IRID Annual Research Report2015

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

Printed on recycled paper.

Published: July 2016

IRID Research and Development Project, FY 2015 (Overview)

In order to promote R&D technology, we committed to gathering expertise from around theworld under an integrated management which encourages mutual coordination among
projects, overlooking the whole R&D projects, and conducted fifteen subsidized projects
and one in-house R&D project in FY 2015.

March 2016

Repair and Water
Leakage Stoppage
Technology for
the Inside of PCVDevelopment of
Repair and Water
Leakage Stoppage
Technology for the
Leakage Points
Inside PCV
Full-scale Test for
Repair and Water
Leakage Stoppage
Technology for
the Leakage Points
Inside PCVContents

Reliability
Evaluation of
Remotely Operated
Decontamination Inside
Reactor Buildings

Development ofTechnology for
Integrity Evaluationof RPV / PCV

Development ofTechnology for
Criticality Control in
Fuel Debris Retrieval

Technology for Investigation /Analysis Inside the Reactor
Development of
Investigation
Inside the PCVDevelopment ofTechnology for
Investigation
Inside RPVDevelopment ofTechnology for
Detection of
Fuel Debris
in the Reactor

Upgrading for
Identifying Conditions
Inside the Reactor
through Accident
Progression Analysis
and Actual Data

Characterizationof Fuel Debris＜Fuel Debris Retrieval＞
Project of Upgrading
Approach and
System for
Retrieval of Fuel
Debris and Reactor
Internal Structures

Project ofDevelopment ofFundamental Technologies
for Retrieval of
Fuel Debris andInternal Structures

Technology for
Treatment andDisposal of Waste

Development ofTechnology for Collection,
Transfer and Storageof Fuel Debris

R&D forTreatment andDisposal of Solid
Radioactive Waste

Main R&D Achievements – Subsidized Projects◆Key Challenge 1 – R&D for Fuel Removal from Spent Fuel Pool

Evaluation of Long-term Structural Integrity of the Fuel Assemblies Removed from Spent Fuel Pool......2◆Key Challenge 2 – R&D for Preparation of Fuel Debris Removal

Development of Technology for Remotely Operated Decontamination Inside Reactor Buildings....4
Development of Repair and Water Leakage Stoppage Technology for Leakage Points Inside the Primary Containment Vessel.....6
Full-scale Test for Repair and Water Leakage Stoppage Technology for Leakage Points Inside the Primary Containment Vessel...8
Development of Technology for Investigation Inside the Primary Containment Vessel............10
Development of Technology for Investigation Inside the Reactor Pressure Vessel.....................12
Upgrading for Identifying Conditions Inside the Reactor through Accident Progression Analysis and Actual Data....14
Development of Technology for Detection of Fuel Debris in the Reactor..................................16
Development of Fundamental Technologies for Retrieval of Fuel Debris and Internal Structures...18
Upgrading Approach and System for Retrieval of Fuel Debris and Internal Structures...........20
Development of Technology for Collection, Transfer and Storage of Fuel Debris.....................22
Development of Technology for Integrity Evaluation of the Reactor Pressure Vessel / Primary Containment Vessel....24
Development of Technology for Criticality Control in Fuel Debris Retrieval................................26
Development of Technology for Fuel Debris Characterization..................................................28◆Key Challenge 3 – R&D for Treatment and Disposal of Radioactive Waste

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

Hirofumi Kenda
President, International Research Institute for Nuclear Decommissioning

Main R&D Achievements – In-house R&D Activities

Reliability Evaluation of Remotely Operated Decontamination Inside Reactor Buildings.........32

Others – Data Sheet

Main Research Results Announced / Published in FY 2015......................................................33
Joint Research / Contract Research.........................................................................................35
Main R&D Installations / Equipment...........................................................................................36
IRID Annual Research Report 2015

1

Data SheetThis annual report is intended to introduce achievements of R&D which IRID had taken
upon as a challenge since FY 2015. We hope this report will help you understand the resultsof our R&D. IRID will continue to take responsibility for the R&D on the steady andefficient decommissioning toward fuel debris retrieval which is facing a crucial stage. We
are deeply grateful for the continuous support and guidance, and would like to express
sincere gratitude and appreciation to all.

Development ofTechnology for
Remotely Operated
Decontamination Inside
Reactor Buildings＜Ensuring of Stability＞

OthersAs a result, robotic technologies were developed. Remotely operated robots for thedecontamination of the reactor building were tested on-site, and the feasibility of thetechnologies verified. Robots for the investigation inside the Primary Containment Vessel(PCV) brought us effective information inside the PCV. Furthermore, Muon tomographyusing permeation method was developed and revealed that fuel debris hardly exists in thecentral region of the Reactor Pressure Vessel (RPV) at Unit 1. We have achieved successin the development of technologies that is essential for fuel debris retrieval.

