. 5). As the results, the study revealed that the canister specification plan is currently
applicable to the throughput, and fuel debris properties and the shape.

Future Development

A verification test for the canister structure will be planned and conducted for evaluation of storage technology. In addition, the transfer
conditions of the canister will be clarified by the hydrogen measurement method and its management. Furthermore, technology for drying
treatment and pre-storage treatment evaluation will be studied development of drying technology and its system.

22

Annual Research Report 2018

Rollover Jig (Bar)

Subsidized Projects

Suspension

Lid side

Simulated
Lid

Collision Surface(Floor surface)

2m

Collision Surface(Floor surface)

Rollover Jig (Bar)Supporting Point

Fig. 1: Appearance of Test Device

Main-bodyside

Fig. 2: Example of Simulated Lid after Test

• The conditions of the feasibility test for the canister lid structure (Canister rollover, Inner diameter ofcanister: 220mm)• For confirming the feasibility of the canister lid structure, the load test simulating events which may havea large impact on the structure strength evaluation was conducted. As a result, confinement performanceof the lid structure was confirmed when shock loads were applied..
Compound Gauge

Micro-manometer

Lead block (for shielding)(Pressure transducer piping)

Safety valve
Pressure transducer
piping

Thermocouple
conducting wire

Test container

Fig. 3: Appearance of Test Device
Inside Cell Outside Cell

*1

Barometer,
Thermometer

Hydrogen gas sensor
Thermocouple

Gas canister

Atmospherictemperaturein cellpressure in cell

Test container-1

Display,
Hydrogen gas sensor

Test container-2*2*3

Lead block(shield)
Test container-3

Vacuum pump

Test container-4

Data logger(Data collecting for temperature/pressure in a test container
and temperature/ atmosphericpressure in a cell)*1 Used by changing the connection at leakage check*2 Used by changing the connection at gas collection*3 Pressure transducer piping: Removed at lid open/close

Fig. 4: Test Device Configuration• For confirming the impact on alpha-ray against hydrogen generation
amount, a measurement test for hydrogen generation amount using
spent fuel as alpha-ray source will be performed, and the obtained
test data will be reflected the hydrogen generation amount evaluation.
strength evaluation was performed to confirm the feasibility and theconfinement performance of the lid structure was confirmed whenshock loads were applied.

(1) Assembled
canister tunk
(Approx. 2mlength)(2) Canister tunk(3) Lid of
canisterFig. 5: Basic Plan of Canister Shape• This is a connection method by rotating the canistertunk and its lid.
• Based on the safety validation results of various studies
and confirmation tests for the canister design, the basic
shape of the fuel debris canister was studied.

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Treatment and Disposal of Radioactive Waste

R&D for Treatment and Disposal of Solid Radioactive Waste
BackgroundAs fuel debris retireval work proceeds, a large amount of wastes will be generated in the Fukushima Daiichi Nuclear Power Station (NPS)
while fallen trees and rubbles have already been stored. Therefore, it is necessary to develop waste management methods of storage,
treatment and disposal and to implement the project while clarifying the properties of the wastes.

PurposeThis project aims at developing technology for the safe treatment and disposal of the solid waste generated in the accident byintegrating the results of research and development (R&D); characterization of inventory evaluation based on waste analysis, and study on
solidification technology the stabilization of secondary wastes generated by water treatment, and concepts of the treatment and the safety
evaluation method.

Major Approach and Results

1 Waste Characterization
The analysis of rubbles, the secondary waste generated by water treatment, and contaminated water continues to be performed.
The analysis revealed that local contamination of rubble samples is distributed unevenly (Fig. 1). In addition, the data repository withaccumulated analysis data was released on website to contribute the decommissioning (Fig. 2). The amount of radioactive wastes wasestimated that the distribution of contamination frequency caused by radioactive nuclides was the normal logarithmic distribution, andthereby, the estimation method considering the increase of analysis data was studied (Fig. 3). The radioactivity inventory of the wastes
by using the improved evaluation approach was calculated and a computation tool was also improved.

2 Preliminary Management① Applicability Evaluation of In-Drum Type Glass Solidification Technology
As part of the stabilization technology required for the preliminary management of the secondary waste generated by water
treatment, in-drum type glass solidification technology was focused upon, and the solidification (crucible test and engineering scaletest) that melted simultaneously with zeolite used for the contaminated water treatment and other simulated waste from the secondary
waste generated by water treatment was produced (Fig. 4). Based on the results of the glass composition study and a crucible melting
test, an engineering scale test under the manufacturing conditions was conducted to evaluate solidification properties, the nuclidetransition to the off-gas system and chemical durability of solidification.
② Study of Storage Measures for High-dose Waste (Hydrogen Generation Measures)
Differences and reasons between Japan and overseas in the concepts , the evaluation methods nd countermeasures of hydrogengeneration and specifications of canister were investigated, which provided the knowledge of the hydrogen gas generation evaluation
method and requirements for vents. In addition, as for the slurry-shape waste and the reactor internals (metal wastes), a series ofhydrogen generation measures for storage, process and disposal were studied to clarify the evaluation method and issues of hydrogengas generation measures.

