Annual Research Report2017

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

Printed on recycled paper.

Published: July 2018

IRID’s R&D Projects in FY2017 (Overview)

Greeting

Hideo Ishibashi

Technology for Investigation/
Analysis inside Reactor[Direct Investigation]

Development ofTechnology for
Investigation inside
PCV

Developmentof Technology
for Detailed
Investigation inside
PCV

Contents[Debris Retrieval]

Development ofTechnology for
Investigation inside
RPV[Indirect Investigation/Analysis]
In-House

Measurement and
Evaluation of Fuel
Debris Distribution inside
Reactor at Fukushima
Daiichi
NPS Unit 3

Upgrading theComprehensive
Identification ofConditions inside
Reactor[Securing Work Environment]

In-House
Developmentof RemoteCollaboration Motion
Control System

Development ofTechnology for Fuel
Debris Analysis/
Fuel Debris
Characterization

Repair Technology
for PCV

Decontamination/
Dose Reduction
Technology

Upgrading ofApproach and
Systems for Retrievalof Fuel Debris
and Internal
Structures

Upgrading ofFundamental
Technology for
Retrieval of Fuel
Debris and Internal
Structures

Development ofTechnology forCollection, Transfer
and Storage of Fuel
DebrisDevelopment ofSampling Technology
for Retrieval of Fuel
Debris and Internal
Structures

Development of
Small Neutron
Detectors[Securing Stable State]
Development ofTechnology for
Criticality Control
Methods

Development ofSeismic Resistanceand Impact
Evaluation Methods
for RPV/PCVDevelopment of
Repair Technology
for Leakage Points
inside PCV

President of International Research Institute for Nuclear Decommissioning

Major Research Achievements–Subsidized Projects

Technology for Fuel Debris Retrieval

Major Research Achievements In-House Research

March 2018

R&D for Fuel Debris Retrieval

Major Research Achievements Subsidized Projects      “The Mid-and-Long-Term Roadmap toward the Decommissioning of Fukushima Daiichi Nuclear Power Station(NPS), Tokyo Electric Power Company (TEPCO) Holdings, Inc.” was revised by the government based on the
updating of the Technical Strategic Plan for 2017 Decommissioning of Fukushima Daiichi NPS, TEPCO Holdings,
Inc.” by the Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF) last September.
      The indication was then given that retrieving the debris from the bottom of the Primary Containment Vessel(PCV) shall take place in advance with a focus on the “Partial Submersion-Side Access Method” as the fuel debris
retrieval policy.
      The International Research Institute for Nuclear Decommissioning (IRID) has been engaged in the researchand development (R&D) of the technology required in the decommissioning of the Fukushima Daiichi NPS as anurgent issue since being established in August 2013.
This has resulted in the situation with the PCV and the reactor being clarified upon as well identifying thetechnical issues to be overcome using the development of technology for use in investigating inside the PCV anddetecting the fuel debris using cosmic rays.
      The Annual Research Report 2017 is intended to summarize the achievements of the R&D projects (15 subsidized projects and 2 in-house research) undertaken by the IRID in FY2017. We would appreciate if this report helps
to cast light on the R&D achievements the IRID has been responsible for.
      Seven years have now elapsed since the accident that occurred at the Fukushima Daiichi NPS after the Great
East Japan Earthquake and 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 timely nuclear decommissioning in thereby fulfilling our responsibilities.
      We sincerely appreciate your kind guidance and continued support.

Subsidized Projects(15 projects)

Full-scale test for
Repair Technology
for Leakage Points
inside PCV

R&D for Treatment
and Disposal ofRadioactive Waste

R&D for Treatment
and Disposal of Solid
Radioactive Waste

In-House Research(2 projects)

R&D for Fuel Debris Retrieval
Development of Technology for Investigation inside PCV ………………………………………………………………… 2
Development of Technology for Detailed Investigation inside PCV ……………………………………………………… 4

Overview of Nuclear Building and IRID’s R&D Project

Development of Technology for Investigation inside RPV ………………………………………………………………… 6
Upgrading the Comprehensive Identification of Conditions inside Reactor ……………………………………………… 8
Development of Technology for Fuel Debris Analysis/Fuel Debris Characterization …………………………………… 10
Development of Repair Technology for Leakage Points inside PCV ……………………………………………………… 12
Full-scale test for Repair Technology for Leakage Points inside PCV …………………………………………………… 14
Upgrading of Approach and Systems for Retrieval of Fuel Debris and Internal Structures …………………………… 16

Fuel Debris
Reactor Building
Spent Fuel Pool

Upgrading of Fundamental Technology for Retrieval of Fuel Debris and Internal Structures ………………………… 18
Development of Sampling Technology for Retrieval of Fuel Debris and Internal Structures …………………………… 20
Development of Technology for Collection, Transfer and Storage of Fuel Debris ……………………………………… 22
Development of Small Neutron Detectors …………………………………………………………………………………… 24
Development of Technology for Criticality Control Methods ……………………………………………………………… 26
Development of Seismic Resistance and Impact Evaluation Methods for RPV/PCV …………………………………… 28

R&D forFuel Debris Retrieval

R&D for Treatment and Disposal of Radioactive Waste
R&D for Treatment and Disposal of Solid Radioactive Waste ……………………………………………………………… 30

Major Research Achievements–In-House Research

Measurement and Evaluation of Fuel Debris Distribution inside Reactor at Fukushima Daiichi NPS Unit 3 …………… 32
Development of Remote Collaboration Motion Control System…………………………………………………………… 32

Others–DataSheet

Major Research Results in FY2017 …………………………………………………………………………………………… 33
List of Joint Research/Contract Research in FY2017 ……………………………………………………………………… 36
Major Research Facility/Equipment …………………………………………………………………………………………… 36

Others DataSheet

Reactor Pressure
Vessel (RPV)

Dry Well (D/W)

R&D for Treatment
and Disposal ofRadioactive Waste

Vent Pipe
Suppression Chamber(S/C)
Torus RoomPrimary Containment Vessel (PCV)1

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval

RPV
PCV

RPV

Major Research Achievements Subsidized Projects

Development of Technology for Investigation inside PCV

Structure

PedestalThis then results in the necessity for the above issues tobe resolved alongside the development of technology thatenables the PCV to be internally investigated.

TIP guide pipe support
CRD exchanger
Fallen grating partFallen object
Slot openingm

The environment is quite severe with high doses of radiationand humidity, and limited visibility was confirmed due tothe steam and accumulated water and inherent darkness.
Moreover, the possibility of falling or interfering objects due tothe accident also exists.

Reactor Building.1

Regarding the conditions inside the PCV, visual images,
the radiation dose, and temperature have been obtained byaccessing the PCV internally via X-100B penetration at Unit
1 and X-53 penetration at Units 2 and 3.

ox

The consideration is that the reactor cores melted andnuclear fuel exists in some parts of the reactor inside as fueldebris in the Reactor Pressure Vessel (RPV) and PrimaryContainment Vessel (PCV) of the Fukushima Daiichi NPS
Units 1–3. Fuel debris presumably spread from the opening
at the bottom of the pedestal to outside it after dropping
through the bottom of the RPV into the pedestal supporting
it, although the status of the fuel debris is yet have been
identified.

pr►Purpose

Flat bar
Distortion of grating

Ap►Background

x. 1m

Appro

A2 Investigation: Image Processing Results Based on Data Acquired from the Pedestal Opening

Cable feedmechanism

Part of fuel assemblies(Upper tie plate)

Lookdown
camera►Major Approach and Results Development of Access, Investigation Device, and System for Specific Area

CameraA2′ investigation: Part of image
data obtained under platform Investigation inside the Unit 2 Pedestal (A2/A2′ investigation)
An A2 investigation device, which accesses the PCV through the X-6 penetrating Unit 2 and investigates above the platform
through the pedestal, and A2′ investigation device, which investigates under the platform, were developed to investigate the conditions within the pedestal and as a verification test of the devices. The results of the investigation revealed parts of the grating tohave fallen onto the platform and deposits that look like pebbles/clay all over the bottom of the pedestal. In addition, the investigation confirmed that part of the fuel assemblies had fallen, and in which deposits suspected to be fuel debris (Fig. 1).

 Investigation inside the Unit 3 Pedestal (Investigation for Unit 3)
A submersible type robot for use in accessing the inside of the PCV from X-53 penetration of Unit 3 and thereby investigating theinternals of the pedestal was developed, and which was then used in a validation test. The results of the survey confirmed that thestructures were damaged and molten material, that had presumably solidified, was attached to the structures within the pedestal.
Moreover, lava-like deposits were confirmed in multiple locations within the pedestal via the verification test (Fig. 2).

 Planning for Investigation of Basement Floor, Element Test
Continuing on from FY2016 an element test for the measurement technology was performed. The test confirmed the capabilities of
the CdTe semiconductor detector in the high dose rate environment. In addition, the technology used to measure the pedestal wall
surface ology was changed from the impact elastic wave method to the low frequency ultrasound method because of the latest
status of the PCV in Unit 1, and which then enabled knowledge to be gained on identifying the aggregate and bottom surface echo.

We intend to perform detailed design/test manufacturing/on-site validations within the project of “Development of PCV inside
Detailed Investigation Technology” and use the data obtained in this project to obtain further information on inside the PCV, andwill then prepare for on-site verification of Units 1 and 2.

A2′ investigation: Tip part of investigation device

Fig. 1: Data Obtained from A2 investigation/A2′ investigation
Underwater ROV injection Port
(PCV penetration Part)

RPV
Water surface

Pedestal

Access route(Submersible type device)
Elevation thruster

Drive thruster

Front camera

CR guide tube
(assumed)

Others DataSheet►Future Developments

External
lighting

Major Research Achievements In-House Research

Dosimeter/
Thermometer

CR index tube
(assumed)(Near CRD housing)Lighting

Neutralfloatation
cable

Solidified lavalike deposits(Lower part of pedestal)Fig. 2: Appearance of Submersible Type Device and Survey Route2

Annual Research Report 20173

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background►Purpose

The reactor cores have been considered to have melted
and nuclear fuel to exist in some parts inside the reactors
and as fuel debris in the Reactor Pressure Vessels (RPV)and Primary Containment Vessels (PCV) in the Fukushima
Daiichi NPS Units 1–3. The PCVs were internally investigated: outside the Unit 1 pedestal (B1/B2 investigations),
inside the Unit 2 pedestal (A1/A2/A2′ investigations) andinside the Unit 3 pedestal using a small submersible ROV(Remotely Operated Underwater Vehicle), thus providing
very valuable information. However, this did not satisfy therequirements with respect to the fuel debris retrieval because
the existing opening has restrictions to it.

“The Detailed Investigation inside PCV” project that follows
the existing “Investigation inside PCV” project requires aninvestigation through a larger access route in the PCV byenlarging the investigation device and upgrading the measurement technology to be installed on the device in thereby
obtaining the necessary data for the fuel debris retrieval
process.

Arm

X-6 Penetration Connection Structure

Air lock(External door)

Rail Establishing Access Route into PCV via X-2 Penetration

 Access/Investigation Device

Abrasive wear jet equipment
Transport cart
Fig. 2: Perforation Device for X-2 Penetration (Air Lock) Internal Door

After selecting an arm type access device, which is able to move through the air to gain access from the X-6 penetration internally
into the PCV of Unit 2, a detailed design and test manufacture were commenced upon (Fig. 3).

 Applicability Validation of Element Technology

Grating
Major Research Achievements In-House Research

Air lock(Internal door) Development of Access/Investigation Devices

Detailed designs and test manufacture of equipment related to establishing an access route into the PCV via X-2 penetration (Airlock) (Abrasive Wear Jet Equipment for use in perforating the grating and air lock door, etc.) for Unit 1, which is a high dose area
near the X-6 penetration, have been completed. On-site validation (function test) was then commenced upon (Fig. 2).

X-6 penetration

Fig. 1: X-6 Penetration Connection Structure►Major Approach and Results

The basic design of an X-6 penetration connection structure (Fig. 1) that is operated by remote control from an isolated room,
which was studied in the FY2016 project of “Development of Technology for Investigation inside PCV”, has been completed. Afterthe basic design was completed a more detailed design and test manufacturing were then commenced upon.

Isolation valve

Isolated room

For this reason, the urgent priorities are obtaining information on a method of retrieving the fuel debris and a detailed
design of a fuel debris retrieval device, along with establishing an access route into the PCV as well as developing
technology for a detailed investigation of inside the PCV in
thereby fulfilling the aims of the surveys.

 Establishing Access Route into PCV via X-6 Penetration

Major Research Achievements Subsidized Projects

Development of Technology for Detailed Investigation
inside PCV

Dual-arm manipulator for maintenanceTool, sensor

Site control panel

Measurement technology to be installed on the access unit was reviewed based on an updated survey plan, and test manufacturing has been commenced upon toward the compatibility validation of size/shape measurement technology and radiation
measurement technology.

Others DataSheet►Future Developments
We intend to continue to establish an access route into the PCVs and perform detailed designs, test manufacturing, and on-site
validation of devices for use in detailed investigations of inside the PCVs and preparing for site validations at Units 1 and 2.
Operation control room
EnclosureArm

Fig. 3: Arm Type Access Unit
4

Annual Research Report 20175

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval
Equipment carry-in entrance
Air lock cell

►Background►Purpose

Information on the fuel debris and inside the Reactor Pressure
Vessels (RPVs) needs to be obtained in advance and including their location, shape, and the internal situation. However,directly obtaining the information is difficult because RPVs
have internally complicated structures and extremely highradiation levels.

Making progress with future decommissioning, including fueldebris retrieval in a safe and smooth manner, necessitatesamassing information on the currently unclear situation.
Therefore, the investigative targets will be clarified based on
the relevant information, and feasible technology developed
for the investigation that requires a method of perforating thereactor core by it accessing from the top or the side of the RPV.