＜Ensuring of Work Environment＞

Technology for Fuel Debris Retrieval

Main R&D Achievements In-house R&D Activities

The International Research Institute for Nuclear Decommissioning (IRID) has beenaddressing various themes in the following three R&D categories as the most urgent issues
toward the decommissioning of Fukushima Daiichi NPS since its start in 2013.
(1) R&D for fuel removal from spent fuel pool
(2) R&D for preparation of fuel debris retrieval(3) R&D for treatment and disposal of radioactive waste

Technologies for Decontamination/ Radiation Dose Reduction

Subsidized Projects

It has been almost five years since the accident at the Fukushima Daiichi Nuclear Power
Station (NPS) was caused by the Great East Japan Earthquake. The situation at the NPS hasmuch improved after the accident; however, we are facing a crucial stage fordecommissioning. In the Mid-and-Long Term Roadmap revised last June, the latest
milestones are clarified, and consistent achievements are required for R&D technology
toward decommissioning more than ever.

Evaluation of Long-term Structural Integrity of the Fuel Assemblies Removed from Spent Fuel Pool

Main R&D Achievements

Message from the President

R&D for Fuel Removal from Spent Fuel Pool

Main R&D Achievements
Key Challenge 1 – R&D for Fuel Removal from Spent Fuel Pool

Fuel assemblies in the spent fuel pools at Units 1-4 at the
Fukushima Nuclear Power Station (NPS) have been stored ina water quality environment different from normal conditions
due to the injection of seawater and rubble falling into thepool. To achieve long-term storage of fuel assemblies in acommon pool in future, it will be necessary to accurately
evaluate the effects of these conditions on fuel assemblycomponent materials and the integrity of the fuel assemblies
for long-term storage.

In order to evaluate whether fuel assemblies retrieved
from the spent fuel pools at Units 1-4 at the Fukushima
NPS can be stored safely or not, it will be necessary to
perform corrosion tests and investigate actual fuel in anenvironment that simulates real storage conditions. The
impact on the fuel integrity during dry storage will be
evaluated by a simulated test.

1

Technical development for the evaluation of the long-term integrityof the fuel assemblies

a- Evaluation on the deposits on the surface of the fuel assemblies
In order to analyze white deposits which had been observed on the surface of the spent fuel stored in the common pool at
Units 1-4 at the Fukushima Daiichi NPS, preparations were made for transport of fuel components and a post irradiation test in
the test facility.
For a smooth transportation of fuel components, loading procedures for transport casks and collecting lock nut were
examined. A mockup for loading lock nut into transport casks was conducted.
In the preparation for testing in the post irradiation test facility, a test guideline for analysis of white deposits was created aswell as electrochemical test using lock nut was studied. It is confirmed that tests are enabled to be conducted in the post
irradiation test facility.

1mm
50μm

50μm

Enlarged scratched part

irradiation 9x9, comparison
irradiation 9x9, Scratches + seawater

The direction of hydride precipitation behavior in a scratched area
is an almost circumferential direction as well as sound parts. Thereis no effect on hydride precipitation behavior with scratches.

irradiation 5x8, Scratches + seawater

0.20

0.10

0.00

0

500

There is no clear difference due to existence (of scratches + seawater)by the 1,000-hr. long test results. Behavior continues to be verified.

Future Developments
Efficient data for the study of corrosion occurrence during storage in the common pool will be acquired through analysis andidentification for white deposits on the surface of fuel assemblies.
Furthermore, we will continue to study the effect on the integrity of fuel assemblies in dry storage caused by specific factors
at Units 1-4 at the Fukushima Daiichi NPS.

Test solution B

CI-36 distribution

Chloride ion concentration (ppm)

Measuring sample for
distribution of radioactivity

Test solution C
Test solution D

10,0009,420 ppm

9,000
8,000
7,000

Chloride ion
concentration increvice parts

Chloride ion
concentrationof test solution

6,000
5,000
4,000

3,101 ppm

3,000
2,000

1,016 ppm

1,000
0

12 ppm

0

25

50

75

100

125

Immersion hrs. (h)Fig. 3: Distribution of CI-36 in upper end plug crevice structure(80℃, 2x dilution of artificial seawater after 50 hrs. of immersion)

Fig. 4: Evaluation result of chloride ion concentration in the upper
end plug crevice structure

According to the results of Fig. 3 and 4, it is confirmed that seawater components transfer to the crevice structure of upper end plug by immersing
in a 2x dilution of saltwater; however, concentration decreased afterwards as water quality improves.

2

IRID Annual Research Report 20153

Data Sheet

In order to evaluate quantitatively the seawater components transfer behavior in a crevice structure of fuel components(upper end plug), tests using tracer were conducted.
The test result proved that concentration of chloride ion decreased in accordance with fluid concentration by improving thequality of the solution and was not enriched in the crevice structure, although chloride irons were taken into the upper end plugcrevice structure by seawater injection (Fig. 3 and 4).

1,500

Fig. 2: Example of a study result (attached scratches + seawater)that can affect the creep speed

Upper end plug

Basic tests for long-term structural integrity

1,000

Others2

Test hrs. (h)Test solution A

Nut

Main test (Long-term integrity at 1F, 2015)0.30

b- Evaluation of Integrity of fuel in dry storage conditions
When fuel assemblies retrieved from the spent fuel pool at Units 1-4 at the Fukushima Daiichi NPS are stored in dry condition,
it is necessary to evaluate the impact on the integrity of fuel assemblies caused by scratches on fuel cladding due to fallingrubbles or seawater due to seawater injection.
Therefore, a low-ductility test (Hydride deposition behavior evaluation test) and a high strain test (creep test) were conducted.
Test conditions in 2015 confirmed that there was a small impact on integrity of fuel assemblies caused by scratches on fuelcladding due to rubbles or seawater by seawater injection (Fig. 1 and 2). We will continue to study the impact on the integrityof fuel assemblies caused by specific factors under dry storage conditions at Units 1-4 at the Fukushima Daiichi NPS.