3 Study of Disposal Concepts and Safety Evaluation Methods for Solid Wastes

Focused on case studies of overseas repositories, solid wastes in the Fukushima Daiichi NPS were clarified about their applicability
and issues. In addition, the methods of the disposal concepts considering properties of the wastes were examined (Fig. 5). The casestudy of multiple disposal concepts was performed with this method. It was confirmed that a policy of the waste treatment and disposal
can be studied depending on the waste properties.

4 Integration of R&D Results

The waste stream was reflected in the latest results obtained from the existing research. The establishment of an integration method
started for the progress, consistency of results and remaining issues, and the operational issues were clarified through trial operations.

Future Development

The project will continuously perform analysis on waste characterizations, as well as the sample collection, efficient analysis andimprovement of the inventory estimation method. In addition, in preparation for high-dose waste segregation and storage methods, will be
studied. Advanced disposal technology will be continuously selected for the stable treatment of the secondary wastes generated by water
treatment. The wastes will be estimated more specifically and multiple disposal methods will be studied.

24

Annual Research Report 2018

Low

Radiation
exposurerate

Surface contamination

High

Contamination after
grinding approx. 2mm

Fig. 1: Appearance of Locally Inhomogeneous
Contaminated Rubbles(by imaging plate measurement)
Probability (Curve)
Frequency (Bar graph, in arbitral unit)

Subsidized Projects

Appearance of test
sampleFig. 2: Released Analysis Datasheet

Improvement of Analysis Data
The uncertainty of normal logarithmic
distribution parameters (average anddeviation) is decreased, which shows thevalidity of the assumptions.
Transfer Ratio(relative concentration against OS in logarithmic scale)Fig. 3: Example of the Uncertainty Decrease of Distribution Parameters Associatedwith Analysis Data Increase

Solidified material (crucible test)
Upper surface

Engineering Scale Test Device

Section A-A

Solidified material (Engineering Scale Test)

Fig. 4: Test Facility of In-Drum Type Glass Solidification Technology
and Solidified Materials
1F Waste Information[Features]• Massive wastes• Wide variety of wastes• Existence of
uncertainty, etc.

Study of Disposal Concept• Disposal depth• Disposal facility structure• Environmental
conditions, etc.

Waste Solidification Plan
• Measurement/ Assortment• Solidification Treatment• Disposal container, etc.

• Study on a disposal policy that issuitable to the features of wastes
in the Fukushima Daiichi NPS byrepeated-study cycle of the “Waste
solidification plan,” “Disposal
concept study,” and “Repositories
capability evaluatio

Repositories capability evaluation• Evaluation scenario• Evaluation model• Evaluation parameters, etc.

Fig. 5: Study Process of Disposal Concepts Considering Waste Properties

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Major Research Achievements - Subsidized Projects

RE P ORT

2 0 1 8R&D for Treatment and Disposal of Radioactive Waste

R&D for Treatment and Disposal of Solid Radioactive Waste(R&D on Preceding Processing Methods and Analytical Methods)

BackgroundIn the Fukushima Daiichi Nuclear Power Station (NPS), multiple and various kinds of wastes have been already generated, and will increase
for a long term in the future. To proceed with the decommissioning, it is necessary to manage the generated wastes in a safe and appropriatemanner, while studying the waste treatment and disposal methods and ensuring the consistency.

PurposeThis project aims at developing technologies for safe treatment and disposal of the solid wastes generated in the accident by the
integration of results of research and development (R&D); Characterizations of inventory evaluation based on waste analysis, study
on solidification technology, stabilization of the secondary wastes generated by water treatment, the concepts of wastes and the safety
evaluation method.

Major Approach and Results

1 Prospect Evaluation of Proven Technology for Solid Waste Treatment① Engineering Scale of Applicability Evaluation on Element Technology for Solid Waste Disposal
This project aims at acquiring information on four treatment technologies (High temperature: Glass solidification and melting, Low
temperature; Cement solidification, AAM: (Alkali Activated Materials), solidification) proven both in Japan and overseas, washing outof the selection index (evaluation axis) of applicable treatment technology and clarification of qualitative/ quantitative information(Table 1). Furthermore, necessary information was clarified. In addition, the insufficient data mainly on the low temperature treatment
was obtained to complement quantitative information. Properties of the AAM solidiication were discovered; early strength developing
property in cement (Fig. 1), the possibility of superior dissolution stability and differences of the setting time depending on composition(Fig. 2).
② Clarification Items Having Impacts on Disposal Safety Evaluation and Study on Analysis Evaluation Method for

Disposal Affects Substances

The project extracted six substances (organic materials, sea water components, boronic acid, ferrocyanide, sulfate and carbonate)since they should be confirmed as impact substances, or they are more likely to have impact on the safety disposal evaluation basedon the information on nuclide adsorbing behavior and wastes. The insufficient information about the interaction with nuclides was
acquired and investigated, provided the quantitative evaluation methods for the influence against nuclide adsorbing behavior dependingon the amount of available data. A trial of the adsorption reducing coefficient evaluation (Fig.3)was performed to extract issues.