►Major Approach and Results Development of Equipment to Access Reactor Core

Major Research Achievements Subsidized Projects

Development of Technology for Investigation inside RPV
Spare nozzle cutting
deviceWork cell

For investigation
deviceFor processing
reactor internals

Tool box

Tool box cart

Operation
floor in reactor
building

PCV head

Guide pipe

Suspension device stored in tool box

PCV head connection part
Guide pipe

RPV head
Steam dryer
RPV head drillingdevice Development of Equipment to Access the Reactor Core from the top

Shroud head

Steam-water
separator

For
For full-scale preliminary
surveyconfirmation

AWJ tool head

Investigation device
for inside reactor

Reactor internals
drilling device

Simulated PCV head

Seal

Boundary function
maintenance device

-1 Development of Drilling Device to Access through RPV Head

Major Research Achievements In-House Research

A possible method was confirmed by reviewing the work plan, specifications, and test manufacture of a device assuming that thewell cover, Primary Containment Vessel (PCV) Head, and RPV Head would need to be drilled through in establishing an accessroute to inject the device which will then require investigating whether the device is able to enter from the operation floor on theupper part of the RPV up to the reactor core of the RPV. In addition, and regarding the RPV Head processing device, an element
test for the removal method of RPV Head spare nozzle located in the access route was conducted to avoid any interruption in theinternal structures of the RPV, and remote processing/workability confirmed.

-2 Development of Boundary Function Maintenance Device/Access Work Device

In order to access the RPV without compromising the boundary function (contamination prevention system) and maintaining anegative pressure environment, and assuming the installation of an access device for the work (work cell) on the operation floor
and connection to the PCV head via a guide pipe, the possibility of method was confirmed by reviewing the work plan, equipment
specifications, and a sealing property element test.

Overview of Investigative Method with Access from the Top

In addition, specifications for a work cell were clarified with respect to the connection method of the tool box that will contain theprocessing equipment and survey equipment, and the carry-in/out and transport method for each equipment.

-3 Development of Device for Drilling through Upper Grid Plate

Assuming a need to drill through the reactor internals (steam dryer, steam-water separator, and shroud head) from the operation
floor to the upper grid plate just below the perforation position of the RPV head, the feasibility of a method was confirmed byreviewing work plan, equipment specifications, and test manufacturing.

PCV

Reactor building

Assuming a need for remote processing in a narrow part of the RPV, an element test of a mock-up facility was performed combining an Abrasive Wear Jet (AWJ) Cutting Device and access equipment installed in a work cell, and the possibility of remote
cutting then confirmed.

 Development of Method of Accessing from SideAn access route into the reactor core from the side of the reactor building was determined upon, and concepts of the methods foruse established that included selecting the applicable tools for excavation and sealing work. In addition, the specifications for all
the equipment were clarified, including the design of a facility for use in maintaining the equipment.

 Development and Selection of Methods of Investigating the Reactor Core

RPV

Biologicalshield wall

Isolation
valve

ACx connection
part System Design and Planning Methods for Overall Investigation Device

ACx cell (used for replacing AC0 cell) withtool (excavation, guide pipe insertion, sealing
installation)

An exposure evaluation during an investigation after accessing the PCV from the side was conducted under the current micro-positive
pressure situation in the PCV. Moreover, An exposure evaluation during an investigation method via access from the top is being
planned for after the establishment of negative pressure and from the aspect of any spreading of the contamination caused by
dust diffusion due to the large amount of the reactor internals that radioactive material is assumed to be adhering to.

►Future Developments
Toward the design of the applicable device for on-site investigations, an element test on the workability will be conducted as necessary based on the environmental information and conditions at the work site, and the test results reflected in the design of thedevice. Furthermore, coordination with the related construction work required for the device design, facilities, and the utility supply
system required for the site survey, along with the applicability of the incidental system, will be reviewed.
6

Annual Research Report 2017

Others DataSheet

Two kinds of specifications for investigation devices in prior confirmation of access routes via a small diameter opening and an
actual investigation via a large diameter opening were studied besides the investigative steps, visibility, and radiation resistancebeing confirmed by an element test. In addition, an accessibility test using a mock-up facility that simulated openings was performed and the remote workability confirmed. The expansion mechanism in investigating the deepest part of the reactor core
via horizontally accessing the center of the reactor core from the upper grid in the shroud head was test-manufactured, and thefeasibility of the mechanism then confirmed by an element test.

Working cell

AC0 cell fixed on east wall surfaceof reactor building

Investigation with Access from the Side: Appearance

7

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background►Purpose

Identifying the conditions inside the reactor and PrimaryContainment Vessel (PCV) is essential in establishing amethod of retrieving the fuel debris and developing safetymeasures. However, a direct survey of and observation
within the reactor are difficult because the radiation levels
within the reactor at the Fukushima Daiichi Nuclear Power
Station (NPS) Units 1–3 are extremely high.

This project aims at a steady approach to the decommissioning of the Fukushima Daiichi NPS. The conditions
therefore inside the reactor and PCVs are being estimatedusing comprehensive analysis and evaluations based on
information gained through accident progression analysis
and other research and development (R&D), and analysis ofmeasured data such as the pressure, temperature, and otherinformation during the accident. This is a joint project with theInstitute of Applied Energy.

►Major Approach and Results
 Comprehensive Analysis/Evaluation of Reactor Internals

 Comprehensive Analysis/Evaluation Based on Actual Data and Results from other Projects

 Establishment of Database Required for Comprehensive Analysis/Evaluation
An English language search system for the database was added and graphic display system for the measurement data, althoughmainly for the 3 weeks after the accident, improved in promoting the utilization of expertise from overseas organizations. In addition, the database with search tags established in this project and information are being updated in contributing to more comprehensive analysis and effective evaluation, as described in .

 Fuel Debris/FP Behavior and Characterization in Comprehensive Analysis and Evaluation
 Reduction of Uncertainty Using Analysis Method

Sensitivity and other analysis taking into consideration the boundary conditions and an analysis model for events assumed to haveoccurred within the reactor were conducted using the accident progression analysis code, and which then provided expertise thatwill contribute to the more comprehensive analysis/evaluation described in . For example, continuing on from the last fiscal
year a simulated fuel assembly plasma heating test (Fig. 2) was performed, which provided expertise leading to reducing theuncertainty of events such as the core meltdown and movement within the BWR fuel assembly system.

 Evaluation of FP Chemical Properties
In evaluating the FP chemical properties, and with a focus on Cs, which is largely contributing to the radiation dose during thedecommissioning, a study has commenced upon the distribution of Cs and its chemical properties, including identifying the chemical species that need to be considered in addition to standard chemical species such as Csl and CsOH, and the insoluble Csparticles whose production has been confirmed in the environment, and the possibility of uneven Cs distribution associated withthe reaction in the upper structures of the RPVs. In addition, on-site test samples have been analyzed, and a study of the composition and spatial distribution of uranium and FP conducted from the aspect of identifying the conditions inside the reactor (Fig.3).

• The current shroud might be damaged and/or sound(General assumption/analysis).
• The increase in fuel temperature in the outer area may
not be high and the pellet may be in the outer area
(General assumption/test/analysis).
• The heat transfer from fuel debris is rather small andpart of the control rod guide pipe may not have melted(General assumption).
• Particulate debris and pellets may have accumulated
in the stagnant area (General assumption).
• The suspected control rod guide pipe have penetrated
the RPV and the RPV may have a damage hole init the size of the control rod guide pipe (Actual measurement).
• Water has been confirmed in the pedestal of the RPV
and the outer. The damaged opening may be present
in the center of the RPV and outer area (Actual measurement).

• Muon measurement results indicate thata mass of fuel debris may not be present
in the original reactor core region (Actual
measurement).
• Muon measurement results indicate thatpart of the fuel debris may remain at thebottom of the RPV (Actual measurement).
• Debris may have entered into the control rod drive mechanism due to damageof the control rod guide pipes (Generalassumption).
• The investigation results inside the PCV
confirmed the pedestal is internally damaged compared with Unit 2. The amountof fuel debris that has fallen into the PCV
is larger than at Unit 2 (Actual measurement).
• It was confirmed that the control rod drive
mechanism housing support bracket haddamaged and solidified molten material
had adhered around where the fuel debris
may be present (Actual measurement).

• Part of the fuel debris may have solidified
without causing MCCI (General assumption).

• Explosions occurred at Units 3 and 4. The consideration is that hydrogen generated from the MCCI may
have caused the explosions (Actual measurement).
• Water may have accumulated on the dry well floor
during the RPV damage because the dry well was
sprayed for more than one hour from 07:39, March 13,
2011. This may have inhibited the spread of the fueldebris (Actual measurement, general assumption).
• Fuel debris may have spread to the outside of the pedestal through the pedestal opening and did not reachthe shell (Actual measurement, analysis).
• The case that water accumulated on the PCV floor
leads to the possibility of particulate debris (Generalassumption).
• The case of particulate debris leads to the possibilityof accumlation in the stagnant part (General assumption).
Explanatory notes

Damaged control
rod guide pipe
Integral control rod
drive mechanism

Fuel rod
Oxidation debris(porous)Particulate debris• The suspected solidified molten material
in the lower pedestal and fallen grating
and deposits were confirmed (Actual measurement).

Major Research Achievements In-House Research

Using the assumed conditions inside the Reactor Pressure Vessels (RPV)/PCV of all the units an information summary map,
which summarizes the various types if information from the RPV, PCV, and reactor building, was created. The information wasthen comprehensively analyzed and evaluated, and used to provide an estimate diagram of the fuel debris distribution (Fig.1), an
estimate of the FP (Fission Products) distribution, and an estimate of the radiation dose distribution.

• Estimating the amount of energy from thePCV pressure rise due to the hydrogenoccurrence and most of the fuel may havemelted (Actual measurement/analysis).
• The assumption was that fuel debris in thereactor core would be minor (less than in
Unit 2) because no temperature increase
was observed in the RPV when the reactor core spray system stopped during
December 9–24, 2013 (total amountinjected was constant through increasing
flow from the water supply system) (Actual
measurement).
• The assumption was fuel debris may haveexisted in the lower part of the plenum
because the temperature in the RPV lower
decreased when the total amount of water
injected was increased (September 1,2011) from the reactor core spray system(Actual measurement).
• Part of Fuel rods may be present in theouter reactor core (General assumption).
• Melted debris may be generally solidified
through oxidation (General assumption).

Major Research Achievements Subsidized Projects

Upgrading the Comprehensive Identification ofConditions inside Reactor

Concrete mixed
debrisIntegral control rod
guide pipe

Control rod drive
mechanism (debris inside)
Integral shroud
Pellet
RPV damage

Fig. 1: Diagram of Estimated Fuel Debris Distribution (Example of Unit 3)

As of February 28, 2018

Deposits in Unit 1
PCV

Curing sheet in
Unit 2

Appearance of analyticalsampleAppearance of analyticalsampleElectronic microscope
image

Electronic microscopeimage

 Utilization of Domestic/Overseas Knowledge through International Joint Research►Future Developments
This project will be completed at the end of FY 2017 (end of March, 2018). Tokyo Electric Power Company Holdings, Inc. intends
to continue comprehensive analysis and evaluations that reflect the results of site surveys results in then estimating the conditions
within the RPVs and PCVs in the future.
8

Annual Research Report 2017

Test equipment appearance

Test result example

Fig. 2: Plasma Heating Test of Simulated
Fuel Assemblies

Others DataSheetThe International Joint Research (OECD/NEA BSAF Phase2) project involves the sharing of accident progress scenarios andplant information with overseas organizations by utilizing the database that was established and which is shown in , and theevaluation results of the accident progression and debris/FP distribution by participating organizations being compared to actual
measured values and the results of site surveys. The results confirmed that the deviation in analytical results of the organizations
could be significantly reduced when compared with Phase-1 through a better common understanding of the progression of theaccident and plant conditions, and the evaluation results of the FP emission amounts to be consistent with the amounts emitted
into the environment. The estimate accuracy of the fuel debris distribution was improved via more thorough knowledge of the
progression of the accident. The expertise obtained through discussions with overseas organizations was utilized in the comprehensive analysis and evaluations described in in .

Fig. 3: Example of Analysis Sample andAnalysis Results9Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background
The reactor at the Fukushima Daiichi NPS was internallyinvestigated in studying fuel debris retrieval methods and thecollection/storage of it after it has been retrieved. Progresswith safe and steady decommissioning work will necessitate
identifying the characteristics of the fuel debris generated in
the reactor. In addition, technology needs to be developed foruse in analyzing fuel debris samples.

►PurposeThis project is aimed at surveying and assuming about thecharacteristics of the fuel debris that include the hardness ofthe fuel debris, which will be useful in studying the necessary
retrieval device, and the behavior of the dry fuel debris, which
will be necessary in studying the collection/storage of fueldebris and in order to provide information for the project thatwill be about proceeding with the actual decommissioning
work. In addition, the properties of the fuel debris, which
consists of complicated materials that include fuel, structural
materials, and concrete, will be promptly and accurately clarified, and analysis technology prepared for updating the “FuelDebris Property List”.

Major Research Achievements Subsidized Projects

Development of Technology for Fuel Debris Analysis/
Fuel Debris Characterization

Upper crustSide boundary
Bottom boundary(metal)Upper part of oxidized
product layer
Lower part of oxidized
product layer

Element map of upper part of crust
cross-section►Major Approach and Results
A surface dose rate evaluation method for the fuel debris was studied and test evaluations of the surface dose rate of fuel debris
sample cases performed. In addition, the “Fuel Debris Property List”, which involves fuel debris property assumptions, wasupdated using the results of analysis of large scale MCCI (Molten Core Concrete Interaction) test products obtained in FY 2017.

 Characterization via Use of Simulated Debris

Element map of bottom partboundary cross-section

Cross sectional element map oflower part of oxidized product layer

Fig. 1: Element Distribution of Large MCCI Test Products

 Evaluation for Characterization of Uneven MCCI Products

Large MCCI test products were analyzed that had been produced in the test facility owned by the French Alternative Energies
and Atomic Energy Commission (CEA) in FY 2016, and for use in obtaining information which includes an element map of MCCI
test product layers (Fig. 1), and the crystal structure and hardness, etc. Confirmation took place that no significant change in theestimation is needed based on the basic test results so far, even if after taking the conditions at the Fukushima Daiichi NPS into
consideration.In addition, expertise related to the time variation with concrete erosion shapes was obtained by analyzing dataobtained from the MCCI test.