}JNESH15、Irradiation of high burn-up 8 x 8, No.64

Enlarged sound partFig. 1: Example of a verification test result for hydride precipitation
behavior(Irradiated 9x9 fuel cladding, 300℃, cooling velocity 0.3℃/h,
stress in a circumferential direction 70MPa, with scratches)

irradiation 8x8, comparison

0.40

Main R&D Achievements In-house R&D Activities

Main Achievements and Approaches

Sound part

Strain of circumferential direction (%)

Aims

360℃, 170Mpa, Irradiated fuel cladding

0.50

Subsidized Projects

Background

Scratchprocessing part

Main R&D Achievements

Evaluation of Long-term Structural Integrity of theFuel Assemblies Removed from Spent Fuel Pool

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Aims

In order to retrieve fuel debris from the Reactor Pressure
Vessel (RPV) and the Primary Containment Vessel (PCV) at
Units 1-3 at the Fukushima Daiichi NPS, various tasks areplanned inside the reactor buildings. For smooth execution, it isnecessary to improve the working environment and reduceoverall radiation levels in combination with decontamination,
shielding and removal of radioactive sources.

Based on R&D achievements up to FY 2014, decontaminationequipment for high places will be improved and verified.
Decontamination equipment of upper floors will be manufactured
and verified, and conceptual study will be conducted fordecontamination of basement floors.
According to the above steps, remotely controlled decontamination
techniques, one of the important technologies required for thesmooth implementation of surveys, repairs and water leak
blockage inside the reactor buildings will be established.

Main Achievements and Approaches

1

Hose reel

ElevatingworkmachineCart (in travelling)

Cart (in travelling)

Charger

For suction blast
Decontamination unit

On-site panel(Insulated transformer,
distribution board
and media converter)

Step-downtransformer＜Relay cart＞＜Support cart＞＜Transport cart＞＜Work cart＞

Suction / Blast
decontamination head(a) Suction/Blast decontamination equipment
Hose feeding device for a relay cart
Hose feeding
Charge reelmechanism

ChargerCable winder

External DG,
Existing HUB

Hosefeeding devise

Hose feedingmechanism

On-site panel(Insulated transformer,
distribution board andmedia converter)＜Relay cart＞

Battery unit
Hose
(4hr. rated
feeding devise
attery unit
output)(4-hr. rated output)
Hose feedingmechanism＜Support cart＞＜Transport cart＞＜Work cart＞
High-pressure water jet
cleaning head

High-pressure
water jet pump,collecting tank,distribution board,
HUB(b) High-pressure water jet decontamination equipment
Cable hose reelFig. 2: Conceptual diagram of upper floor decontamination equipment

OthersData Sheet

Development of decontamination equipment for upper floors

Cable winder

External DG, Existing HUB

b- Verification test for decontamination equipment

2

Compressor(Used for both dry ice blastand suction blast)

Cable winder

Cable winder

a- Improvements of decontamination equipment

In a mock-up test facility in which upper parts of the first floor in the reactor building were simulated, verification re-tests wereconducted to mainly improve features of high-pressure water jet decontamination equipment, dry ice blast decontaminationequipment and suction/blast decontamination equipment such as decontamination performance, remote travelling, operability andsafety function. The results showed that it would be applicable in actual use.

(c) Suction/Blast
decontamination equipment

Fig. 1: Decontamination equipment for high places

Development of decontamination equipment for high places

Based on the results of verification tests in FY 2014, various improvements were made. High-pressure water jet
decontamination equipment was improved for visibility and easy-maintenance; dry ice blast decontamination equipment wasadvanced for operability during travelling or decontamination work and suction/blast decontamination equipment was improved forvisibility and collecting operability in case of emergency (Fig. 1).

(b) Dry ice blast
decontamination equipment(a) High-pressure water jet
decontamination equipment

Main R&D Achievements In-house R&D Activities

Remotely operated decontamination equipment is mainly used in lower parts of the first floor, upper parts of the first floor, upper
floors and basement areas. By FY 2016, development of decontamination equipment for lower parts of the first floor has
completed. Decontamination equipment for high places was manufactured and verified by tests and that for upper floors was
designed and manufactured. Based on the aforementioned results, the following tasks were executed in FY 2015.

Decontamination head

Subsidized Projects

BackgroundArm
movement

Arm/Head

Main R&D Achievements

Development of Technology for Remotely Operated
Decontamination Inside Reactor BuildingsArm(2 sets)

Work cart

a- Manufacturing of decontamination equipment
Following FY 2014, decontamination equipment for upper floors was manufactured. Decontamination equipment is designed toaccess upper floors of the reactor buildings (the second and the third floors) by using an elevating work cart for decontaminationof the floor and wall surface (up to approx. 2 meters). Equipment is comprised of a work cart, a transport cart, a support cart anda relay cart. Mounted in each cart are the decontamination units: high-pressure water jet decontamination unit, suction/blast
decontamination unit and dry ice blast decontamination unit (Fig. 2, 3, 4).

b- Verification test for decontamination equipment
A mock-up test facility that simulates the second floor of the reactor building and elevating test facilities were created, andverification tests were conducted. Various safety tests conducted for decontamination performance of each decontamination
system such as remote travelling performance, operability, accessibility to upper floors, and safety function on collection, etc.,
during emergency. The results showed that it would be applicable in actual use.