2 Development of Technology for Solid Waste Storage and Management① Development of Contamination Evaluation Technology for Solid Waste Segregation
As for alpha contamination measurement technology, the parameters impacting the measurement of environmental temperatureand radiation dose rate were clarified, assuming the site environment, the requirement specifications of the measurement device were
established based on the measurement needs. Additionally, based on the required specifications, the test device (Fig.4) added enlarged
lens and temperature adjustment mechanism were produced, and element tests were conducted to clarify issues for the site application.
The scope of the site application for the penetration alpha contamination measurement technology was studied and the measurement
needs were investigated, and thereby, the technological investigation was performed aiming at using the technology under the assumed
environmental conditions. Moreover, the measurement method performance of the gamma nuclides penetration depth including Cs-137
which is considered to accompany alpha nuclides (Fig. 5) was evaluated.
② Study of Simple and Rapid Analysis Method
The analysis method for a massive and wide dose rates of waste samples was developed for the safe, and effective analysis while
ensuring the reliability of analysis values. Specifically, the methodology with least sample collection was established to ensure
appropriate samples. Furthermore, the analysis process (Fig. 6) that was installed a mass analysis to the minimize nuclide separation
processes and to improve a detection lower limit was established. In addition, the automation technology of seven-step operations(from the melting phase to the ion exchange/solid-phase extractions) which is the most complicated analysis operation was developed,
and confirmed that the system has the same accuracy as skilled workers. The specific methods including the analytical accuracy
management to secure the reliability of acquired data and to maintain analyst’s skill were systematically developed.

Future Development

This project will perform additional investigations and tests for the insufficient information and data about the applicability evaluationof processing technology and the impact evaluation with toxic substance in parallel with other R&Ds. In addition, technology forcontamination evaluation and the analysis methods were established, and technology and management verification of these methods will
be promoted for the site application.

26

Annual Research Report 2018

Verification test of small device in Japan(30L/ batch)• Number of process 14
Process • Treatment temperature 1450°C
performance • Treatment capacity 0.4t/ batch/ d• CS volatile rate
CS 10 ~ 70%• Waste composition (Ca concentration,
impurity), waste moisture content,• High impact
molten metal temperature
Operability parameterSolidified substance performance•• Major process linkdegradation due to internal pressure riseand insufficient melting• Major facility• Drying facility, high-frequency power
Economic configuration
source, water-cooled coil, off-gas system
Efficiency • Generated• Bug • HEPA filter, residue and exhaustsecondary waste
water scrubber
olidified substance 200L drum
Solidified • Sdimensions
Substance • Leach rate
10-4kg/m2/d
No hydrogen generation
Products • G-value

←

Slow

Starting Time (min)●Metakaolin 100%Concretion

Slag 20%

3 7 28 91

3 7 28 91

Slag 40%OPC※

3 7 28 91

3 7 28 91

Portland Cement Association

Fig. 1: Relationship between Compressive
Strength and Transit Time
1E+1

Adsorption of ferrocyanide
coexistence system tobentonite

1E+0

pH4
pH8
pH10

pH4-ref.

1E-1

pH8-ref.
pH10-ref.

1E-2

pH4-ave.
pH8-ave.

slag fine◆Blast-furnace
powder 40% kneading

1E-3
1E-4

1E-3

1E-2

Power
of
(pH10)
Number(pH 10)

1E-1

Na4[Fe(CN)6] concentration [mol/L]■Cement (OPC)

fast

Metakaolin only

Curing days

slag fine▲Blast-furnace
powder 20% kneading→

Flash setting

AAM

Compressive strength(N/mm2)(Qualitative/ Quantitative Information)

Kd [m3/kg](Evaluation Axis)
Technological Achievements

Subsidized Projects

Table 1: Major Evaluation Axis and Clarified Information(Example: Superhigh Frequency Molten Solidification)

Fig. 3: Evaluation Example of Adsorption
Decreasing Factor for Artificial Barriers

Na/Si mol ratio in solidified substance(mol/mol)

Fig. 2: Relationship between Composition
and Concretion Time
Accepted Object
Blank Burst

Camera with
distance
sensorSample Burst

Combined Alphacontamination
distribution on
camera image

Alpha-ray signal(UV) receiving partFig. 4: Surface Alpha Contamination
Measurement Device (Alpha Camera)
x-y-z Distribution
Measuring screen

Smash

Drying/ Heating/ Collection

Distilling Distilling Sample ADJ
H-3

LSC

C-14
LSC

I-129

ICP-MS

Alkali Melting

Cs-137HPGe

LSC(or LBC)

Aid Digestion

Adsorptive Adsorptive Adsorptive
Separation Separation Separation

Separation

Sediment
Production

EuCo-60 Nb-94 152,154

ICP-MS

Sediment
Production

Separation

Separation

LSC

Separated Separated Separated
Production Production Production

HPGe HPGe HPGe

Cs-135Ca-41

LSC(or Ge-LEPS)