Major Research Achievements In-House Research Assumption of Fuel Debris Properties

With regard to the large MCCI test products produced in FY2016, the porous regions and metal layers wereobserved with their uneven status. The results of analyzing each layer revealed that the test samples, mainly
consisting of the upper part of the crust and lower part of the oxidized product layers consisted of U-Zr oxidationand other products in a high Si containing (containing Ca, Al) matrix. In addition, the discovery was made that themetal block at the bottom of the boundary consists of an alloy which mainly consists of Fe.

 Emission Behavior Evaluation of Fission Products during Dry Heat Treatment

The emission behavior of Fission Products (FPs) was studied for use in the design of an off-gas facility and the dry treatment whichis under consideration as method of pretreatment when the fuel debris is stored. A reference investigation was performed withregard to the FP emission behavior via heating in FY2017, and a medium volatile FP candidate that has a large impact on the environmental emission evaluations selected (Fig. 2). In addition, confirmation took place that emission start temperature/emission
speed could be clearly measured using a thermogravimetric device/differential thermal analyzer and test samples containing Ru.
The necessary element technology was selected and technology for fuel debris analysis is being developed. And with regard
to the melting of the fuel debris and development of a multi-element analysis method, confirmation took place that components
derived from the structural materials of the fuel debris, including Cr2O3 and Fe3O4, could be melted using the alkali fusion
method. In addition, results of analysis using the multi-element simultaneous analysis method and the inductively-coupled plasma
emission spectrometry device (ICP-AES) with MOX simulated debris proved to have high levels of repeatability.

TeO2
SnO

UO3

CdO

Others DataSheet

Fuel debris analysis technology using an X-Ray CT was developed, and porosity measurements and simulated debris samples
performed with the same porosity measurements as with an optical microscope. The components of samples proved capable ofbeing clarified using the methodology of a combination of X-Ray CT and γ-Ray tomography and test samples with a mix of spent
fuel and covering materials (mixed melted sample materials).

Steam pressure [atm] Development of Elemental Technology for Fuel Debris Analysis

Cs

With regard to the development of technology for the multiple nuclide rationalization analysis method using an induction coupling
plasma mass spectrometer (ICP-MS), interfering ions with the nuclide measurements were specified and target values determined
for their removal, and an impact reduction method for the hindering nuclides (Nb, Mo). A removal method is also being studied.
In addition to the above achievement’s analysis items were examined and an analysis procedure was prepared for the fuel debris
sampling that is presumed to be taking place to be conducted at the existing analysis facility in Ibaraki Prefecture.

►Future Developments

The dose rates on the surface of the fuel debris are continuing to be evaluated for use in assumptions regarding the characterization of the fuel debris and the results will be reflected in the “Fuel Debris Property List”. In addition, and regarding the FP
emission behavior evaluations during dry heat treatment, the emission behavior of selected medium volatile FPs will be evaluated.
Furthermore, the development of a multiple nuclide rationalization analysis method using the ICP-MS is intended for use in establishing a method of reducing the impact of any interfering ions. An analysis procedure is being examined in combination with theflow of the overall analysis. In addition, transporting the fuel debris samples will be studied.
10

Annual Research Report 2017

Temperature [K]Fig. 2: Calculation Results of Steam Pressure of Oxidized Products
The steam pressure of oxides to be generated to counteract the nuclides that will have a significant impact on
the environment was calculated. With the fuel debris dry heat treatment the conditions of the volatile fuel uraniumoxides are yet to have been determined and hence chemical compounds with higher steam pressures than UO3pressure are being focused upon. Meanwhile, the off-gas treatment needs to be considered as all the high volatile
FPs such as Cs will get emitted. The compounds that therefore have a lower steam pressure than Cs but a higher
steam pressure than UO3 were selected as the candidate medium volatile FPs and included Te, Sn and Cd.

11

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval

Welding device
Major Research Achievements Subsidized Projects

Development of Repair Technology for Leakage Points
inside PCV

Guide pipe
Bentonite sludge water►Background►Purpose

The assumption is that melted fuel has not only fallen within
the Reactor Pressure Vessel (RPV) but has also reached thePrimary Containment Vessel (PCV) at the Fukushima Daiichi
nuclear power station (NPS). In order to retrieve fuel debris,the plan is to accumulate water in or inject water into the PCV.
Water will therefore need to be prevented from leaking fromthe PCV.

This project aims at establishing repair technology for thewater leakage points of the PCV in order to substantiate afuel debris retrieval method by submersing it in or injecting
water into it.

Repairmaterial

Vent pipe 1/1 scale test
Rubber material
injection port

►Major Approach and Results

500A×2000mm×2
Steel pipe tilt angle: 20 degrees

Water stoppage
material

 Process Study and Plan for Submersing PCV

115mm×95mm×750mm
Particle size of aggregate:
minimum 7mm

Rubber material pressure
test*Self-injecting concrete composition test was conducted
at Kochi University of Technology (Ouchi laboratory)

Concrete water
stoppage test

Fig. 1: Overview of Repair of Lower Part of PCV (Filling in Vent Pipe and S/C) Development of Repair Technology for Lower Part of PCV
 Reinforcement/Water Stoppage Technology for Suppression Chamber (S/C) and Vent Pipes

Front arm

-1 Reinforcing Technology for S/C Support Columns

The effectiveness of the reinforcement was confirmed based on the results of evaluating the strength after placing reinforcement
materials in a full-scale mock-up facility, flow analysis, and simulations performed in FY 2017.

Flexible guide pipe
Water stop pluginsertion device

-2 Water Stoppage Technology via Filling in Vent Pipes (Fig. 1)

Water stoppage materials using radiation resistant rubber and self-injecting concrete, and repair materials using sludge water
were developed in securing water stoppage capabilities. The workability and water stoppage capabilities were confirmed in a 1/1
scale test using self-injecting concrete as the water stoppage material.
A function validation test on the S/C guide pipe implementation required in inserting the hose for the concrete injection, the waterlevel control, and water drainage in the S/C were performed in confirming their workability.

Water stop plug
Water stop plug
Drawing of water stop plug for vacuum breakage line

Installation of water stoppage plugs and a plug insertion device through the flexible guide pipe were improved upon, and a workability test using a 1/1 scale test facility performed in thereby confirming their water stoppage capabilities. No leakage was confirmed at 0.45 MPa (fully submersed condition) during a water pressure resistance test.

►Future Developments
In order to establish a fuel debris retrieval policy for each of the units, repair and water stoppage technology will be reevaluatedas part of an investigation of the PCV internals and studying the fuel debris retrieval method, and in case the prerequisites andconditions change.
12

Annual Research Report 20170.2

No leaking water

0.1
0
1

0

1

2

3
時間[h]
Time
[h]

4

5

6 0

1

2

Exposure amount

Total dosage,
36% reduction

51.5Sv
33.2Sv

Exposure during
repair work

30.5Sv

S/C water filling stoppage device
Camera device
Water stoppage work area

• Placement locations: 7→4• Shared camera and perforation for
injection

Others DataSheet

The radiation exposure during the water stoppage work needed to repair the lower part of the PCV was evaluated, and a methodof reducing the radiation exposure by decreasing the amount of water stoppage work and the place studied in thereby identifying
the issues that need to be solved. In addition, the exposed dose reduction effect through improving the environment was also
confirmed.

0.3

Fig. 2: Overview of Repair of Lower Part of PCV (Vacuum Breakage Line Filled) Development of Repair Technology for the Upper Part of PCV

 Conceptual Study of Environment Improvement toward Applicable Repair Method (Fig. 3)

0.4

Water Pressure Test Results (check valve side)

Development of water stoppage materials for the inside/outside of the various connection pipes and a repair method, and element
development of remote construction and an access devise were performed in confirming the feasibility of the method.

Abrasion of the equipment hatch water stoppage device, each of the welding heads, and the access device were studied with viewto their improvement, and with the workability of the sealing method then enabling their improvement.

0.5

0 Boundary Formulation Technology for Connecting Pipes

Water Stoppage Technology for Sealant Part (equipment hatch)

Rear arm

Water stoppage material
Increasing(injecting in step 1)
surface pressure
Overview of Water Stoppage Plug

Bellows

 Water Stoppage Technology via Filling Vacuum Break Line (Fig. 2)

BalloonCompressing
all the parts

Vacuum breakage
valve

圧力[MPa]Pressure[MPa]

-3 Water Stoppage Technology via Filling in S/C (Fig. 1)

Major Research Achievements In-House Research

In order to retrieve the fuel debris in a dry condition, several feasible PCV water levels were proposed based on guidelines that
including the purpose of repairing the PCV, the safety requirements, work exposure levels, and seismic capacity, and the target of
the water stoppage capabilities of the individual repair technologies determined.

Total dosage:
45% reduction

Environment
improvement
Exposure with environment
improvement
2.6Sv

Ambient dose rate
Exposure Evaluation with Water Stoppage Repair
work (Unit 2)

Review of total dose reduction during repairs(Unit 1)

Fig. 3: Conceptual Study of Environment Improvement for the Applicable Repair Method (Example)
13

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Purpose

The Fukushima Daiichi NPS has a severe environment
with high doses of radiation and narrow spaces. There is anumber of places where it is extremely difficult for people toaccess and do the necessary decommissioning work. This
then necessitates the development of a method of repairingand stopping the water leaking from the Primary Containment
Vessel (PCV), and for a remote operation device for use in
retrieving the fuel debris.

A full-scale test to validate the technology and operational
training was conducted in order to apply the technology
developed to repair and stop the water leaks on site (methods
and remote operation device, etc.). This project is taking placeat the Naraha Remote Technology Development Center thatwas organized by Japan Atomic Energy Agency (JAEA).

►Major Approach and Results

Vent pipe
Suppression
Chamber
approx.
12m►Background

Torus room
wall surface

Downcomer

app
rox
. 18
m

Suppression
Chamber

Quencherapprox. 20m►Plants to be simulated:
Units 2 and 3

Major Research Achievements Subsidized Projects

Full-scale test for Repair Technology for Leakage
Points inside PCV►Facility to be simulated:
1/8 of the area of the torus room
wall surface and the lower part of
the PCV►Test facility size:
Approx. 18 m × approx. 20 m × approx. 12 m
Unit

S/C volume

Unit 1

Approx.
4800 m3
Approx.
6500 m3

Unit 2-3

S/C cross- Center
sectional diameter
diameter of S/C ring
Approx.
8m
Approx.
9m

Approx.
30m
Approx.
35m

Vent pipe Full-scale Test of Repair Technology for the Lower Part of PCV
 Suppression Chamber (S/C) Support Columns
A test wherein high-fluidity reinforcement materials were placed and injected into the lower part of the S/C was conducted in confirming its feasibility using the assumed procedures and the actual situation. In addition, the feasibility of the operation monitoring,
including the installation height, was also confirmed (Photo-1).

 Vent Pipe Water Stoppage
A workability verification test was conducted in which interfering objects were removed and the vent pipes perforated via remote
operation. Accessibility to the operation area was thereby confirmed with the actual situation (Photo-2).

Major Research Achievements In-House Research

Downcomer

A full-scale test using a test facility (Fig.1) was conducted with respect to the following items.

QuencherUpper part of torus room wall surface

Cross-section of PCV lower part

Inside of S/C

Fig. 1: Overview of Test Facility

 Water Stoppage by Filling S/C (Down-comer Water Stoppage)
A workability verification test confirmed that there is no problem in the facility with the applicability of the prepared remote operation, the repair of the PCV, and work in the high dose environment. In addition, improved workability was considered based on
the knowledge obtained from the test.
The operation monitoring feasibility, including installation height, was also confirmed using the assumed procedures in the actual
situation (Photo-3).

 VR Data Preparation for Preliminary Simulation Test
A remote operation device to be used for stopping the water in the vent pipe was duplicated using a virtual reality (VR) system forthe operational training environment as below.
• Operation measurements of remote device via motion capture, etc.
• Operation comparison validation with remote device and VR system reflecting its operating data.
VR system effectiveness evaluation during operational training was performed (Photo-4).

Photo-2: Implementation of Vent Pipe Water
Stoppage Workability Verification Test

Photo-3: Supplying Water Stoppage Material for
Injecting Filling to Stop Water inside S/CPhoto-4: Operating Panel of VR Systemand 3D Screen

Others DataSheet

Photo-1: Implementation of S/C Support
Column Reinforcement Test

►Future Developments
Technology for strengthening the S/C support columns and stopping the water in the vent pipes and by filling in the S/C requires
the identification and clarification of the issues with practical application based on the results of full-scale tests. Vent pipe water
stoppages are being prepared for via placement tests. VR data for preliminary simulation tests will then be completed for use inestablishing training environments for remote device operators.

14

Annual Research Report 201715

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background►Purpose

The status with the fuel debris in the Reactor Pressure
Vessels (RPVs) and Primary Containment Vessels (PCVs)of the Fukushima Daiichi NPS is currently ensuring stablecooling. However, the reactor buildings, RPVs, and PCVs
were damaged in the accident and the plant itself is unstable.
The aim is to retrieve the fuel debris from the unstable conditions and establish safer conditions by preventing the spreadof any radioactive material.

The technology for upgrading the approach and a system touse in retrieving the fuel debris and internal structures has
safety issues: ensuring the confinement function, capturing
and removing the dust derived from the fuel debris, and themonitoring of alpha-nuclides (collective term for the radioactive nuclides that release alpha rays). This project aims at
developing technology that solves the above issues and optimizing the method and system in ensuring the utmost safety.

►Major Approach and Results Development of Technology related to Confinement Function* PCV repair technology developed by other PJ

Dissolved
nuclideremoval facility

Supply pump Supply water
sourceEmergency
buffer tankParticle
removalfacility

To watertreatment facility
To watertreatment facility

Boric-acidadjustment
deviceParticle
removalfacility

Particle
removalfacility

To buffer tankParticle
removalfacility

Circulationpump

Monitoring instrument
abbreviation
L: Water level gauge
T: Thermometer
P: Pressure gauge
F: Flow meter
n: Neutron monitor
R: Radiation monitor

Exhaust port

From water
treatment facility

Monitoring
systemExhaust Particle
PretreatmentfanremovalsystemfilterHydrogen
Dust Criticality
densitometer monitor monitor
To watertreatment facility

Cooling
unit
From torus
room

Buffer tank

Torus room intake
D/W intake
S/C intake
pump (regular/
pump (regular/
pump (regular/
emergency use) emergency use) emergency use)

Exhaust Particle PretreatmentfanremovalsystemfilterExhaust portaCooling
unit
Hydrogen
Dust Criticality
densitometer monitor monitor

Particle
removalfacility

Monitoring
systemNitrogen supply system

In-leak fromnegative PCV
pressure

Particle Pretreatment
removalsystemfilterNitrogen supply system

Pretreatment
systemParticle removal
system (HEPA Filter)*This drawing is an example installation that satisfies
the functional requirements and has no specified
installation location (inside/outside the R/B) for theequipment nor an equipment model.