3

Conceptual study on decontamination equipment for basement areas

A conceptual study was performed on stagnant groundwater that had been considered as a radiation source on the basement
floors of the turbine building and sludge accumulated in the bottom as well as treatment for highly contaminated water
accumulated in the main condenser (primary stagnant water). The study showed that existing technologies would be applicable.
In order to confirm the feasibility of conceptual study on the basement of the reactor building, the necessary information on thecontamination contributing to radiation dose was organized.

4

IRID Annual Research Report 2015

Relay cart
Support cart
Transport cart

Fig. 3: Upper floor decontamination equipment (Showing
mounted suction/blast equipment)

Fig. 4: Upper floor decontaminationequipment (Elevating)Future Developments
Development of the decontamination equipment for high places and upper floors has been completed in FY 2015. Dry ice blast
equipment for high places was applied in actual use in the second semester of FY 2015. On the other hand, other decontaminationequipment will be utilized according to on-site requirements and timing for start of construction.
In regards to the basement floor of the turbine building, decontamination will be conducted accordingly as water level decreases.
The results from the conceptual study and on-site investigation will be used as reference. These results will be utilized since
on-site investigation requires clarifying needs for the basement of the reactor building.
5

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

SRV pipe
Inner catwalk

Vent pipe

Vacuum break line
Outer catwalk
Vacuum break line bellows

Main R&D Achievements

Development of Repair and Water Leakage Stoppage Technology
for Leakage Points Inside the Primary Containment Vessel←Placing pipe

Aims

It is assumed that fuel debris at the Fukushima Daiichi NPS
has not only fallen down within the Reactor Pressure Vessel(RPV) but has also made its way into the Primary Containment
Vessel (PCV). In order to retrieve fuel debris, the PCV isplanned to be submerged. Therefore, it is necessary to ensurethat water leakage is prevented from the PCV.

Toward the realization of the submersion method for fueldebris retrieval, this project is intended to establish
technology for repair and water leakage stoppage fromleakage points in the PCV.

Vent pipe bellows

Vent header OP4660

r
inne
shell
S/C

eter
diam

Subsidized Projects

Background

Simulated test device of S/CS/C center OP3570(Height to which the reinforcement
material needs to be ﬁlled to support S/C)

Upper pin

Upper pinT-quencher OP970

Ring header
Column support
S/C lower bottom OP485

Columnsupport10m

Reinforcement material placing height in S/C

Fig. 2: Verification test for evenly balanced flow behavior of strengthening S/C
support column (1/1 scale: simulated Unit 2,3)

Main Achievements and Approaches
1Fig. 1: Reinforcement material placing height in S/C(green area)

It was confirmed whether reinforcement materials poured into a placing pipe
flowed evenly toward the opposite side through under S/C.

On the technology for strengthening the suppression chamber (S/C) support columns (Fig. 1), tests in FY 2013 proved that theintensity of a reinforcement material decreased due to a long distance travel after placing. In order to solve this problem, further
test confirmed that the improved material is enabled to maintain sufficient strength. Furthermore, the performance target was
secured for long-distance fluidity, evenly balanced behavior (Fig. 2) and effects on surmounting obstacles.

2±5°

Water stoppage technology involving injection of sealing material into the vent pipe

Vertical：258㎜

a- Water stoppage technology inside of the vent pipe facility

Floor drilling

Water stoppage material(underwater inseparable concrete)

A full-scale deployment test for an inflatable sealing bag (a cloth bag), which will be installed at the tip of vent pipe, wasconducted. Challenges were identified. In addition to dealing with the challenges, an elemental test for sub-inflatable bag that
contains materials to help water stoppage (a clogging prevention material for a gap between a vent pipe and an inflatable sealingbag) will be conducted. Candidate materials will be selected for actual use.

R/B 1st floor
360°

Bending mechanismb- Water stoppage technology involving ﬁlling inside of the suppression chamber (S/C)

Considering the vacuum break line that is not accessible from right above the first floor of the reactor building at Unit 1, a
full-scale test for installation of a flexible guide pipe and a water stoppage plug that can be accessed from oblique direction was
conducted. In the result, workability and feasibility were confirmed through identifying pressure resistance, installation of water
stoppage plug, opening a hole of the vacuum break line and setting a guide pipe (Fig. 4).

3

Water stoppage technology for leakage from the seal section and of vent pipe bellow

Fig. 3: Status of down-comer long-distance fluidity test
Water stoppage materials were placed from both ends of a 24-m watertank. Capacity for repletion/water leak blockage in the interior of thedown-comer around the center of the merging area from both ends was
confirmed.

Expansion joint：200㎜

Data Sheetc- Water stoppage technology involving injection of sealing material into the vacuum break line

Bending mechanism：40°±3°（Position 1 m away fromthe vacuum break line）

Fixing mechanism

OthersMaterial compounding of water stoppage materials were determined. Downcomer water stoppage test, quencher water
stoppage test and strainer water stoppage test were conducted. Additionally, basic performance for water stoppage by using thematerials was confirmed through tests: a long distance fluidity test at the merging point of the two-way flows and a vacuum breakvalve water stoppage test (Fig. 3).