Sediment
Melting

Sediment
Melting

Ni-63
LSC

Cl-36

Sr-90

Separation
Separation
Se-79
Tc-99

ICP-MS

Separation

Zr-93 Mo-93

ICP-MS ICP-MS

Dilution, etc.
Pd-107
Sm-151
Sn-126
ICP-MS

Ni-59

ICP-MS(or Ge-LEPS)

Fig. 6: Developed Analysis Process
Detector appearance

Fig. 5: Example of Investigation Results of Penetration
Alpha Contamination Measurement Technology:
Detector appearance and image of measured results
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Major Research Achievements - In-House Research

RE PORT2 0 1 8Investigation Research on Deposits at the Lower Part of Pedestal in
Fukushima Daiichi NPS Unit 2

Background and Purpose
In the Fukushima Daiichi Nuclear Power Station (NPS) Unit 2, the investigation inside the primary containment vessel (PCV) revealed that
deposits exist at the bottom of the pedestal. This study aims to confirm the changes of the deposits behavior in contact with the deposits at thelower pedestal based on development of the investigation device for the PCV internals.

Major Approach and Results

The investigation unit that was used for investigation inside the PCV was changed in
January 2018, and confirmed the state change of the deposits by applying a mechanicalforce of the finger structure action. This investigation revealed that pebblelike deposits
at the pedestal bottom was movable. In addition, it was confirmed that the deposits on
the platform can be gripped and movable.

Future Development

The useful information on fuel debris retrieval was obtained, therefore, this projectcontinues to study toward determining the fuel debris retrieval method for an initial unit.

Photo 1: Investigation Result ofPedestal Bottom(In contact with deposits)Test and Research for the Stable Operation of Advanced Liquid
Processing System and Slurry Stabilizing Facility

Background and Purpose
The feasibility of the stabilizing facility with pressurized press filtration (filter press) was confirmed under the subsidized project of
supplemental budgets of FY2013 and FY2014. This project aims at verifying the applicability of the treating range by testing for crushing,transfer and filling of dehydration products as well as the system obstructions.

Major Approach and Results

1 Applicability Test of Technology for Crushing and Transfer

of Dehydration Products

A dehydrate crushing test using proven wires in general industries was conducted,
which provided the prospects of application. In addition, a conveyer transfer test was
performed, and it revealed that issues remain in the operation side.

2 Applicability Test of Dehydrate Filling Technology

The bulk density and the scattering condition of the dehydration products after
discharge were confirmed, assuming that a container is filled with dehydration
products by free fall. In addition, as for the measurement of the water content rate on
the dehydration products, the applicability of two patterns calculated from remotemeasurement and process monitoring data was studied.

3 Applicability Verification for System Obstruction

Photo 1: Appearance of Crushing
Test by Simulated Dehydration
Products

The removal effect using a detergent was confirmed, assuming that the system wasobstructed due to slurry adhesion.

Future Development

The results of the project will be reflected into a basic design of the slurry stabilized treatment facility and develop the practicalapplication.

28

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Others - Datasheet

RE PORT2 0 1 8Major Research Results in FY2018
No.

Presented by

Date

1

Japan Robotics Society (Official English-Language Journal);
Advanced Robotics <Research paper >May 2018

Shape Changing Robot for Investigation Inside PCV named
PMORPH

3

Academic Network contributing to Fukushima
Reconstruction and Reactor Decommissioning (ANFURD)(Workshop on “Potential Issues on Fuel Debris Retrieval”)

May 15, 2018

Current Status of Technology for Fuel Debris Retrieval andExpectation to the ANFURD

2

5

6

May 11, 2018

IAEA Regional train trainer source on NDT for civil
infrastructure, Workshop

May 31, 2018

The 35th International Nuclear Air Cleaning Conferenceby International Society for Nuclear Air Treatment
Technologies

Jun. 4, 2018
Jun. 14 ~15, 2018

8

The 8th Robot Award

Jun. 29, 20189

The 10th Radiation Measurement Forum, Fukushima

Jul. 4, 201811

Japan Society of Maintenology; The 15th Academic
Conference

Jul. 10 ~12, 2018

International Conference on Nuclear Engineering(ICONE26)

Jul. 23, 2018

7

10

12
13
14
15
16
17

18
19

The 26th ATOX Technology Workshop

The 26th International Conference on Nuclear Engineering(ICONE 26)
The 56th Team Meeting on Countermeasures for
Decommissioning and Contaminated Water Treatment
Conference

Jul. 10, 2018

Jul. 22, 2018

Jul. 26, 2018

IEEJ (Institute of Electrical Engineering, Japan) Journal,
August Issue

Aug., 2018

IRID Symposium 2018

Aug. 2, 201824

The 3rd International Forum on the Decommissioning of
the Fukushima Daiichi Nuclear Power Station

Aug. 6, 201827

Subsidy Project on Decommissioning and Contaminated Water
Management in FY2017 Supplementary Budgets; Developmentof Technology for Investigation inside PCV in Unit 2-3 ofFukushima Daiichi NPS

Subsidy Project on Decommissioning and Contaminated Water
Management in FY2017 Supplementary Budgets; Developmentof Technology for Investigation inside PCV “InvestigationPCV”
Investigation inside the Pedestal for Unit 3”
Development of Debris Investigation Equipment at Fukushima
Daiichi NPSATOX Approach for National Project