Fig. 1: Scope of Element Technology Development for System Formulation

Equipment carry-in crane• An investigation of the technology for use in dust collection/removal process with respect to the gas and liquid systems was performed using existing technology and the benchmarks identified, and which also identified the necessary items for confirmationin the selection of technological advantages and the element tests. The element tests will be conducted to obtain the findingsin the future.

• The alpha-nuclide monitoring technology was clarified regarding the necessity and purpose of the fuel debris retrieval as well as
the required measurement range of the gaseous system examined.

“Explanatory notes”
Blue: Regular use system
Red: Emergency use system
Green: Monitoring system
Confinement technology via differential
pressure control

Boric-acid water Boric-acidsupply pump water tank

 Development of Technology for Dust Collection/Removal Derived from Fuel Debris Study of Alpha-Nuclide Monitoring System Associated with Fuel Debris Retrieval

Boric-acid water
injection pump Boric-acidwater tank

Water injection
Water injectionpumpsource• The opening area estimation method for the damaged PCV boundary was studied. A confirmation test with the actual use wasthen proposed.

• Assuming the possibility of the alpha-nuclides having melted into the water an investigation of the necessary technology forthe collection/removal of dissolved nuclides in the liquid system was performed using existing technology and the benchmarks
identified, and which also identified the necessary items for confirmation in the selection of technological advantages and theelement tests. The element tests will be conducted to obtain findings in the future.

:R ange of Component Technology Planin this Project:R eflected range of development resultsfrom Criticality Control PJ

Gaseous system(Overall Image) *Major Research Achievements In-House Research• Elemental technology was developed to ensure the confinement function works via differential pressure control, and an implementation policy that combines analysis and element tests formulated. Analysis and element tests will continue to take place intheir confirmation.

Liquid system(Overall image) *Major Research Achievements Subsidized Projects

Upgrading of Approach and Systems for Retrieval of
Fuel Debris and Internal Structures

Equipment transport cart

Equipment hatch cell
Maintenance cell

Equipment cell

Reactor building wall opening
Reactor building wall
Fuel debris retrieval cell
Cell transportation
systemWaste transport cart

PCV wallPedestal

Waste transport cell

Mini-cask system• A survey of existing technology used with alpha-nuclide monitoring of gaseous systems was performed and issues with theactual use dealt with.

 Study on Optimization of Ensuring Safety of Methods and Systems• The results of public exposure evaluations conducted up to last fiscal year were re-confirmed as well as an important study on
workers’ exposure during fuel debris retrieval commenced upon.

Element technology for the collection/removal of dust derived from the fuel debris and alpha-nuclide monitoring is currently being
developed, and the study results will be reflected in the optimization of the fuel debris retrieval method and systems. In addition,
the safety and functional requirements will continue to be monitored and their feasibility confirmed. The partial-submersion side
access method has priority but other methods with the applicability to each unit will be taken into consideration and further examined in confirming their feasibility.
16

Annual Research Report 2017

Fuel debris
retrieval cell
Maintenance cell• The safety requirements and functional requirements during fuel debris retrieval were reviewed and brushed up in re-confirming
the required systems.

►Future Developments

Transport cell

Others DataSheet• With a focus on the partial-submersion side access method, the design conditions of the method with the applicability to each
unit were taken into consideration and reviewed and the cell installation method substantiated (Fig. 2).

Waste carry-out
cell

PCV
Fuel debris
carry-out cell
Pathway (access tunnel)

Fig. 2: Installation Image of Fuel Debris Retrieval Cell

17

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background►Purpose

The fuel debris retrieval policy was determined upon andhence the focus of the study is on the partial-submersion
side access method. This project aims at acquisition of thenecessary data and information via element tests and aconceptual study in thereby evaluating the feasibility of theretrieval method.

The subsidized project of the Upgrading of Approach and
System for Retrieval of Fuel Debris and Internal Structures
is being used to study theoretical retrieval methods. This
project aims at the highest possibility of its realization not
only through theoretical study but also through obtaining therelevant data through element tests.

ChiselSimulated debris (MCCI)
Provided by HITACHI-GE
Nuclear Energy Ltd.

In addition, the development plan will be formulated after
identifying any issues that need to be reviewed.

Major Research Achievements Subsidized Projects

Upgrading of Fundamental Technology for Retrieval of
Fuel Debris and Internal Structures
Inflatable seal
PCV wall simulation Biologicalshield wall
test
facility
Remote equipment
for installing inflatableseals (manipulator)
Equipment
When sealingtransport cart
(Develops
after opening)►Major Approach and Results Development of Technology for Prevention Fuel Debris Spreading

Fig. 1: State of Chisel Processing Preliminary Test

Fig. 2: Inflatable Seal Element Test

 Development of Fuel Debris Collection System Development of Fuel Debris Cutting/Dust Collection System

Simulated fallen andinterfering objects
within pedestal

Interfering object
removal device• Fuel debris to be cut and collected was clarified with respect to where and what shape the fuel debris would be, and the mosteffective processing method collection method studied.
• MCCI products included in the fuel debris have been estimated to be largely present at the bottom of the PCV. Two processing methods, which are the most efficiently (chisel processing method and ultrasound core boring method), were identified as
requiring confirmation via element tests.
• The test sample produced for use in the simulated fuel debris processing tests were used to study the components of the MCCI
products, their size, and the trial production method required for the test.

 Development of Preventing Fuel Debris Spreading• Associated with the fuel debris retrieval work technology for preventing the fuel debris at the bottom of PCV from spreading into
the vent pipes and Suppression Chamber (S/C) etc. was studied in the implementation plan and based on results of investigating the PCV internals.

 Development of Element Technology for Retrieval Device Installation
 Element Technology Development for Work Cells• Technologies for the cell confinement and connecting to the PCVs were compared and clarified.
• The inflatable seals, or the sealing method used to connect the cells to the PCVs, involved identifying the items that would
require confirmation via element tests, and test preparation launched.

Major Research Achievements In-House Research• Exiting technology for the collection of powdery fuel debris was clarified and a conceptual study on applicable technology
conducted.
• Using the assumed collection process for the fuel debris at the bottom of the Primary Containment Vessel (PCV) (including fueldebris generated from cutting process) a conceptual diagram of powdery fuel debris collection was created.

Approx. 10 holes

Pedestal opening

Fig. 3: Element Test of Removal of Interfering
Objects in Pedestal

Biological shield wallremoval device(core boring)

1.8m

Biological shieldwall single test
facility

Fig. 4: Element Test of Biological Shield Wall
Removal Development of Technology for Interfering Object Removal during Fuel Debris Retrieval

Others DataSheet• With a focus on the partial-submersion side access method a processing method that could be used to remove any objects
interfering with reaching the fuel debris at the bottom of PCVs was clarified.
• The element test plan regarding the interfering object removal was substantiated through the results of investigating the interfering objects, and the facility required for the work steps and element test discussed.
• The element test plan for confirming the operational performance of the combination of a robotic arm and access rail was substantiated upon.

 Development of Remote Maintenance Technology for Fuel Debris Retrieval Equipment• The basic idea of remote maintenance, for example the cell-inside equipment of the partial-submersion side access method,
was studied, and the area classification and maintenance equipment clarified.

►Future Developments
The element test plans will be used in the tests that will be prepared and including the production of a test device. The element
test results with respect to the collection of the interfering objects that require retrieval and the processing of the fuel debris will bereflected in the fuel debris retrieval method in thereby increase its feasibility.

Fig. 5: Combined Element Test Image of Robotic Arm and Access RailA conceptual study of the technology needed to collect the powdery fuel debris has already taken place and a development plan
will be formulated as well as the conceptual study substantiated.
18

Annual Research Report 201719

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval
Enclosure

PCV

RPV

Carry-in/out cell
Manipulator

Major Research Achievements Subsidized Projects

Development of Sampling Technology for Retrieval of
Fuel Debris and Internal Structures

Panel house
Manipulator

Acceptance/carry-in/
pretreatment cell

Carry-out cell

Transport containeron the premises►Background►PurposeCollecting the fuel debris from the lower part of the PrimaryContainment Vessel (PCV) and the reactor and identifying
the components and mechanical properties in thereby ensuring appropriate safety management and for use in the facility
design for the fuel debris retrieval are very important. This
project therefore concerns the study of a fuel debris sampling
method as part of the analysis facility and in order to obtain
information that cannot be obtained using a camera.

This project involves the scenario of the realization of theformulation of a sampling survey that will utilize the results ofinvestigating inside the PCVs. It aims at studying the design,
test production, and safety facilities of sampling devices aswell as proceeding with a study of the capability to carry out
collected samples to the analysis facility, and thus obtaindebris information in a safe and prompt manner.

Transportationon thepremises

Arm type accessequipment

Container fortransfer

Debris test sample
Measurement container
Cartridge storage container

Sample storage box

Sampleinjection bottle

Transportation
container

Canister

Measurement container

Transportation
outsidepremises

Fuel debris

Analysis facility

Fig. 1: Sampling Concept for Debris in PCV►Major Approach and Results
Number of reduced links

 Study and Formulation of Fuel Debris Collection/Sampling Scenario
Collection place, amount, and various types of collection devices were considered based on the needs and results of preceding
investigations inside the PCVs.

Major Research Achievements In-House Research

An overall scenario for the fuel debris collection was formulated and a developmental plan considered and updated through thefollowing steps.

Sampling tool and sensor

A development plan for the necessary technology, for example the sampling tools, was formulated.
The safety system during the sampling that depends on the type of collection method was evaluated.

Enhanced motors for each part

 Fuel Debris Sampling System and Device Design/Test production in PCV
The design and test production of the required technology specific to the fuel debris sampling were promoted for the items below
while utilizing the results of other technology developments.

Fig. 2: Example of Access Device Used for Sampling Basic Design of Fuel Debris Sampling System
The system requirements were clarified with respect to ensuring exposure/criticality safety and depending on the collected typesof debris including cylinder shape and pebbly/sandy debris. In addition, this study explored the use of monitoring sensors that
detect neutrons in ensuring the utmost safety with the relatively large cylinder shape debris to be collected, and the technology
needed for the remote handling of the canisters and the transport canisters for transporting the collected samples from the highradiation area at the site (the enclosure in the reactor building) to the analysis facility.

 Design and Test Production of Devices for Accessing Fuel Debris
In order to design the access devices that depend on the type of debris collection, this study explored the versatility of arm typeaccess devices that will make a detailed investigation inside the PCV, a link mechanism reduction for the installation of sampling
tools and sensors, and advanced access device.
After collection

Collection

The technology for use with the pebbly and sandy debris collection was evaluated through a confirmation test and concerning
the elements that included the adhesion, and with both grabbing and scooping being performed and a concept collection device
designed. In addition, the conceptual design of the tools needed for effective collection of the powdery debris generated fromcutting was conducted in an element test.

Gripping type collection tool

Sample container
head

Sample container

Adhesive type collection device
Others DataSheet Design and Test Production of Fuel Debris Sample Collection Device

Cutting test

Sealing test Conceptual Study of Fuel Debris Sampling System in RPV
Based on the status of the development of the internal investigation of the RPV, this project involved studying the concept of thesystem to be used to collect the fuel debris from inside a reactor via access from two directions, namely the upper surface and sidesurface of the reactor building, and with identification of the technical issues and element tests being planned.

►Future Developments
Based on progress made in FY2017 and the results of the latest internal PCV investigation, the necessary tools and an arm typeaccess device toward early debris collection will be developed in FY2018. This project will be developed effectively and promptly
in cooperation with the project of “the Development of Technology for Detailed Investigation inside PCV”.
20

Annual Research Report 2017

Cutting blade

Cutting mechanism
Post-cut powder
collection mechanism

Sealingmechanism

Grating
Sealing mechanism around grating

Cutting collection toolFig. 3: Example Collection Device

21

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval►Background►Purpose

According to the Mid-and-Long-Term Roadmap towards theDecommissioning of the Fukushima Daiichi NPS of the Tokyo
Electric Power Company (TEPCO) Holdings, Inc., the planwith the fuel debris retrieved from the Fukushima Daiichi
NPS is for it to be stored after being removed from the reactor building until such time effective treatment and disposal
methods have been determined. This therefore necessitates
the establishment of storage/collection, transfer, and storage
systems for the fuel debris.

Based on the experience of Three Mile Island Nuclear Power
Generation Station Unit 2 (TMI-2) in the United States, andexisting technology used to transport and store spent fuel,this project aims at developing a fuel debris canister (hereinafter referred to as “canister”) and a canister handling device
for safe and efficient collection, transfer, and storage. This
project mainly aimed at optimizing the specifications andshape of the canister and the canister handling device for the
relevant fuel debris retrieval method in FY2017.

Micro-manometer

Inner diameter: several 10 cm
Height: Approx. 2m Study on Safety Requirements and Specifications and Storage Systems for the Transfer/
Storage of Fuel Debris Canisters
A flowchart of the process to use with the fuel debris retrieval, collection, transfer, and storage (including drying treatment) was
created, and a primary evaluation of the throughput (including amount of required treatments, number of canisters, and storagearea) based on the prerequisites were used in a revised plan about reducing the canisters (Fig. 1). In addition, confirmation took
place that the revision would not currently be necessary based on the latest information from related projects regarding the wet
and dry storage systems for fuel debris that took place in FY2016.

 Development of Safety Evaluation Methods and Safety Validation(Buffer

The basic specifications for the canisters determined in FY2016 were studied with respect to the collection method and with thefuel debris properties taken into account and based on the safety requirements and safety validation results described in 2 and 3
above, and the canister specifications updated.