Main R&D Achievements In-house R&D Activities

Technology for strengthening the suppression chamber (S/C) support columns

Fig. 4: Flexible guide pipe overview
A guide pipe was developed to be accessed from oblique position as it is notable to drill holes right above the vacuum break line due to an obstruction at
the drilling position of the 1st floor.

Surveillance camera
for the construction area

Welding torch

Surveillance camera
for the wholeCart

As equipment hatch is not accessible due to high radiation dose, conceptual study of remotely operated welding method hascommenced (Fig. 5).

4

Boundary construction technology for PCV connector pipes

In order to isolate connector pipes from the PCV, the study of water stoppage proposals is ongoing. Atmospheric Control System(AC System), Radioactive Waste System (RW System) and Reactor Component Cooling Water System (RCW) were selected.
Water stoppage tests were conducted after clarifying the required performance of water stoppage materials.

Part A

7-axis articulated
robot arm

Future Developments
In order to establish water stoppage technology for each part to be repaired (water stoppage), various tests continue to be
conducted for practical use, and challenges will be reviewed, analyzed and evaluated. In addition, considering the applicability inactual use, the required performance will be reflected in the equipment and the study of long-term water stoppage maintenance willproceed further.
6

IRID Annual Research Report 2015

Driving mechanism
for front and rearFig. 5: Water leakage stoppage technology for the sealing parts (water
stoppage for the equipment hatch)

Welded part of
the flange surface

Water leakage stoppage method for equipment hatch has
been studied through methods involving injection of
cement-based materials; however, it proved difficult tocompletely stop water from leaking. Therefore, a remotewelding water stoppage method is now being considered.

7

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Aims

Toward the decommissioning of the Fukushima Daiichi
NPS, it is essential that repair and water leakage stoppage
technology for leakage points inside the PrimaryContainment Vessel (PCV) is established for fuel debris
retrieval by full submersion (Submersion method).

This project is intended to perform a full-scale test for theverification of developed technologies and operational
training toward the application of repair and water leakage
stoppage technology on-site (construction methods,
remotely controlled equipment, etc.).
Testing is planned at the Naraha Remote Technology
Development Center as a joint proposal with the Japan
Atomic Energy Agency (JAEA).

Subsidized Projects

Background

Main R&D Achievements

Full-scale Test for Repair and Water Leakage Stoppage Technology
for Leakage Points Inside the Primary Containment Vessel

Main Achievements and Approaches

Full-scale mock-up test of equipment to repair and stop leakage
in the lower part of PCV

In order to confirm the operability of reinforcement equipment for suppression chamber (S/C) support reinforcement developedin the project entitled “Development of Repair and Water Leakage Stoppage Technology for Leakage Points Inside the PCV,” afull-scale mock-up test using water has commenced (Photo 1).
In addition, necessary tests and confirmation items were identified for the study of actual construction and were thoroughly
examined for further full-scale test.

2

Photo 1: Full-scale testing
Upper left: Placement device, Upper right: Surveillance camera
image of the placement device tip, Lower left: Lowering nozzle ofthe placement device

Photo 2: Assemblies of full-scale test device
Lower left: Assemblies for S/C shell, Upper right: Assemblies for S/Cshell and down-comer, Lower left: Assemblies for vent pipes, Lower right:
Simulated torus room wall (panoramic view)

Design and production of full-scale mock-up

Continuing from FY 2014, a full-scale mock-up facility simulating the lower part of the PCV at the Fukushima Daiichi NPS Unit
2 was designed, manufactured and installed. This year saw the completion of installation for a full-scale mockup facility in the
Naraha Remote Technology Development Center (Photo 2).

ResearchandAdministrative
Bldg.

OthersThe main specifications of the full-scale mockup are as follows:●A full-scale model is a one-eighth sector simulating the lower parts of the PCV (the suppression chamber [S/C], the vent pipes[excluding the bellows], vent head, downcomer and surface of torus room wall) at Unit 2 at the Fukushima Daiichi NPS.

Turbid water
treatment facility

Study/design/production/installation of required equipment
including plumbing and turbid water treatment systems

Data Sheet

3

Main R&D Achievements In-house R&D Activities

1

Test Bldg.

Work floor

In a full-scale testing, we simulated actual conditions at Units 1-3 at the Fukushima Daiichi NPS as much as possible such as
size and temperature. Continuing from FY 2014, the following equipment was studied, designed, manufactured and installed.
Installation of necessary equipment has completed as scheduled in the Naraha Remote Technology Development Center.
[Heating/Feed water equipment]
Heated water, which simulated actual stagnant water at the Fukushima Daiichi NPS, is planned to be used for a full-scale
mock-up test. Installed equipment is capable to heat water and supply heated water at the required volume.
[Turbid water treatment equipment]
Water stoppage materials (grouting material [including cement] ) is used for stopping water at the lower part of the PCV, therefore, watercontaining cement generates after the test. Turbid water treatment equipment was installed to drain water outside the facility.