Subsidy Project on Decommissioning and Contaminated Water
Management in FY2017 Supplementary Budgets; Development
and Results of Technology for Investigation inside PCV in Unit2-3of Fukushima Daiichi NPS
Example of R&D on criticality control technique of Fukushima
Daiichi Nuclear Power Plants
Development of ROV to investigate inside of primarycontainment vessel at Fukushima Daiichi Unit 3

Analytical Results of Waste Sample (rubbles, secondary wastes
generated from water treatment faciity contaminated water,
treated water, and soil)
Development of Investigation inside PCVTomography Technology inside the Reactor through Cosmic-ray
Muon

Status of IRID’s R&D

Status of IRID’s R&D; Conceptual Design to Enhance Safety andFeasibility, and Future Technological Development
Crawler Travel Test in narrow parts

Characterization of the large-scale MCCI test products for fueldebris removal from the Fukushima Daiichi Nuclear Power Plant
Progress of 1F PCV contaminant analysis by TEM observation
Development of a remote concrete sampling device“NDA System for Fuel Debris Characterization at Fukushima
Daiichi”

25
26

Development of Working Robot under High Radiation
Environment for the Decommissioning of Fukushima Daiichi NPSCore Material Melting and Relocation tests

21
23

R&D for Fuel Debris Retrieval at Fukushima Daiichi NPSApplication of ICP-MS to analysis of samples from 1F site atRadioactive Material Analysis and Research Facility

20
22

Development of Investigation Equipment for inside PCV (A2, A2’,
Underwater ROV) and Investigation Results under the Ultimate
Environment of Fukushima Daiichi NPS

DataSheet

The 23rd National Symposium on Power and Energy
Systems

Robot Development toward Decommissioning of Fukushima
Daiichi Nuclear Power Station (NPS)In-House Research

4

The 10th Anniversary Memorial General Meeting for
Tohoku and Hokkaido Branch Division Establishment, Japan
Society of Maintenology

Details

The 15th “Nuclear Power Technology” Summer Seminar,
Nuclear Power Group Meeting; Atomic Energy Society ofJapan

Development of a remote concrete sampling device
Aug. 9, 2018

Estimation of fuel Debris Distribution at Fukushima Daiichi NPS
Unit1-3(Report including Subsidy Project on Decommissioning andContaminated Water Management in FY2017 Supplementary
Budgets (Upgrading the Comprehensive Identification ofConditions inside Reactor.

29

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No.

Others - Datasheet

RE P ORT

2 0 1 8Presented byDateDevelopment of Canister for Fuel Debris;(11) Measurement of Hydrogen Generation Amount due to theWater Radioactive Degradation under Gamma-ray Irradiation(Part-2)(12) Analytical Evaluation of Hydrogen Generation Amountdue to the Water Radioactive Degradation under Gamma-ray
Irradiation (Part-2)(13) Study on Alpha-ray Effect by Hydrogen Generation Test
Using Spent Fuel

28

Study on the Element Analysis Method for Fuel Debris in
Fukushima Daiichi NPS

29

Results of “Development of Fuel Debris Criticality Control
Technology” in FY201730

Comparative study of Cs silicates properties between DFT
calculation and experimental data

31

32

2018 Fall Meeting, Atomic Energy Society of Japan

Sept. 10,2018(1) Development of the Inventory Evaluation Method for
Fukushima Daiichi Accident Wastes, (13) Study on the
Introduction of Statistical Approach to Analytical Estimation
Technology(2) Development of the Inventory Evaluation Method for
Fukushima Daiichi Accident Wastes, (14) The Effect on Nuclide
Adsorption to Zeolite due to change of Contaminated Water
Quality(3) Establishing Analysis Database of Radioactive Waste
Generated from the Fukushima Daiichi NPS Accident(5 topics among the presentations related to the Project of
Identifying inside the Reactor as below);
Analysis/ Evaluation of Identifying inside the Reactor in
Fukushima Daiichi NPS, TEPCO(107) Long-term Corrosion Evaluation of PCV Floor by theSensitivity Analysis of Debris Deposition Situation (Hitachi-GE)(109) The Difference of Reactor Core Energy in Process of theReactor Core Materials Transport and its Impact (JAEA)(113) Summary of Comprehensive Analysis/ Evaluation in Unit3 (TEPCO HD)(114) Summary of Comprehensive Analysis/ Evaluation in Unit1 (Hitachi-GE)(115) Summary of Comprehensive Analysis/ Evaluation in Unit2 (JAEA)
Approach by Toshiba Energy Systems & Solutions Corporation
toward Decommissioning~ Progress of Investigation inside PCV for Fuel Debris Retrieval
and Issues ~Fission Product Behavior under Light Water Reactor Sever
Accident in the light of Fukushima Daiichi NPS

Sept. 21,2018

Introduction of R&D Related to the Fuel Debris Retrieval in Fukushima
Daiichi NPS (focused on criticality control technology development)

Sept. 28,
2018

Development of Robotic Technology Associated with
Decommissioning Work and Site Application