►Future Developments
The results of studying the fuel debris retrieval and collection that were conducted in the project involving the fuel debris retrieval
method and system, and relevant project on upgrading the fundamental technology, will be reflected in the canister specifications,along with the handling equipment specifications being optimized using an updated evaluation of the safety, handling properties,
and throughput. In addition, the required specifications for the transfer and storage systems for the canisters will be clarified.
22

Annual Research Report 2017

Thermocouple（ユニット缶）（緩衝構造）(Unit can)structure)

Piping for gas/
test water supply（ユニッ
ト缶収納状態）(Unit
cancollection)

Test container
Inner container

Compound gauge

Safety valve
Piping for gas collection
Test water
Spent fuelPellet piece(2) Test equipment configuration

Fig. 1: Basic Draft Canister Plan
One example of the plan of a canister in a mockup test is
shown. The shape will be optimized in cooperation withthe project involving the fuel debris retrieval method andsystem in the future.

Fig. 2: Test Measurement of Amount of Hydrogen
Generated Using Spent Fuel
The amount of hydrogen generated was measured in
tests using fuel debris in thereby confirming the impact
on hydrogen generation with or without the existence ofalpha-rays. The test data obtained will be reflected in thehydrogen generation amount evaluations.

Evaluation of Convection Flow
State in Canister (hatched part)
Cross-section(1)Cross-section(2)Cross-section(3)Canister

Cross-section(4)Fuel debris

Cross-section(5)Cross-section(6)(1) Evaluated Object

Others DataSheet Study on Fuel Debris Collection Method

Micro-manometer

メッシュ
Mesh
（側面
底面）
(Side・
surface/bottom surface)

The Fukushima Daiichi NPS requires validation of the safety of the canisters with respect to criticality, the structures, aged deterioration of materials, and hydrogen generation in thereby ensuring the safe collection, transfer, and storage of the fuel debris,
including the MCCI products which are assumed to contain seawater components at the Fukushima Daiichi NPS, and thus differ
from the case of the TMI-2 fuel debris.
Continuing on from FY2016 this project involved a sub-criticality evaluation (study on the applicability of restricting the amountof water but increasing the inner diameter of the canisters in thereby increasing the workability of the fuel debris retrieval andmitigating the sub-criticality conditions), a study on the canister structure (canister-lid structures with the handling flow and safety
requirements taken into consideration), evaluation of aged deterioration of materials (corrosion evaluation with the environmental
conditions taken into consideration), a study on countermeasures against any hydrogen gas generation (measurement tests of theamount of hydrogen generated from spent fuel (Fig. 2), and convection evaluation with respect to the canisters and an efficiency
evaluation of a hydrogen recombination catalyst (Fig. 3). The necessary knowledge for the establishment of a safety evaluation
method has therefore been gained.

Thermocouple

Major Research Achievements In-House Research

An investigation of the approval and authorization documents of the Rokkasho Nuclear Fuel Reprocessing Facility, Japan Nuclear
Fuel Limited (JNFL), and information on the transportation and storage of damaged fuel at overseas’ facilities (including the Idaho
National Laboratory in U.S.) obtained by the end of FY2016 were reexamined, and can be considered to have contributed to thestudy in section – below. Technical knowledge regarding regulatory safety requirements and management items for the transportation and storage of radioactive waste was also obtained.

Piping for gas
collection

Test container(1)Testing►Major Approach and Results Investigation and Establishment of Research Plans for Transfer and Storage

Safety valve

Piping for gas/
test water supply(Lid structure)（蓋構造）

Major Research Achievements Subsidized Projects

Compound gauge

Development of Technology for Collection, Transfer
and Storage of Fuel Debris(Convection flow at crosssection: cross-section-(3)(Convection flow at eachand above)
cross-section)“In the case of high levels of heat generated with the fuel debris
then UO2 only considered”(2) Evaluation of Convection Flow State (Example)

Fig. 3: Evaluation of Convection Flow State in Canister
Since any hydrogen concentrations generated in the canisters should be lower than an explosive level
thermal flow analysis using numerical fluid dynamics (CFD) will be performed in thereby confirming theconvection flow state within the canisters, and results of the analysis reflected in the efficiency evaluationof hydrogen gas countermeasures, including the allocation of the hydrogen recombination catalyst.

23

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval

Development of Small Neutron Detectors

Major Research Achievements Subsidized Projects

Processor
HUB

Recording device (PC)

►Purpose

Retrieving the fuel debris necessitates identifying the position and amount. The neutron measurement method, whichmakes measurements using the neutrons generated from thespontaneous fission of elements in the fuel debris, is one of
the methods for retrieving the fuel debris. However, existing
sensors may not be capable of being used due to the dimensional restrictions, thus making a smaller censor necessary.

The weak neutrons need to be measured under the highdose gamma-rays by accessing the vicinity of the fuel debris
with a detector in order to actually detect the fuel debris.
However, the route used to access the vicinity of the fueldebris is rather narrow, thereby making a smaller sensor
necessary. This project aims at developing a small neutron
detector that will satisfy that requirement.

Gammaray source(Co-60)

Number ofelements►Major Approach and Results Specifying the neutron detection technology applicable to the fuel debris retrieval at
the Fukushima Daiichi NPS and element test survey results of its feasibility (Phase-1)

Sensor part expansion

Neutron-ray source(Pu-Be or Cf-252)

Fig. 1: Layout of Performance Validation Test

►Future Developments
Following Phase-2, the production of the sensor as well as the sensor unit and performance evaluations will be conducted in thecompletion of a small  neutron detector by the end of the second quarter of FY2018. More specifically, a survey of an applicationin which a smaller neutron detector when compared with the existing one, and the required specifications determined. After completion of this project the decommissioning activities at Fukushima Daiichi NPS, including a detailed investigation inside the PCVs
and monitoring during the fuel debris retrieval etc., will be substantiated for application in the project.
24

Annual Research Report 2017

600

800

1000

1000
100

Neutron count (cp)

Neutron count (cps)

105 Gy/h

0.1
0.01
0.001

10

772 Gy/h
983 Gy/h1
0.1

Others DataSheet

The test production of a small neutron detector was commenced upon based on the results of Phase-1.
The  sensor design and production were commenced upon; and more specifically, the design of a CMOS sensor, the sensor
installation, a study on the performance validation method using a single CMOS sensor, and the design of peripheral equipment
and software.
The test production and evaluation of the sensor unit were commenced upon with respect to the basic design of the sensor unit,
the design of the peripheral equipment, and the planning of the performance evaluation tests.

400

Fig. 2: Detection Performance at Single
Gamma-ray Location• Straight line shows regression
1 • Error bar shows detection
range of multiple elements

In order to study the capabilities of the feasible neutron detector and the specifications the results of a performance validation
via element tests were clarified, and the capabilities and specifications of the neutron detector in actual use considered.

 Performance Validation of Element Technology

200

CMOS (Number of elements differ in each test. 2, 4 and 6)

Phase-1 involves the applicability to Fukushima Daiichi being evaluated after element tests.

The study involved the establishment of a neutron detector development plan, and discussing the relevant schedule up to
the production period toward practical application, the cost, the delivery due date, and the handling of the developed detector (Availability of measurements by a plant operator and the necessity of preliminary training, etc.) Practical application was
thereby determined.

0

Gamma-ray dose rate (Gy/h) Other Phase-1 Studies

With respect to the conceptual design of the neutron detector the use of a neutron measurement system was assumed in an
investigation of the outside of the Unit 1 pedestal (a method to where a sensor descends from the grating on the 1st floor), anda concept neutron detector designed.

30
25
20
15
10
5
0

Test conditions• Neutron flux 0n/(cm2 ·s)• Gamma-ray dose rate:
63–983 Gy/h Max. total pixels
approx. 31% of• Irradiation time: 5min.
strength

Major Research Achievements In-House Research

Element test of a CMOS semiconductor type neutron detector was performed and its performance verified (Fig. 1).
Basic property tests were conducted at a single neutron location in confirming that the sensor was capable of obtaining theneutron flux from the neutron count (Fig. 3). In addition, confirmation also took place that the signal strength obtained against
the gamma-rays was linear in the single gamma-ray location (Fig. 2).
The detection performance under the combined irradiated environment of gamma-rays and neutron-rays was confirmed in
identifying the cluster pattern using alpha-rays derived from neutrons in the combined location (Fig. 4).
The results of the radiation resistance evaluations confirmed that there would be no incorrect detection of neutrons up to the
cumulative gamma-ray dose, or degree of 1000Gy. In addition, the results of irradiating a single neutron revealed no remarkableincrease in the count, even with irradiation of more than 6.5e+6n/(cm2·s)·s.
The long distance signal transmission properties confirmed that transmission could take place over more than 50m because theneutron count with a 50m long cable was almost the same as that of a 16m cable length.
In confirming the effect of the reaction layer neutrons were irradiated at the sensor without a 6Li coating, which is the reactive
layer of the sensor. The results confirmed that neutrons could basically not be detected practically if there were no reaction layer.
The results of measuring the neutron flux with the direction of the sensor normal line and neutron generation in evaluating thedirectional dependency revealed that the angle of the sensor face provided no significant change up to 45 degrees. However,the neutron flux did decrease beyond that.

Moderating
material

Number ofelements

Total of gray value for eachpixel

►Background

0.01
0.1

1

10

100

1000

Neutron flux (n/(cm2/s))
Fig. 3: Detection Performance with Single
Neutron Location

0.1

1

10

100

1000

Neutron flux (n/(cm2/s))
Fig. 4: Detection Performance with Gamma-Ray/
Neutron-Ray Combined Locations

25

Major Research Achievements–Subsidized Projects

R&D for Fuel Debris Retrieval

Unit 1

Unit 2

Unit 3

Reactor • Submerging of
remnant fuelcore

MinimalMediumLow• Submerging of

RPV lower debris• State variation
part

during retrieval►Background►Purpose

The assumption is that the fuel debris has not currently
reached criticality. The development of a criticality control
method is underway in thereby ensuring criticality can be prevented in which the shape of the fuel debris and the change
in the amount of water are estimated for during the fuel debris
retrieval work in the future, as well as the termination of any
criticality in a safe manner, even if it does occur.

An impact evaluation method for use with criticality shall beestablished via the formulation of criticality scenarios and by
developing a behavior evaluation method during a criticality
situation. Moreover, criticality prevention technology using
neutron-absorbing material with soluble/insoluble properties
and criticality monitoring technology, including technology forcriticality approach detection and recriticality detection, will
be developed and their technical feasibility confirmed.

CRD• Submerging of
Housing attached debris
PCVbottompart(almost no
remnant fuel)Exposure via leakage(Fuel may remain (Fuel may remain
in/around thein periphery)reactor core.)

Sky shine

Direct line

Worker on
operating floor

Submerging: Low Submerging: Medium Submerging: Medium
Retrieval: Minimal Retrieval: Low Retrieval: Low(Small remnants)(Large remnant,exposure)Gas
controlsystem(Large remnant,exposure)

Low ~ Minimal Low ~ Minimal Low ~ Minimal(attached shapeand amount)(attached shapeand amount)

Partial-submersion
methodSubmersion method(attached shapeand amount)

Exposure via
environmental
release• Submerging of
Submerging: Low Submerging: Medium Submerging: Low
exposed debris
Retrieval: Low Retrieval: Low Retrieval: Low• State variation(large residue,(Large remnant, (Large remnant,during retrieval
large exposure)small exposure)(including curling) small exposure)

Table 1: Degree of Importance for Criticality Control by
Units and Sections (relative evaluation)

The concrete deployment of criticality prevention and criticality
approach detection technologies are being studied depending on
their degree of importance in controlling criticality.

►Major Approach and Results

Major Research Achievements Subsidized Projects

Development of Technology for Criticality Control
Methods

Sections Major shape change

A workeroutside thebuilding

PublicGrand shine

Fig. 1: Exposure Evaluation Model
These models are used in studying the specifications of the safetymeasure facilities and are reflected in designs in thereby avoiding any
excessive exposure in the case of criticality occurring“B-10 Neutron Detector”
Reflector cell

 Establishment of Criticality Evaluation Method

Fuel assembly cell

A behavior evaluation method during criticality via an exposure evaluation model was developed, and an evaluation method that
could be used to evaluate the behavior from the occurrence of criticality through to its termination and the impact of any exposure
to the public and workers established (Fig. 1). They will be utilized in studying the procedures to use to mitigate any impact in thecase criticality occurs and safety measures to use at facilities in the future.

Control rod cell
Neutron detector

 Development of Criticality Approach Monitoring Method“Appearance of KUCA Test”

The development of subcriticality measurement technology that is based on neutron measurements has been promoted.
Operational verification tests of the B-10 neutron detector in a radiation environment were performed in FY2016 in order to confirm
its feasibility with the site environment, with the discrimination of neutron-rays and gamma-rays confirming their feasibility whenused with appropriate shielding. A reactor core, which simulates the various fuel debris conditions, was configured at the Kyoto
University Critical Assembly (KUCA) laboratory in FY2017, and the degree of subcriticality evaluated using the reactor noise
method then confirmed the estimate from the neutron signals obtained (Fig. 2). The feasibility of the criticality approach monitoring
technology was thus confirmed by the neutron measurements.

This technology is based on gamma-ray measurements using a germanium (Ge) detector and the gamma-ray energy measurement range was increased to 3MeV in order to be capable of collecting gamma-ray data at the Fukushima Daiichi Unit 1 whichis equipped with a Ge detector. The results confirmed that the objective nuclides had no impact in the high gamma-ray energy
regions and the peak of Kr-88 (Fig. 4). This study confirmed the possibility of early criticality detection using the enhanced gas
control system.

 Development of Criticality Prevention Technology
Neutron absorbing materials for use in preventing criticality from occurring that are of the water soluble/non-soluble type havebeen developed.

In addition, the soluble neutron absorbent material (sodium pentaborate) confirmed the required boron concentration estimate to
be about 6,000ppm as a conservative assumption. This estimate exceeds the concentration proven to exist at the power generation reactor and hence a nuclear characteristic verification test was performed at the KUCA in thereby confirming that the nuclear
properties could be evaluated with the same accuracy as previous. Moreover, the basic specifications of the boron concentration
maintenance facility were examined. Herewith, the prospect of boron concentration evaluations and the technical feasibility of thenecessary facilities, including the concentration maintenance facility, were obtained.