Heating feedwater equipment[Work floor]
The height of the first floor of the basement at Unit 1 differs from the height at Units 2 and 3 at the Fukushima Daiichi NPS. As
the water stoppage equipment is expected to be installed from the first floor in actual operation, work floors were installed tochange the floor level for each unit when a full-scale test is conducted.

Fig.1: Preparation of a test device facility[Transfer rail for test device]
Since the weight of the mock-up is very heavy (approx. 5,400t), it is difficult to move once it is assembled. Therefore, a transfer
rail that enables to move the mock-up was installed.
8

IRID Annual Research Report 2015

Mock-up transfer rail

Full-scale test device

Future Developments
A full-scale mock-up test will be conducted accordingly to utilize the achievements of the development activities for repair and water
stoppage technologies.
Considering the actual construction, a guideline will be created and procedures will be verified. In addition, performance validation will
be performed for repair and water leakage stoppage materials through observing the conditions of filling for the materials. These results
will be the criterion to determine the methods of fuel debris retrieval.
9Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

It is considered that reactor cores at Units 1-3 have melted,
and fuel has partially fallen with reactor internals into theReactor Pressure Vessel (RPV) and Primary Containment
Vessel (PCV). It is assumed that fuel debris has fallen fromthe RPV bottom into the pedestal, which supports the RPV,
then spread out of the opening of the pedestal bottom into
outside of the pedestal and remained in the PCV bottom.
However, actual conditions have not been identified.

With regard to the conditions of the PCV internal,
information such as remotely visual image, radiation dose
and temperature in the vicinity of the penetration wereobtained through access to the PCV internals from X-100B
Penetration at Unit 1 and X-53 at Unit 2. Based on the
results, it is confirmed that the visibility is limited due to thepresence of stream and stagnant water in the dark, in
addition to the harsh, highly radioactive and extremely humid
environment. Moreover, there is the possibility of unforeseen
obstructions generating from the accident. In order toovercome these challenges, this project aims to develop
technology for the investigation inside the PCV.

Grating

X-100B penetration

Main R&D Achievements In-house R&D Activities

Aims

Photo 3: B1 investigation device(outside the pedestal)

Subsidized Projects

BackgroundPhoto 1: A2 investigation device(inside the pedestal)

Main R&D Achievements

Development of Technology for Investigation Inside
the Primary Containment Vessel

Main Achievements and Approaches
1

Formulation of a plan for investigation inside the PCV
and development of investigation equipment

Prior to formation of the plan, requirements were organized and analyzed in cooperation with other related projects. Moreover,
we followed up the discussion on the requirements and studied further investigation plan based on the results of the investigationon the grating zone (B1 investigation).

2

X-6 penetration

Development of investigation devices, etc.

a- Technology for accessing inside the pedestal：Access route
Photo 2: Shielding block removal equipment(in front of X-6 penetration in Unit 2)

Data Sheetb- Technology for removal of the shielding block

Pedestal
OthersEquipment for the investigation on the platform inside the pedestal (A2 investigation) was improved as reflected in test results.
Preparation has been completed for on-site verification at Unit 2 (Photo 1).
Moreover, for further investigation of the inside pedestal (A3 investigation), an element test on research techniques was
conducted in parallel with test production/testing for equipment to free a hatch of X-6 penetration.

Fig. 1: B2 investigation (Image)＊(Basement outside of Unit 1 pedestal)
Camera for travelling (each on both sides)

Remotely operated equipment for removal of the shielding block placed in front of the X-6 penetration at Unit 2 (equipment forremoving the X-6 shielding block) has been developed. An on-site verification test was conducted from June to July 2015. Results
showed that 128 out of 135 shielding blocks and two iron plates in the back of three plates were removed. It was confirmed thatremaining blocks and a plate were strongly adhered. Installation structure was different from what had been expected; therefore,
an end effector that would remove the adhesion was additionally developed (Photo 2).

Winchc- Technology for accessing the the outside of the pedestal
In order to obtain information e.g. visual image, radiation dose and temperature access through investigation device working on
the grating of the first floor inside the PCV at Unit 1 (B1 investigation), an investigation was conducted in April 2015. Through thisinvestigation, it showed that there was no extensive damage on the existing facilities inside the PCV. Data on the radiation dose
and temperature in the 3/4 range of the grating on the 1st floor were obtained (Photo 3).
In addition, conceptual study, elemental and manufacturing tests were conducted. Investigation methods were established forfurther investigation (investigation on the basement floor [B2 investigation]) following B1 investigation (Fig. 1).

Cable specification
・Composite cable including light fiber
Direction to the lensDirection to the light
Dosimeter + underwater camera
Size (target): around Φ20mm

Devices currently being developed and their applications

d- Technology for measurement of fuel debris
Fuel debris measurement equipment, which identifies the position and distribution of melted materials that are suspected fueldebris, under extremely harsh environments – darkness, high level of radiation, raindrops and fog – was designed and prototyped.
Taking into account the severe environment inside the PCV, optical cutting method was applied. Development is ongoing towards
verification tests after FY 2016.

10

IRID Annual Research Report 2015

＊Fig. 1 is a current idea based on the concept examined by IRID. This
image might be changed due to further improvements depending on the
future development and achievements of element tests.