Sept. 5 ~7, 2018

33

Summer Seminar, Hokkaido Branch Office, Atomic Energy
Society of Japan

Sept. 10,
2018

Geochemical Society of Japan Annual Convention in 2018

Sept. 11 ~13, 2018

38

Lecture Meeting on the Decommissioning of NPS,
Fukushima Prefectural Centre for Environmental Creation

Sept. 21,
2018

40

Intensive Course “Introduction to Nuclear Decommissioning
Engineering/ Special Topics, 2018”, Intensive Course of
Tohoku University

Sept. 29,
2018

34
35
36

37

39

41
42

43

44

45
46

47

30

Details

JOINT ICTP-IAEA International School on Nuclear Waste
Actinede Immobilization

The 7th Reactor Physics/ Nuclear Data Research Workshop,
Tohoku University

Intensive Course “Introduction to Nuclear Decommissioning
Engineering/ Special Topics, 2018], Intensive Course of
Tohoku University

The Nuclear Almanac, 2019

Oct. 20182018 International Severe Accident Management
Conference (ISAMC)

Oct. 15, 2018

4th International Conference on Maintenance Science andTechnology 2018 (ICMST-Tohoku 2018)

Oct. 23 ~26, 2018

NFD Workshop, 2018

Oct. 24, 2018

International Conference on Dismantling Challenges:
Industrial Reality, Prospects and Feedback Experience

Journal of Nuclear Science and Technology

Annual Research Report 2018

Oct. 22, 2018

Oct. 24, 2018

Study on the Generation Mechanism of Insoluble Cesium
Particle, Type-A

Current Situation of Technological Development involved in theDecommissioning of Fukushima Daiichi NPS

Characterization and Treatment of Fuel Debris

Status of R&D for Decommissioning of Fukushima Daiichi NPS
conducted by International Research Institute for Nuclear
Decommissioning (IRID)

Challenging Issues and International Joint Research Project
regarding fuel debris characterization toward decommissioningof Fukushima Daiichi NPP
The CMMR Program: BWR Core degradation in the CMMR3 Test
Shape-Changing Robot, “PMORPH”

Muscular Robot with Flexible Structuring Mechanism
Development of Robotic Technology Associated with
Decommissioning Work and Site Application

Fiber optic type compact dosimeter using Nd: YAG for basement
investigation inside primary containment vessel

No.

Presented by

Date

Details48

2018 Fall Lecture Meeting by Fluid Power System, Japan
Fluid Power System Society

Oct. 25 ~26, 2018

Estimation and Control of Hydraulic Cylinder Output Using
System Identification

50

The 116th Robotics Seminar, Japan Robotics Society

Oct. 31, 2018

Restoration of Robot Technology in NPS

49

51

MATLAB EXPO 2018 JAPANThe 2nd Meeting in FY2018; Atomic Energy Society ofJapan; “Behavior of Fission Products at Severe Accident,” FP
Experiment Working Group, Research Expert Committee,

The 6th Regular Lecture Meeting,
Inertial Sensor Application Technology Research
Association, NPO

Oct. 30, 2018
Nov. 2, 2018

Design Development of Fuel Debris Retrieval Robot for
Fukushima Daiichi NPS to utilize MBD

Development of Remote-operated Decontamination Technology
in Reactor Building; Introduction of Results Involving
Investigation of Contamination

Nov. 9, 2018

Utilization of Inertial Sensor Robot for NPS

Materiaux, Strasbourg, France

Nov. 19 ~23, 2018

First analyses of concrete-corium solids surrogate for the
Fukushima Daiichi severe accidents

Robot & Aerospace Festival in Fukushima, 2018

Nov. 22, 201857

E-Journal of Advanced Maintenance Vol.10 No.3

Nov. 25, 201858

The 7th Reactor Physics Professional Study Meeting, Atomic
Energy Society of Japan

Nov. 26, 201859

Subcommittee of Building Structure Performance,
Decommissioning Study Committee, Atomic Energy Societyof JapanNov. 28, 2018

“The 11th Radiation Measurement Forum, Fukushima”, The
186th Committee Meeting, Japan Society for the Promotionof Science

Dec. 3, 2018

52

53
54

55
56

61
62

Membership Lecture Meeting; The Japan Machinery
Federation, General Incorporated Association

Nov. 13, 2018

Nov. 30, 2018

The 15th Ibaraki Area Analysis Technology Exchange
Meeting

Dec. 7, 201865

Nuclear Decommissioning Technology/ Human Resource
Development Forum and IRID Small Work Shop in FY2018,
Tokyo Institute of Technology

Dec. 26, 2018

66

Maintenology Academic Journal, Japan Society ofMaintenology

63
64

Jan. 2019

Three Dimension Analysis by Muon Scattering Method

Subsidy Project on Decommissioning and Contaminated Water
Management in FY2015 Supplementary Budgets; Developmentof Technology for Investigation inside PCV, “Investigation,”
Investigation inside the Pedestal for Unit 3” Development ofTechnology for Investigation inside Primary Containment Vessel
Status of R&D Conducted by IRID; Investigation and Retrieval of
Fuel Debrisl by using Robot
Development of a MAV Equipped with Rotatable Attachment
Adhesive Mechanism for Movable Observation System