►Future Developments
Prospective establishment of a criticality evaluation method for use with the fuel debris retrieval and feasible criticality control
technology were obtained, and the results will be utilized in studying the optimal practical implementation of a fuel debris retrieval
device and relevant systems, and thus ensuring greater safety in the future.
26

Annual Research Report 2017

KUCA can be used with variable amounts of fuel and moderator according to the purpose of the test and to simulate
various fuel debris conditions and hence was used in the criticality approach detection and nuclear characteristic tests.

Rare gas

For overallmonitoringGas control
systemFor local
monitoring
Neutrondetector

Xe-135Rare gases, etc.
Rare gas

Gas chamber

Fig. 3: Overview of Criticality Monitoring
Localized criticality will be monitored by the neutron detector andentire PCV monitored by the gas control system (Gamma-ray
monitoring of rare gases).
Liquid→Solid
Solid(Solidified material)

B/Gd containing
glass

Glass material image
after formation

Gd2O3 particles

Kr-88

Cs-137
Cs-134
Kr-85

Kr-88 detection isavailable in highenergy region.

Kr-88
Kr-87

Shield (lead)“Image of Gamma-Ray Measurements by
Gas Control System”

B4C sinteredmetal material

Kr-87

Granulated cement/Gd2O3 powder(Left; before hardening: Right; after
hardening)

Granulated liquid glass/Gd2O3
powder (after hardening)

Measurement range of existing radiation monitor

Gamma-ray energy(keV)

Fig. 4: Result of Gamma-Ray Measurement Tests
by Gas Control System
Absorption material thatspreads over the surfaceof rock shape simulated
debris with unevenness
in the water

Others DataSheetWith respect to the non-soluble absorbent material nuclear characteristic verification tests at the KUCA, dissolution tests withlong-term exposure, and performance workability tests (Fig. 6) took place using candidate materials selected by their fundamental
physical properties and radiation resistance performance tests (Fig. 5), and a prospect absorbent material candidate obtained.

Fig. 2: Criticality Approach Test at Kyoto University Criticality Assembly (KUCA) Laboratory Device

Ge detector

The gas control system used to monitor slight amounts of FP gas concentrations present in the Primary Containment Vessel(PCV) (Fig. 3) was improved for use in detecting any recriticality in the early stages, and Kr-88 detection technology, which canbe used for prompt responses during criticality in addition to the Xe-135 technology that is currently being used in monitoring, isbeing developed.

“Image of Neutron Signals”“Example Test Reactor Core”

Number of counts

 Development of Recriticality Detection Technology

Major Research Achievements In-House Research

Based on the estimated information on the fuel debris distributions obtained from the results of evaluations and investigations
inside the reactor, and muon measurements, the relative importance of criticality control while the fuel debris is being retrieved
from each unit was evaluated (Table 1). The results will be used in the studying the concrete deployment of technologies for use
in both monitoring and preventing any criticality from occurring.

Absorption material
flowing down to granularsimulated debris in water

Underwater hardening resin/Gd2O3
powder (Left; before hardening: Right;
after hardening)

Fig. 5: Appearance of Insoluble Neutron Absorption
Material (candidate material)

Fig. 6: Example of Workability Verification Test of Water
Glass Type Absorption Material27

Major Research Achievements–Subsidized Projects►Background►Purpose

Due to the severe events caused by the Great East Japan
Earthquake the Reactor Pressure Vessel (RPV) and PrimaryContainment Vessel (PCV) at the Fukushima Daiichi NPS
were affected by high temperatures, seawater, and falling
fuel debris and thus concerns about the material further deteriorating. Measures to maintain the structural integrity of the
PCV/RPV are thus required over the long term that retrieving
the fuel debris from the reactor core will take.

This project aims at clarifying what important equipment in
RPVs/PCVs got damaged in the event of the large-scale
earthquake and the spread of the impact on the water levelsin the PCVs that can be assumed during the debris retrieval
work and installation of important equipment in the building.
Countermeasures to prevent or suppress that impact will
be developed and its efficacy confirmed through seismic
assessments.

PCV

Basic test data
• Thermal degradation at
the time of the accident,
assumed erosion• Long-term erosion
prediction after theaccident

Seismic response analysis(building – equipment coupled)
Time history elasticity/elastic-plastic analysis(RPV pedestal, S/C support column)

Reactorshieldingwall
PCV stabilizer

Air lock forworkers
Vent pipe

Fig. 1: Overall Development Flow of Seismic
Analysis Method

Vent header

Downcomer

Vent pipe
1500mmMajor Research Achievements In-House Research

Vent header

745mm

Downcomer
burial height
S/C Development of Seismic Resistance/Impact Evaluation Methods in Establishing Safety
Scenarios

Downcomer
Seismic resistance support

Column support

Concrete injection conditions
in S/C with water stopped in
downcomer assumption

Fig. 3: Vent Pipe–S/C System Coupled Analysis Conditions andTime History Seismic Response Analysis Model (Unit 2–3)

The development of seismic resistance and impact evaluation methods  has been promoted based on the flow and feedback
shown in Fig. 1.

 Development of Seismic Resistance/Impact Evaluation Methods for Suppression Chamber (S/C) Support Column

Suppression Chamber

Dry wellColumn support

Vent pipezySeismic resistance
supportVent headerx

The data obtained uses assumptions about the pedestal temperature history and its distribution in the development of the following evaluation method for erosion impact evaluation from the fuel debris and material data.

• The earthquake response impact evaluation method uses the building/large equipment system coupled model (Fig. 2) in which
the stiffness degradation is a parameter that assumes the erosion in the RPV pedestal.

Displacement

Elasticity/elastic-plastic
properties (image)

Bonds

Vent pipe

Safety scenarios (safety functions that maintain or can be used in responding to restoration after accidents have occurred) were
established based on countermeasures for the facility which will be performed before the start of any fuel debris retrieval and asa swift preparatory response to the potential risk resulting from any damage to major equipment due to a large-scale earthquake.

• Simplified evaluation where the amount of erosion is a parameter and the detailed evaluation method uses the 3D FEM elastic
analysis model shown in Fig. 5.

Foundationof mainreactorstructure

Fig. 2: Reactor Building/Large Equipment
System Coupled Model

1/2 sector solid
component model Establishment of Safety Scenarios during Large-Scale Earthquakes

 Development of Seismic Resistance/Impact Evaluation Methods for Pedestals

RPVstabilizer

RPV pedestal

Evaluation/
countermeasure study►Major Approach and Results

In order to evaluate seismic resistance while stopping the water in the downcomer by injecting filling into the S/C coupled analysis model (Fig. 3) connected with the vent pipes and the S/C systems in Units 2–3 was created and time-based history seismic
response elastic analysis performed. The important evaluation parts: the column support, the seismic resistance support, and thevent header were all evaluated via elastic analysis as well as a detailed FEM model created in confirming the downcomer burial
conditions.

RPV

Major Research Achievements Subsidized Projects

Development of Seismic Resistance and Impact
Evaluation Methods for RPV / PCV

Reactor Building

Analysis model/
conditions

Load

R&D for Fuel Debris Retrieval

Downcomer

Detailed FEM analysis model of S/C system

Detailed FEM analysis model of vent system

Fig. 4: Detailed Analysis Model of Vent Pipe for Elastic-Plastic Analysis and S/C System (Unit 2–3)
CRD opening• High temperature erosion and strength degradation amounts for use in assuming the strength degradation of reinforcing steel
in concrete.

CRD opening

Others DataSheet Upgrading of Safety Scenarios
CRD opening

A confirmation method that includes analysis and tests was studied for the upgrading of the evaluation method above . The
following detailed analysis and material tests were conducted.

►Future Developments
The safety scenarios will be reviewed according to the progress status of related projects (fuel debris retrieval system, PCV internal survey, etc.).
28

Annual Research Report 2017

Internal• Acquisition of PCV material test data through taking the temperature history at the time of the accident into consideration.

External• Evaluation via time history elastic seismic response analysis of the S/C support column in Unit 1.

CRD opening

Cross-sectional
drawing

Centralopening

Lower Access
Opening PartFEM analysis model of 1F-Unit 1

Internal only
deficiency(element reduction)

Deficiency range,
inside/outside(free boundary)

Modeling of debris erosion parts

Fig. 5: Development of Seismic Evaluation Method for Pedestal due to Debris Erosion29

Major Research Achievements–Subsidized Projects

The fuel debris will be retrieved from the Fukushima Daiichi
NPS in addition to the debris etc. that has already been
stored and hence a large amount of waste can be estimatedto be generated in the future. Because of this situation, methods to store, treat, and dispose of the waste in the implementation of the project will need to be developed in parallel with
their characterization.

This project aims at developing technology for the safe treatment and disposal of the solid waste generated in the accident by integrating the results of Research and Development(R&D) resulting from characterization of an inventory evaluation based on waste analysis and a study of solidification
technology, the stabilization of secondary waste generated
from the water treatment, and concepts of the treatment andsafety evaluation method used.

►Major Approach and Results

14
12
10

1.5E-018

1.0E-0164

5.0E-022

Data frequency
Probability density(Normal logarithmic
distribution)

Data frequency

Probability density

16

2.0E-01

0

3.
16
.E
-0
1.
4
00
.E
-0
3.
4
16
.E
1. -03
00
.E
-0
3.
2
16
.E
-0
1.
2
00
.E
-0
1

►Purpose18

2.5E-01

0.0E+00

Transport ratio (Compared to Cs-137)

101

1.E+11

Nuclide concentration (Bqt)►Background

Unit 1
Unit 2
Unit 3103

20

3.0E-01

Major Research Achievements Subsidized Projects

R&D for Treatment and Disposal of Solid Radioactive
Waste

Radioactive concentration of
Sr-90(Bq/g)

R&D for Treatment and Disposal of Radioactive Waste

10-1
10-2

10-3
100

10-4

(Outlined characters show
lower detection limits)
102

104

106

Radioactive concentration of
Cs-137(Bq/g)

Explanatory notePu-2381.E+095%ile

1.E+07

medlanaverage

1.E+0595%ile1.E+03

Analysis resultsfrom actual waste(Concentration)

1.E+01

0

0.05

0.1

0.15

0.2

Probability density

Fig. 2: Examples of Analysis Value Distribution(Upper Drawing) and Inventory
Assumption (Lower Drawing)

Fig. 1: Correlation of Sr-90 and Cs-137
in Rubble in R/B

 Characterization

Major Research Achievements In-House Research• Production of glass solidification of secondary
waste generated in the simulated water disposal(mixed zeolite, sludge from the decontamination
device, and additives such as B2O3)

Analysis of the rubble, contaminated water, and secondary waste generated from the water treatment continues to takes placein amassing data. In addition, the concentration of various nuclides is compared to Cs-137 in thus investigating contamination
behavior using the analysis data obtained so far. The results revealed that the Sr-90 and Cs-137 concentrations indicate a correlation with the rubble samples in the reactor building, regardless of the unit (Fig. 1), with the same behavior being observed withthe other nuclides. Furthermore, the contamination frequency distribution of radioactive nuclides indicates the normal logarithmic
distribution used with the amount of radioactivity (inventory), including the radioactive waste, and which is being estimated (Fig. 2).

Fig. 3: Glass Solidification Tests Using
Simulated Waste

 Preliminary Management
 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 in the water
treatment in-drum type glass solidification technology was focused upon from the aspect of contamination dispersion prevention
and equipment size. The solidification results from simultaneous melting with zeolite being used for the contaminated water treatment and other simulated waste from the secondary waste generated by the water treatment (Fig. 3). Basic data, which includes
chemical analysis and a leaching test from the solidification, was collected, and the composition and conditions of the solidification
thus provided.

 Study on Decontaminant Device Sludge Stabilization (including sample collection)

 Study of Concepts of Disposal Applicable to Solid Waste and Safety Evaluation Method

 Integration of R&D Results
The waste stream was reflected in the latest results obtained from existing research and the establishment of a method of ensuring consistency in the progress made, results, and remaining issues commenced upon.

Actual collected sludge(10mL vial container)

Storage layer-D

Ground water
level
B1F

Underwater
cameraApprox. 40cm

B2F
West – East

Overview of investigation insidestorage layer-D (Cross-section)

The condition of sludge at thebottom of storage layer-D (Yellow)

Fig. 4: Investigation inside Main Process Building Storage Layer-D andCollection of Sludge from Decontamination Device
Acceptanceof widevariety of
waste

Increasing
the capacityof disposal
facility

Correspondence
to risk
uncertainties

Example, Large
container of mixed
waste

Example, Large vault

Example, Human
entrance prohibition

Others DataSheet

A detailed survey (Fig.5) of overseas repositories: the U.K. Low Level Radioactive Waste Repository (LLWR), the Swedish Final
Repository for low and intermediate level short-life radioactive waste (SFR), the U.S. Waste Control Specialist (WCS) for low level
radioactive waste disposal, among others, was conducted in the development of information to use with the concepts of applicablesolid waste disposal and safety evaluation method, as well as the applicability to the solid waste at the Fukushima Daiichi NPS
being studied based on various case scenarios and with domestic disposal cases and the characterization of the solid waste taken
into consideration.

Ground level

Outer wall

The internal conditions of reservoir tank D in the main process building where sludge generated from the decontaminant device
has been stored were surveyed in collecting some actual sludge (Fig. 4). An underwater camera revealed the thickness of theaccumulated sludge layer to be approximately 40cm, and hence its volume was estimated to be 37m3. In addition, basic data
related to fluidity was obtained using simulated sludge in the study on the sludge retrieval.

Solid content

Aboveground floor:
Decontaminant(AREVA) installation
Worker1F area►Future Developments
Analysis will be continued to be reflected in improving the inventory estimation model as well as tests related to the stabilizationof the secondary waste generated by the water treatment. In addition, a detailed survey of overseas’ disposal concepts and theevaluation method, and evidence and background will continue, and applicability to the solid waste treatment at the Fukushima
Daiichi NPS reviewed. The above results will then be examined with respect to the waste stream.
30

Annual Research Report 2017

Fig. 5: Detailed Survey of Overseas Repositories
31

Major Research Achievements–In-House Research

Others–DataSheet

Major Research Results in FY2017►Background and Purpose

No.