Future Developments
Development continues toward the verification test for B2 investigation at Unit 1 and A2 investigation at Unit 2 which arescheduled to be conducted in FY 2016. In addition, development of equipment also continues for further investigation both into theinterior and exterior of the pedestal inside the PCVs.
11

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Aims

In order to develop equipment for retrieving fuel debris andcore internals, information about the location, shape andconditions must be understood in advance. However, it iscurrently a difficult situation to acquire directly information
about the inside of the Reactor Pressure Vessel (RPV) due toextremely high radiation level, in addition to the complicatedreactor internal structure.

Based on the compiled needs from related projects and theconstruction site, the development plan and investigation planestablished in FY 2014 will be updated. In addition, thefeasibility of the system and technologies will be confirmed
through elemental test of the system that maintains theboundary functions, which is a primary technology for theinvestigation of the RPV internals, access technology for thereactor core and sampling technology for fuel debris.

Guide pipe
Base plate
Operating floor

Steam dryer
Position of
opening holes

Well cover

Guide pipe main device

Main Achievements and Approaches

Opening holes
from the top

Formulation and updating of plan for investigation/development

The information needs that had to be obtained through the investigation of the RPV internals were collected from related projects
on fuel debris retrieval and on-site parties. Moreover, information, particularly high necessity items like radiation dose level insideof the RPV, situation of molten fuel remains and distribution and composition of fuel debris were analyzed, evaluated andextracted. In order to obtain the information, investigation items such as visual checking in the reactor cores and sampling
methods for fuel debris were updated and clarified. In addition, based on the evaluation results for the feasibility of thetechnologies as shown below ( 2 , 3 ), the development plan was updated after FY 2016.

2

Elemental testing and feasibility assessment for technology
of investigation by drilling holes at the top of the reactor

Guide pipe sealingmechanism

PCV

Upper grid plate

Fig. 1: Image of boundary function maintaining device

Fig. 2: Elemental analysis for opening holes using partial simulated
test device of the reactor internals

a- Conceptual study on the system for maintaining boundary function

Removing sample from access device
and taking out from the site

Access/cutting/collection of fuel debris

Transport container
(T/C)

3

Access device

Sampling cell

Cutting/collection device

Investigation of the technology for fuel debris sampling and feasibility assessment

As a result of the fuel debris sampling required from related projects, composition of fuel debris and mechanical properties(hardness) were mainly chosen. In order to determine design conditions of the sampling system, sampling items such asmeasurement, number and position were organized.
In addition, the entire flow was studied from the sampling of fuel debris to its analysis (Fig. 3). Conceptual design (including
elemental test) was performed for cutting and collecting devices as the main component device, access device (entry from theside of the PCV), measuring device for uranium/plutonium, sampling cell, etc. The feasibility of the sampling technology, which isunder development, was verified. In order to analyze fuel debris, necessary evaluation items were clarified when transporting thesampling of fuel debris outside the Fukushima Daiichi NPS.

Data Sheetb- Conceptual study on the technology for accessing the reactor core
In order to build an investigation route to the reactor core, the conceptual study on the technology of drilling holes, remotely
controlled in the complex reactor internals (steam dryer, steam separator, upper grid pate and etc.), was conducted. Taking into
account the difficulty of supporting the reaction force via remote operating, contactless cutting methods such as razor
cutting/plasma cutting/water jet cutting (abrasive cutting) were considered before fundamental test was conducted. Elemental test
was also conducted for drilling holes using a partially simulated test device of the reactor internals. Also, the feasibility of theinvestigation route was verified (Fig. 2).

Refilling in the intermediate cell and taking outside the site

Based on results from the needs investigation, conceptual study and elemental test conducted in FY 2015, development oftechnology for investigation inside the RPV continues. In regards to the technology of investigation for drilling holes at the upper
part of the reactor, trial manufacturing of the investigation system and securing the boundary system and mock-up test also willproceed. Regarding the technology of fuel debris sampling, test production of cutting and collecting devices and performance test
will be conducted. Upgrading of access device and cell will be studied as well.
12

IRID Annual Research Report 2015

Intermediate cell

Transportation
公道輸送
on the public roadAnalysis facility off-site

Transport container

Manipulator

U/Pu measuring device

Transportation boxof test sample

Manipulator

Examination and study for the technologies/
Items required to evaluate debris transport

Scope of study for the research
Transport container
Sample capsule

Future Developments

Others

The conceptual study was conducted for an alternative boundary facility (guide pipe) that will be placed on the head of the PCV(Fig. 1) as a prevention system against the scattering of radioactive material caused by the drilling of holes for the building of aninvestigation route. In addition, an elemental test was conducted for sealing properties between the alternative boundary and theconnecting part of the PCV head. Through the test, the feasibility of boundary function was verified.

Main R&D Achievements In-house R&D Activities

1

Position of
opening holes

Opening holes
from the top

Subsidized Projects

Background

Tool box
Connecting flange

Gate valve

Main R&D Achievements

Development of Technology for Investigation Inside
the Reactor Pressure Vessel

Sample sealed bottle

Transport
Container

Sample sealed bottle

・This overview chart shows an image of theintermediate cell and access device. The
method, boxes, amount, etc., will be confirmed
based on further evaluation.