Criticality Behavior Analysis Assuming Fuel Debris Retrieving
Work Based on Criticality Scenario; Development of Criticality
Approach Monitoring System(1)Validation of criticality approach monitoring system by using
KUCA, Development of criticality approach system(2)Validation of MVP2.0.33 Time-list Mode
R&D Results in FY2017; Development of Seismic Performanceand Impact of RPV/ PCV, in Fukushima Daiichi NPS
Reinforcement High Temperature Corrosion Test on RPV
Pedestal

DataSheet60

Research Seminar “Possibility of Nondestructive Analysis
through Muon”

Current Situation of Technological Development for theDecommissioning of Fukushima Daiichi NPS

Development of Alpha Contamination Remote-operated
Measurement Technology Assuming Fukushima Daiichi NPS

Study of Application to New Material Hardness Evaluation
Method by the Multiple Spectra Simultaneous Measurementof Laser-induced Breakdown Spectroscopy (LIBS), (1)Physical
Properties Evaluation of Test Sample for Hardness Measurement
by Using LIBS
Study of Application to New Material Hardness Evaluation
Method by the Multiple Spectra Simultaneous Measurement ofLaser-induced Breakdown Spectroscopy (LIBS), (2) Study ofApplication to Material Hardness Measurement by LIBS
Current Situation of Technology Development for Fukushima
Daiichi NPS by IRID
Status of Investigation Inside PCV in Fukushima Daiichi NPS;
Development Status of Submersible-type Boat Device

Subsidy Project on Decommissioning and Contaminated Water
Management in FY2015 Supplementary Budgets; Developmentof Remote-operated Investigation Device for Investigation inside
PCV and Situation of inside Pedestal in Fukushima Daiichi NPS
Unit-2-3.

31

A NN UA L

R ES EA RC H

Others - Datasheet

RE P ORT

2 0 1 8

List of Contract Research in FY2018
No.

Project Name

Category

Development of Technology for Fuel Debris
Analysis/ Characterization

1

Upgrading of Approach and Systems for
Retrieval of Fuel Debris and Internal Structures(Development of Technology for Establishmentof Criticality Control Method)2
3Upgrading of Fundamental Technology
for Retrieval of Fuel Debris and Internal
Structures

4

5

R&D for Treatment and Disposal of Solid
Radioactive Waste

67

Subject
ResearchonRadioactive
Fine Particle
Contracton Surface of Liquid Phase
Research Behavior
and Gas Liquid Phase9

R&D for Treatment and Disposal of Solid
Radioactive Waste (R&D on Preceding
Processing Methods and Analytical Methods)

1011

Evaluation on Trajectory Generation
Contract Considering the Avoidance of
Research Interference with the Environment ofMulti-freedom Robotfor the Estimation/
Contract Investigationof Hydraulic Driven
Research Control
Manipulator Finger Load ForceContract Study on the Accuracy Improvement ofResearch Analytical Evaluation Method

The University of Tokyo

Apr. 1, 2018 ~Jan. 31, 2019

Kobe University

Jun. 15, 2017 ~Feb. 15, 2019

Osaka University

Central ResearchInstitute of ElectricPower Industry
National Nuclear
Laboratory (UK)Contract Research on Waste Management
Research (Hydrogen Generation)of Highly
Contract Analysis and Evaluation
Nuclear Fuel
Materials generated from Nippon
Research Contaminated
Development Co., Ltd.
Fukushima Daiichi NPS Accident
Study on theContract Research and of
New Material including
Research Solidification
Recycling Materialof Water Containing
Contract Development
Evaluation Method in New
Research Component
Cement Material
Contract Investigation for the Physical Property
Research Change of Cement based Material
Research on Simple and Prompt
Contract Analytical Method for the Treatment
Research and Disposal of Solid Waste in
Fukushima Daiichi NPS

12PeriodTest of Criticality Approach
Contract Verification
System including Fuel Debris Kyoto University
Research Detection
to be Applied to the System

Contract Research and Study on Evaluation
Research Approach

8

Partner

Central ResearchInstitute of ElectricPower Industry

Taiheiyo Consultant
Company, Limited
Japan Chemical Analysis
CenterList of Major Research Facility/Equipment in FY2018

32

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

Project nameFull-scale test of repair and water stoppage technology for leakage points inside PCVDevelopment of Technology for Investigation inside PCVAnnual Research Report 2018