Date

Details

The 50th JAIF Annual Convention 2017

Apr. 11, 2017

Robotics Technology Utilization

2

Special Lecture on Nuclear Reactor Decommissioning, Tokyo
Institute of Technology

Apr. 21, 2017

Current State of Fukushima Daiichi NPS and Road to Reactor
Decommissioning

3►Major Approach and Results2017 International Congress on Advances in Nuclear Power
Robot Exhibition
Apr. 26 – 28, 2017
Plants (ICAPP 2017)
Lecture: Overview of IRID R&D - Focusing on Debris Investigation

4

Societe Francaise d’Energie Nucleaire, Les accidents graves(French Nuclear Society, The severe accidents)

May 10, 2017

Fukushima Daiichi: Clarification of the accident progression,
current plant status and R&D programs in Japan

5

Electrical Review 2017, Summer Special Issue

May 12, 2017

Development of Inside Survey Equipment for Nuclear Reactor
Containment Vessel– Shape Variation Type Robot [PMORPH] –6[The contemporary Science and Technology Giant Seminar “TheInspiration of Knowledge”], Kyoto University

May 17, 2017

Working for Radioactivity

7

NSC2017 (Chernobyl Workshop)

May 18, 2017

Fuel debris and MCCI product characterization for Fukushima
Daiichi NPS

8

The 42nd Task Team Meeting for Decommissioning / Polluted
Water Control / Secretariat Conference

May 25, 2017

Analysis of Solid Waste Sample, Fukushima Daiichi NPS(Accomplishment report to the present situation), etc. 2 cases

9

Robotics Consortium, Shibaura Institute of Technology

May 26, 2017

Future of Decommissioning and Robotics Technology

10

National Institutes for Quantum and Radiological Science andTechnology (QST)

Jun. 1, 2017

Impact Evaluation of Gamma-Ray Exposure Exerting the Local
Erosion of Carbon Steel in the Diluted Artificial Seawater in which
Rust-Proof Agent is Added, etc. 2 cases

11

NPO Japan Association for Automation Advancement General
Meeting

Jun. 3, 2017

Future of Decommissioning and Robotics Technology

12

Professional Meeting for Basic Study Related to the Reliability
Keeping of Containment Vessel / Building and Waste Disposal for
Decommissioning Action and Core Human Resource Cultivation Program

Jun. 12, 2017

Proof Strength Evaluation of Severe Accident Experienced
Reinforced Concrete Structure

Development of Remote Collaboration Motion Control
System13

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Jun. 12, 2017

Development of the model for advection and diffusion of eroded
concrete into debris in molten core-concrete interactions

14

The 30th Regular Seminar of [Water Chemistry Committee],
Atomic Energy Society of Japan

Jun. 23, 2017

Impact Evaluation on Rust-Proof Agent Exerting the Local Erosionof Carbon Steel in the Diluted Artificial Seawater►Background and Purpose

15

FY2017 Summer Special Issue, [Electrical Review]Jun. 30, 2017

Front Line of Disaster Corresponding Robot– Prospect at Flammable Gas Generation and Approach to theFirefighting Activity at Large-Scale Fire –The dry ice blast decontamination device for high placesand decontamination device for the upper floors that weredeveloped using the FY2013 supplementary budget consistof multiple crawlers, with each of the carts being operated byoperators.

16

Knowledge Innovation Research Group Meeting, Knowledge
Management Society of Japan

Jun. 30, 2017

Overview of Activities of International Research Institute for
Nuclear Decommissioning

17

ACTINIDES 2017, Sendai
Progress in Nuclear Science and Technology

Jul. 2017

Characterization of the VULCANO test products for fue debris
removal from the Fukushima Daiichi Nuclear Power Plant

18

International Trade Fair [1 FPEX 2017] Hosted by Japan Fluid
Power Association

Jul. 2017

Water moves machine!
Safety of food! Safety of person! Safety of town! Safety ofCountry!

19

The Second International Forum of the Fukushima Daiichi
Decommissioning

Jul. 2 – 3, 2017

Estimation of Situation inside the Reactor Core and PCV
Fukushima Daiichi NPS

20International Conference Actinides 2017; and Progress in Nuclear
Science and Engineering

Jul. 10, 2017

Chemical state analysis of simulated corium debris by EXAFS,
etc. 2 cases

21

The 16th Materials Division Summer Seminar, 2017; Atomic Energy
Society of Japan

July 12, 2017

Development of Technology for Corrosion Control of RPV/ PCV
in 1F

22

The 14th Academic Conference, Japan Society of Maintenology

Aug. 2, 2017

Development and Verification of Investigation Device inside PCV– Basement Survey by Shape Variation Type Robot [PMORPH] –23

IRID Symposium 2017 in Iwaki

Aug. 3, 2017

Conceptual Description of Element Technology for upgrading
Fundamental Technology

24[The 14th Nuclear Power Technology Summer Seminar], Nuclear
Power Group Meeting

Aug. 4, 2017

Approach to Domestic and Foreign Nuclear Power 

25Technical Information Center Seminar

Aug. 22, 2017

The Latest Technique Development of Nuclear Power Plant Robot
toward Decommissioning/ Convergence Support

26

Study on Characterization of Fuel Debris and Human Resource
Cultivation Conference (ConFDeC2017)

Sep. 6, 2017

Current Status of Fukushima Daiichi NPS and R&D for Fuel
Debris Retrieval

27

Creative Human Resource Cultivation Program through Basic
Study on Decommissioning– Transdisciplinary Challenge from Fukushima Using InterCollege Network –FY2017 Section Meeting and the 1st Research Report Meeting

Sep. 6, 2017

R&D Status of International Research Institute for Nuclear
Decommissioning

Muon measurements, which can penetrate the reactor building, confirmed major structures such as the shielding concreteof the Primary Containment Vessel (PCV) peripheral, the spent
fuel pool, and the reactor building wall.

 Evaluation Results of Unit 3

The results of evaluating the fuel debris distribution in the
RPV through measurements made with the muon penetration
method are as follows.
 No mass of fuel debris is present in the original reactor coreregion.
 Fuel debris may remain at the bottom of the RPV, although
uncertain.

Reactor building

Estimated positionof reactor core

West

South

Spent
Fuel Pool
(SFP)

East
Measurement
deviceMeasurement
deviceFig. 1: Image of Muon Measurements
Fig. 2: Image of Muon Measurements
Penetrating Reactor Building (EastPenetrating Reactor Building (SouthWest Cross Sectional Drawing)North Cross Sectional Drawing)
Building wall
(Major steam isolation valve room)
RPV
Spent
Fuel Pool

Reactorcore region
PCV
peripheral
concrete(As of September 8, 2017)
Building wall
(Major steam isolation RPV
valve room)
ReactorSpentcore region
Fuel Pool
PCV
peripheral
concrete

High placeon mountainside Summary

The results of the evaluation will be utilized in the approach
used with the fuel debris retrieval in the future in addition to otherexpertise obtained from investigations made inside the PCV.

North

Turbinebuilding

Fig. 3: Evaluation of Distribution ofaa
Substance Quantity (density length) valuation of Distribution of
via Simulation (Case of fuel debris Fig. 4: E
Substance Quantity (density length)
being present at the reactor coreusing Muon Measurements
and at the bottom of the reactor)

This research aims at developing the technology of an automatic tracking vehicle that following carts can automaticallyuse with respect to the route to take in order to reduce thenumber of required operators.

►Major Approach and Results

Photo 1: Decontamination Device for Upper Floors

 Operational Test of Route Tracking
Testing of the route tracking on a test course that was a simulation of the reactor building 1st floor (straight, narrow areas,
avoiding obstacles, cornering, and searching for the lead cart)
confirmed that the route tracking operation had no problems.

 Future Developments
In the case of use of decommissioning devices that consistof multiple carts on-site a remote collaboration motion control
system will be developed in thereby reducing the number ofnecessary operators.

Photo 2: Testing of This Research
Annual Research Report 2017

Others DataSheet

1

Major Research Achievements In-House Research

As an approach to identifying the conditions inside the reactor and toward retrieving the fuel debris muon transmission measurements* have been made in thereby identifying the distribution of substance quantities in the Reactor Pressure Vessel (RPV) using
the penetration rate of muons that could penetrate the reactor as a method of obtaining information on the fuel debris distribution
at the Fukushima Daiichi NPS Units. The same measurements and evaluations were conducted for Unit 3.
* The measurements were developed using technology of the High Energy Accelerator Research Organization

 Measured Results at Unit 3

32

Presented at / byMajor Research Achievements Subsidized Projects

Measurement and Evaluation of Fuel Debris Distribution
inside Reactor at Fukushima Daiichi NPS Unit 333

Others–DataSheet
Presented at / byDate

Details28

Decommissioning Study / Human Resource Cultivation Enhancement
Program, FY2017 Summer School, University of Tokyo

Sep. 6, 2017

Remote Technology for Decommissioning and its Design
Methodology, etc. 2 cases

29

The 3rd Asian Nuclear Fuel Society

Sep. 8, 2017

A study of cesium chemisorption onto surface of Stainless Steel30

The 35th Japan Robotics Society

Sep. 11, 2017

Robotics Technology for Decommissioning and Expectation to
Academy, etc. 3 cases

Sep. 13, 2017

Filling Water Stop in S/C(1) Overview Related to 1/1 Scale Actual Size Filling Test
(2) R elationship between Stiffening Ring Trans-lesion andConcrete Mix (Mix-A, Mix-B)(3) Placing up State and Finishing State regarding 1/1 Full Scale
Filling Test
(4) Concrete Core regarding 1/1 Scale Full Scale Filling Test
(5) Water Stopping Performance of Damaged Hole, Quencher
and Strainer regarding 1/1 Full Scale Filling Test

31

32

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Development of Fuel Debris Criticality Control Technique (29) andCriticality Control Policy, etc. 12 cases

332017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Chemical Reaction Evaluation on Cs and Steel Member at LightWater Reactor Serious Accident(1) E valuation on Speed of Reaction between Cs and Steel
Member toward Modeling2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Development of Housing-Can for Fuel Debris(1) Colledtion/ Transfer/ Storage Techniques for Fuel Debris, etc.
10 cases35

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Study on Identifying of Conditions inside Reactor by Severe
Accident Analysis Code MAAP
(17) Detailed Analysis on the Accident Progression of Fukushima
Daiichi Unit 2 by MAAP, etc. 2 cases

Sep. 13, 2017

Development of Poly-Nuclear Species Analytical Method by ICPQQQ-MS
Development of simultaneous measurement for multi isotopes by
ICP-QQQ-MS

37

2017 Fall Meeting, Atomic Energy Society of Japan2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Review on the Preprocessing for Dry Type Storage of Retrieved
Fuel Debris– Evaluation on the Drying Property Focused on the Shape of Fuel Debris –Pre-treatment of defueling Fuel Debris for drying storage– Evaluation of Drying Characterization focused on form for Fuel Debris –Presented at / byDate

Details51

Decommissioning Technique, Radioactive Waste Management andNuclear Facility Decommissioning Technology Center

Sep. 20, 2017

Current State of R&D for Fuel Debris Retrieval Technique toward
the Decommissioning of Fukushima Daiichi NPS– Overview of R&D Approached by International ResearchInstitute for Nuclear Decommissioning (IRID) –52[Introduction to Nuclear Decommissioning Engineering/ Special
Topics], Intensive Course of Tohoku University

Sep. 23, 2017

D e v e l o p m e n t o f R o b o t Te c h n i q u e A s s o c i a t e d w i t h
Decommissioning Work and State of Site Application

53

Kansai Nuclear Conference[Research Committee on the Upgrading Technique of Atomic Fuel]

Oct. 11, 2017

IRID R&D State toward Fuel Debris Retrieval

54

Academic Journal; The Institute of Systems, Control andInformation Engineers (ISCIE)

Oct. 15, 2017

Introduction to ROS for Control Researchers

55

The 34th Japan-Korea Nuclear Experts Meeting

Oct. 16, 2017

Overview of IRID R&D Focusing on Debris Retrievel

56

The 2nd Committee on Decommissioning Ground Engineering,2017

Oct. 16, 2017

R&D on Treatment and Disposal of Solid Wastes from Fukushima
Daiichi NPS

57

MIT-Tokyo Institute of Technology Innovative Nuclear Event
System (MT-INES) Program

Oct. 19, 2017

R&D activityes for Fukushima Daiichi Nuclear Power Station
Decommissioning

58

The 64th Materials and Environmental Discussion Session (Japan
Society of Corrosion Engineering)

Oct. 24, 2017

Criticality Prevention Measures at Fuel Debris Retrieval by Using
Neutron Absorption Material, and Impact on Erosion

59

MIT-Tokyo Tech workshop on Innovative Nuclear Energy Systems(TM-INES)

Oct. 27, 2017

Restoration and decommissioning activities for Fukushima Daiichi
Nuclear Power Station by Toshiba60

2017 IEEE/SICS International Symposium on System Integration(SII 2017)

Oct. 30, 2017

Outcome of [Cooperative Movement Control of Multiple Carts]

61

The University of Tokyo-IRID workshop

Oct. 30, 2017

Current State and the Issues of R&D toward Fukushima Daiichi
Decommissioning, etc. 2 cases

62

The 64th Materials and Environmental Discussion Session, Japan
Society of Corrosion Engineering

Nov. 8, 2017

Development of RPV/PCV Erosion Suppression Technique for
Fukushima Daiichi NPS: About Selection Result and Management
Procedures of Rust-Proof Agent, etc. 3 cases

63

Robot Festival Fukushima, 2017

Nov. 9, 2017

State of R&D Approached by IRID
[Survey/ Retrieval of Fuel Debris by Robot]

64[Outcome and Issues of Robot Technique] Hosted by Expert
Committee of Production and Management Knowledge, The
Japan Society for Precision Engineering

Nov. 13, 2017

Decommissioning of Fukushima Daiichi and Robot Technique65

Vol. 35, Number 9; Japan Robotics Society

Nov. 15, 2017

Development of Fuel Debris Shape Measuring Device

66

The 26th Polymer Materials Forum, The Society of Polymer
Science, Japan

Nov. 16, 2017

Current State of Technical Development by IRID toward theDecommissioning of Fukushima Daiichi NPS