Aug. 6, 2018 ~Jan. 31, 2019

Aug. 1, 2017 ~Feb. 15, 2019

Jul. 1, 2018 ~Jan. 31, 2019

Aug. 3, 2018 ~Feb. 28, 2019

Sept. 3, 2018 ~Jan. 31, 2019

Apr. 1, 2018 ~Feb. 28, 2019

Apr. 1, 2018 ~Feb. 28, 2019

Apr. 1, 2018 ~Feb. 28, 2019

Apr. 1, 2018 ~Feb. 28, 2019

Jul. 2, 2018 ~Jan. 31, 2019

Over 1 Million Yen

Details
Heating/ Water Supply Facility
Muddy Water Treatment Facility
Working Floor
Test Sample Mitigation Rail
Full scale Mock-up Facility
B1 Survey Equipment
Scattering Prevention Facility for B1 Survey Equipment
Ancillary Facility for B1 Survey Equipment
Simulated Object for B1 Survey Equipment Mockup Test
Shielding Block Removing Device
Fuel Debris Shape Measuring Device
Element Test Device for Fuel Debris Shape Measuring Device
A2 Survey Equipment (Chamber and Guide Pipe included)
X-6 Penetration Hole Punch: 1 set
In-Penetration Advanced Check Device: 1 set
Sediment Removing Device: 1 set (Chamber included)
Ancillary Facility for A2 Survey Equipment: 1 set
In-Pedestal preliminary Confirmation Device: 1 set
In-PCV Structure Simulated Mockup Body: 1 set
A3 Survey Component Test Equipment: 1 set
Hatch Opening Device Associated Machine: 1 set
B2 Survey Equipment
Scattering Prevention Equipment for B2 Survey Equipment
Ancillary Facility for B2 Survey Equipment
Simulated Object for B2 Survey Equipment Mockup Test
Underwater Swimming Type Equipment Trial Model: 1 set
X-6 Penetration Hole Remote Punch: 1 set

No.

Project name84

Tank Storage Tent for Slurry Preparation Tank Facility

28
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
54
55
56
57
58
59
60
61
62
63
64
65
66
67
68
69
70
71
72
73
74
75
76
77
78
79
80
81

Fuel Debris Characterization

Development of Technologies for Fuel Debris Characterization and Analysis

Development of Fundamental Technology for Retrieval of Fuel Debris and Internal
Structure

R&D for Treatment and Disposal of Solid Radioactive Waste

R&D for Treatment and Disposal of Solid Radioactive Waste (R&D on Preceding
Processing Methods and Analytical Methods)

Full Scale Filter Press

83
85
86
87
88

DataSheet

82

Details
Large Capacity Thermogravimetric Scales andSimultaneous Thermal Analyzer
Crystal Piezoelectric Type 4-Component Cutting Dynamometer
Elemental Analyzing System for SEM
Hydraulic Type Automatic Embedded Equipment
Inverted Metal Microscope
Carbon Coater
Vacuum Replacement Arc Melting Furnace
Fuel Debris Compression Test Device
Fuel Debris Sonic Speed Measurement Device
Metallographic Image Analyzer
Dynamic Micro Hardness Measurement
Simultaneous Thermal Analyzing System
Gas Piping Valve Heater
Sample Cutting Machine
Sample Abrasive MachineCore Sampling Collection Unit
Laser Diffraction Type Gain Size Distribution Measuring Device
Dry Automatic Density Meter
Heating Furnace for Thermal Analyzer
Alpha Particle Aerosol Basic Test Device
Particle Measuring Instrument
Full Scale Test Facility
Full Scale Test Facility
1/4 Scale Test Facility
1/4 Scale Test Facility
1/4 Scale Model Test Facility
1/4 Scale Reactive Force Retaining Mechanism Bonding Test Sample
Test Facility of PRV Upper Part Water Shielding
Flexible Structure Arm: 1 set
Flexible Structure Arm Control Device
Facility for Mockup in PCV
Facility for Equipment Hatch Carry-In Test
1/1 Scale Hydraulic Type Reactive Force Retaining Mechanism
1/1 Scale Electric Reactive Force Retaining Mechanism
Laser Gouging Power Measurement Unit
Laser Gouging Head
Equipment for Access Equipment Element Test
Robot Arm
Access RailPCV Welding Device
Chamber for Alpha Nuclide Analysis
Digital Spectrometer
Efficiency Calculation Program for Gamma-Ray Measurement
Aerosol Migration Observation Instrument
Well-Type Ge Detector
Core Data Collection Unit
Zeolite Sample Collection and Test Equipment
Small Core Sample Collection Test Model
Portable X-ray Fluorescence Spectrometers
Evaluation System Software
Material Cross-Section Processing Machine
Solidified Material Setting Testing Machine
Alpha-ray Telemeter-1; (Equipment set packaging
detector, camera and temperature adjustment
mechanism in a housing: 1 set)
Alpha-ray Telemeter-2; (Combination of detector andcircuit: Used for the comparison to above Alpha-ray
Telemeter-1, for backup)
Dust Testing Device

Test and Research for the Stable Operation of Advanced Liquid Processing System andSlurry Stabilizing Facility

Slurry Preparation Tank Facility

Addition of Full-Scale Filter Press, Dehydrator Pump
and Cleaning Discharge Pipe
Filter Cloth Replacement Platform for Full-Scale Filter
Press Testing Device
Crushing and Transfer Device

33International Research Institutefor Nuclear Decommissioning

5 F, 3 Toyokaiji Building, 2-23-1, Nishi-Shimbashi, Minato-ku, Tokyo 105-0003, Japan
Phone Number: +81-3-6435-8601 URL: http://irid.or.jp/en/

Printed on recycled paper.

Published: July 2019