Nov. 23, 2017

Numerical Analysis for Fulushima-Daiichi Unit 22017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 201739

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Analysis/ Evaluation on Indentification of Conditions inside
Reactor Fukushima Daiichi NPS, TEPCO(91) Evaluation on Debris Spreading State by Actual Machine
Sensitivity Analysis for In-Core State Assumption

67

The 10th International Symposium on Nuclear Science andTechnology

402017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Analysis on Inside Temperature Distribution of Fukushima Daiichi
Unit 3 by Using CFD Tool

68412017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Development of Compact Dosimeter toward the Inside Survey of
PCV

Preliminary Evaluation of Radiolytic Hydrogen Amount in Fuel
[Conference on the Erosion Prediction and Mitigation of Major
Debris Canister
Components in Fukushima Daiichi NPS], the 4th Fukushima Nov. 27 – 28, 2017 (Provisional Calculation of Hydrogen Generation Amount by the
Research Conference FY2017
Radioactive Degradation in Fuel Debris Containment Can), etc.
4 cases

69

Discussion Session on X-Ray Material Strength, The Society ofMaterial Science, Japan

Dec. 1, 2017

Nondestructive Inspection of Imaging by Muon Scattering Method

70

QST Takasaki Science Festival 2017

Dec. 12, 2017

Evaluation on Hydrogen Gas Generation Associated withGamma-Ray Irradiation Targeted at the Cement Solidification
Sample of Contaminated Water Treatment Secondary Waste

71

The Society of Instrument and Control Engineers
Lecture Presentation, Systems Integration Category

Dec. 20, 2017

Development of Operation Training System Related to the
Fukushima Daiichi NPS by Using VR

72

Lecture Presentation on the Measures for Decommissioning and
Polluting Water

Dec. 20, 2017

Survey Robot Technique Development at IRID

73

Decommissioning Technology and Human Resource Cultivation
Forum, Tokyo Institute of Technology; 2017

Jan. 10, 2018

Review on Realistic Criticality Risk Evaluation MethodJan. 24, 2018

Current R&D State of International Research Institute for Nuclear
Decommissioning (IRID) toward the Decommissioning ofFukushima Daiichi NPS

42

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Development of Water Stop Technique in S/C by Underwater
Inseparable Concrete Filling(11) Overview of Filling Property/ Water Stop Property Confirmation
Tests by Inseparable Concrete Using Full Scale S/C Model43

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13, 2017

Evaluation on Long-Term Integrity of Fuel Assemblies Retrieved
from Spent Fuel Pool
(4) Evaluation on Fuel Soundness at Wet Type Storage

44

2017 Fall Meeting, Atomic Energy Society of Japan45

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 13 – 14, 2017 Development of Inside Survey Equipment for PCV [PMORPH]
Sep. 14, 2017

Development of Multi Nuclides Analysis Method by ICP-QQQ-MS

74

Journal of Japan Electrical Manufacturers’ Association [DENKI]

75

Research Chronology 2016 (brochure), Takasaki Advanced
Radiation Research InstituteFeb. 2018

Investigation of Hydrogen Gas Generation by Radiolysis for
Cement-Solidified Products of Used Adsorbents for Water
Decontamination

76

Journal of Nuclear Engineering and Radiation Science

Mar. 2018

Mechanical properties of cubic (U, Zr) O2

46

2017 Fall Meeting, Atomic Energy Society of Japan

Sep. 14, 2017

Review on Preprocessing for Dry Type Storage of Retrieved Fuel
Debris– Evaluation on Dry Property Focusing on the Shape of Fuel Debris –472017 Fall Meeting, Atomic Energy Society of Japan

Sep. 15, 2017

Current State of Technical Development by IRID toward theDecommissioning of Fukushima Daiichi NPS

48

The 3rd Strategical Workshop Related to the Decommissioning ofFukushima Daiichi NPS

Sep. 16, 2017

Recommendation to the Decommissioning from Perceptions ofLarge MCCI Test at CEA

49

6th Asia-Pacific Symposium on Radiochemistry (APSORC17)

Sep. 17, 2017

Application of ICP-MS to analytical methods for samples from 1F
site at Okuma Analysis and Research Center

50

HOTLAB2017 (Annual Meeting on Hot Laboratories and Remote
International collaborations at JAEA/CLADS toward decommissioning
Sep. 17 – 19, 2017
Handling)of Fukushima Daiichi NPS, etc. 4 cases

Annual Research Report 2017

Others DataSheet

38

Development of Inventory Evaluation Method for Fukushima
Daiichi Accident Waste(11) Review on Analytical Estimation Technique for Contamination
State in Nuclear Reactor Building

Major Research Achievements In-House Research

34

36

34

2017 Fall Meeting, Atomic Energy Society of JapanNo.

Major Research Achievements Subsidized Projects

No.

35

Others–DataSheet
No.

No.

Project nameCategory

Subject

21

Development of Technology for Investigation inside PCV

Hatch Opening Device Associated Machine: 1 set

22

Development of Technology for Investigation inside PCV

B2 Survey Equipment

23

Development of Technology for Investigation inside PCV

Scattering Prevention Equipment for B2 Survey Equipment

Period24

Development of Technology for Investigation inside PCV

Ancillary Facility for B2 Survey Equipment

Apr. 1, 2017 –Feb. 28, 201825

Development of Technology for Investigation inside PCV

Simulated Object for B2 Survey Equipment Mockup Test

26

Development of Technology for Investigation inside PCVUnderwater Swimming Type Equipment Trial Model : 1 set

27

Development of Technology for Investigation inside PCV

X-6 Penetration Hole Remote Punch: 1 set

28

Fuel Debris Characterization

Large Capacity Thermogravimetric Scales and Simultaneous Thermal Analyzer

29

Fuel Debris Characterization

4-Component Piezoelectric Cutting Dynamometer

30

Fuel Debris Characterization

Elemental Analyzing System for SEM

31

Fuel Debris Characterization

Hydraulic Type Automatic Embedded Equipment

32

Fuel Debris Characterization

Inverted Metal Microscope

33

Fuel Debris Characterization

Carbon Coater

Partner2

Upgrading the Comprehensive
Contract
Identification of Conditions inside
Research
Reactor

Piping Deformation Analysis at Reactor Core
University of Tokyo
Substance Slumping

Upgrading the Comprehensive
Contract
Identification of Conditions inside
Research
ReactorInverse Problem Evaluation by Virtual Nuclear
University of Tokyo
ReactorUpgrading the Comprehensive
Contract
Identification of Conditions inside
Research
ReactorStudy on Cesium Supplementary Reaction
Osaka University
and Diffusion Behavior

Apr. 1, 2017 –Feb. 28, 2018

5

Upgrading the Comprehensive
Contract
Identification of Conditions inside
Research
ReactorStudy on the Thermodynamic Amount Evaluation
Tokyo Institute of Technologyof Cs-Si-(Fe)-O System Chemical CompoundMay 1, 2017 –Feb. 28, 201834

Fuel Debris Characterization

Vacuum Replacement Arc Melting Furnace

35

Fuel Debris Characterization

Fuel Debris Compression Test Equipment

6

Upgrading the Comprehensive
Contract
Identification of Conditions inside
Research
ReactorResearch on Silicon Behavior in Material

May 1, 2017 –Feb. 28, 201836

Fuel Debris Characterization

Fuel Debris Sonic Speed Measurement Device

37

Fuel Debris Characterization

Metallographic Image Analyzer

7

Upgrading of Fundamental
Contract
Technology for Retrieval of Fuel
Research
Debris and Internal Structures

Evaluation on Trajectory Generation Considering
the Avoidance of Interference with the Environment Kobe Universityof Robot Multiple Degrees of Freedom

Jun. 15, 2017 –Feb. 15, 201938

Fuel Debris Characterization

Dynamic Micro Hardness Tester

39

Fuel Debris Characterization

Simultaneous Thermal Analyzing System

Survey on the Estimation/ Control of Hydraulic
Osaka University
Driven Manipulator Finger Load Force

Aug. 1, 2017 –Feb. 15, 2019

Fuel Debris Characterization

Gas Piping Valve Heater

8

Upgrading of Fundamental
Contract
Technology for Retrieval of Fuel
Research
Debris and Internal Structures

40
41

Fuel Debris Characterization

Sample Cutting Machine

R&D for Treatment and Disposal Contractof Solid Radioactive Waste
Research

Review on the Accuracy Improvement of Central Research Institute of Jul. 1, 2017 –Analytical Evaluation Method
Electric Power Industry
Jan. 31, 201842

Fuel Debris Characterization

Sample Polishing Machine

943

Fuel Debris Characterization

Core Sampling Collection Unit

Fuel Debris Characterization

Laser Diffraction Type Grain Size Distribution Measuring Apparatus

4

MELCOR Analysis Related to Event Transition
WASEDA University
at Reactor Core Substance Slumping

Tohoku University

Apr. 1, 2017 –Feb. 28, 2018
Apr. 1, 2017 –Feb. 28, 2018

R&D for Treatment and Disposal Contractof Solid Radioactive Waste
Research

C o - O p e r a t i o n i n Wa s t e M a n a g e m e n t National Nuclear Laboratory Sep. 27, 2017 –Technology Fields on Gas Generation
Limited (NNL) in UK
Feb. 28, 201844
45

Fuel Debris Characterization

Dry Automatic Density Meter

11

Development of Technology for Contract
Criticality Control Methods
Research

Verification Test of Criticality Approach Detection
System to be Applied to the System Including Fuel Kyoto University
Debris and Neutron Absorption Material (Part-2)

Oct. 20, 2017 –Jan. 31, 201846

Fuel Debris Characterization

Heating Furnace for Thermal Analyzer

47

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Full Scale Test Device

12

Development of Repair Technology Contract
for Leakage Points inside PCV
Research

Liquidity Improvement of Self-Compacting
Kochi University of Technology
Concrete

Oct. 1, 2017 –Mar. 20, 201848

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Full Scale Test Facility

49

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Equipment for 1/4 Scale Test

50

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Facility for 1/4 Scale Test

51

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals 1/4 Model Test Facility

52

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals 1/4 Scale Reactive Force Retaining Mechanism Combination Test Sample

53

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Test Sample of PRV Upper Part Water Shielding

54

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Flexible Structure Arm: 1 set

55

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Flexible Structure Arm Control Device

56

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Facility for Mockup in PCV

57

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Facility for Equipment Hatch Carry-In Test

10

Major Research Facility/Equipment
No.

Project name

Over 1 Million Yen

Details

Full-scale test of repair and water stoppage technology for leakage points inside the Primary Containment Vessel Heating/ Water Supply Facility

58

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals 1/1 Scale Hydraulic Type Reactive Force Retaining Mechanism

Full-scale test of repair and water stoppage technology for leakage points inside the Primary Containment Vessel Muddy Water Treatment Facility

59

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals 1/1 Scale Electrically-Assisted Type Reactive Force Retaining Mechanism

3

Full-scale test of repair and water stoppage technology for leakage points inside the Primary Containment Vessel Working Floor

60

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Laser Gouging Power Measurement Unit

4

Full-scale test of repair and water stoppage technology for leakage points inside the Primary Containment Vessel Test Sample Mitigation Rail61

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Laser Gouging Head5

Full-scale test of repair and water stoppage technology for leakage points inside the Primary Containment Vessel Full scale Mock-up Facility

62

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Access Equipment Component Test Equipment

6

Development of Technology for Investigation inside PCV

B1 Survey Equipment

63

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Robot Arm

7

Development of Technology for Investigation inside PCV

Scattering Prevention Facility for B1 Survey Equipment

64

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals Access Rail8

Development of Technology for Investigation inside PCV

Ancillary Facility for B1 Survey Equipment

65

Development of Fundamental Technologies for Retrieval of Fuel Debris and Reactor Internals PCV Welding Device

9

Development of Technology for Investigation inside PCV

Simulated Object for B1 Survey Equipment Mockup Test

66

R&D for Treatment and Disposal of Solid Radioactive Waste

Chamber for Alpha Nuclide Analysis

Shielding Block Removing Device

67

R&D for Treatment and Disposal of Solid Radioactive Waste

Digital Spectrometer
Efficiency Calculation Program for Gamma-Ray Measurement

10

Development of Technology for Investigation inside PCV

11

Development of Technology for Investigation inside PCV

Fuel Debris Shape Measuring Device

68

R&D for Treatment and Disposal of Solid Radioactive Waste

12

Development of Technology for Investigation inside PCV

Element Test Device for Fuel Debris Shape Measuring Device

69

R&D for Treatment and Disposal of Solid Radioactive Waste

Aerosol Mitigation Observation Instrument

13

Development of Technology for Investigation inside PCV

A2 Survey Equipment (Chamber and Guide Pipe included)

70

R&D for Treatment and Disposal of Solid Radioactive Waste

Well-Type Ge Detector

14

Development of Technology for Investigation inside PCV

X-6 Penetration Hole Punch: 1 set

71

R&D for Treatment and Disposal of Solid Radioactive Waste

Core Data Collection Unit

In-Penetration Advanced Check Device: 1 set

72

R&D for Treatment and Disposal of Solid Radioactive Waste

Zeolite Sample Collection Test Equipment

73

R&D for Treatment and Disposal of Solid Radioactive Waste

Compact Core Sample Collection Test Model

15

Development of Technology for Investigation inside PCV

16

Development of Technology for Investigation inside PCV

Sediment Removing Device: 1 set (Chamber included)

17

Development of Technology for Investigation inside PCV

Ancillary Facility for A2 Survey Equipment: 1 set

18

Development of Technology for Investigation inside PCV

In-Pedestal preliminary Confirmation Device: 1 set

19

Development of Technology for Investigation inside PCV

In-PCV Structure Simulated Mockup Body: 1 set

20

Development of Technology for Investigation inside PCV

A3 Survey Component Test Equipment: 1 set

Annual Research Report 2017

Others DataSheet

1
2

Major Research Achievements In-House Research

1

Upgrading the Comprehensive
Contract
Identification of Conditions inside
Research
Reactor3

36

Details

Major Research Achievements Subsidized Projects

List of Joint Research/Contract Research in FY2017

Project name* R&D completed major research facility/equipment have been deleted from the list, FY2017 version.

37