Fig. 3: Flow chart of the whole process from fuel debris sampling to analysis (Image)13

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

CoreCore support plate

Fuel support metal fitting
Improved fission productdeposit model
Control rod guide tube
Control Rod Drive mechanism(CRD) housing
Pedestal

Particulate debris
Metal layer
Molten pool
Solid crust

Improved lower plenum
debris bed layering model

Improved lower RPV failure model1

Corium leakage through sump pit
and added erosion model

Added thermal conduction to
CRD housing and radiation model
between CRD housing

PCV floor sump pit

Improving thermal conduction
model for the entry phase ofwater addition on top of corium

Added Corium leakage throughsump piping model

In accordance with the upgraded specifications established in FY 2014, improvement was made on a physical phenomenon
model for fission product transition and fuel debris behavior of accident progression analysis code (MAAP and SAMPSON) (Fig.
1). Also, technology for accident progression analysis was enhanced.
●MAAP (Modular Accident Analysis Program): Comprehensive analysis code that enables the consistent evaluation of thermal-hydraulic phenomena/fission
product behavior in the Reactor Pressure Vessel (RPV) and the Primary Containment Vessel (PCV) at the time of severe nuclear power plant accidents.

●SAMPSON (Severe Accident analysis code with Mechanistic, Parallelized Simulations Oriented toward Nuclear fields): A model, based on a multidimensional

Fig. 2: Example of MAAP code accident progression analysis (Unit 2)

Improvements were made on a physical phenomenon model on
the transfer of fuel debris behavior and fission products.
Technology for accident progression analysis was upgraded.

Analysis results were obtained from the improved code andsensitivity analysis which was almost the same as those of theactual measurement in Unit 2 with complicated accident
progression.

The comprehensive analysis and assessment database for the interior of the reactor was developed to collect and organize data
information obtained from other R&D projects, on-site operation and related information of accident progressing analysis. Results
were utilized for comprehensive analysis and assessment.

4

Sump pit
Simulated instrumentation piping
inside the reactor test device
Simulated ICM housing
Heating pot
for molten
debris(Produced
by KAERI)

Thermocouple block(5 blocks)
Thermocouple(15 blocks)

3600 sComprehensive analysis/assessment based on results of actual data and other projects

Comprehensive analysis and assessment (see below) were performed utilizing the data information obtained from the results ofaccident progressing analysis, actual investigation. Existing location, quantity and composition of the fuel debris and fission
product were evaluated. As a result, it is estimated that most of the fuel debris had fallen from the Reactor Pressure Vessel (RPV)
into the Primary Containment Vessel (PCV) at Units 1 and 3. On the other hand, a certain portion of the fuel debris was assumed
to exist in both of the reactors at Unit 2.
Items

Unit 1

Unit 2

Unit 3

Accident progression
analysis

Most of fuel debris has fallen down
to PCV.

Distribution of fuel debris largely depends on theamount of injected seawater by fire engines.

Most of fuel debris has fallen down
to PCV.

Evaluation of heatbalance method

Heat source is few in RPV.

A certain portion of fuel debris exists
both in RPV and PCV.

A certain portion of fuel debris exists
both in RPV and PCV.

Muon Radiography

High density structures (fuel debris)
hardly exist in the reactor core.

A large mass of fuel debris hardly exist
in the reactor core.※

Not measured

Investigation of inside of
PCV

The investigation shows no large-scale
damage to PCV wall( in the past researches.)

No large-scale damage in RPVlower head.

The investigation shows no large-scale damage
in the structures (in PCV in the past researches)

Comprehensive
evaluation

Most of fuel debris has fallen down
to PCV.

A certain portion of fuel debris
exists both in RPV and PCV.

Most of fuel debris has fallen down
to PCV.

100kg of fuel debris equivalent(actual corium), which was melted
by induction heating, was injected
into a pot for 15 sec..

Pedestal wall

Sump pit

3.95 mDevelopment of comprehensive analysis and assessment database

Concrete wall
Concrete floor

Fig. 3: Example of Fuel debris behavior evaluation in PCV (Unit 1)
Analysis code was developed to evaluate Molten Core ConcreteInteraction (MCCI) in detail and estimate the behavior of fueldebris fallen into PCV.

Cover tube/
Simulated cable
Simulated molten
penetration pipe

Data Sheet

3

Molten debris
Solidification debris

OthersAccident progressing analysis and sensitivity analysis were conducted through improved analysis code. Existing position,
quantity and composition of the fuel debris and fission product were evaluated (Fig. 2). In addition, analysis code was developed
to evaluate the Molten Core-Concrete Interaction (MCCI) and fuel debris in detail, and fuel debris behavior that had fallen in thereactor estimated (Fig. 3). Furthermore, tests were conducted to verify the conditions of the damaged in-core instrumentation byusing real-size penetrating pipes and actual corium in Korea Atomic Energy Research Institute (KAERI) (Fig. 4).
Additionally, the three-year Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Phase-2
Project has been executed. Its results have been utilized for evaluating conditions inside the reactor.

Date and time

Fig. 1: Improvement / upgrading for MAAP Analysis Model

14.2 mAssumption and evaluation in identifying inside the reactor utilizing accident progressing analysis

Leakage from RPV gas phase part (Estimation)

Improved debris/
concrete mixture model

numerical formula presenting a physical phenomenon and theory, that is suited to analyze the dispersion and properties of fuel debris.

2

Leakage from D/W gas
phase part (Estimation)

Main R&D Achievements In-house R&D Activities

Improvement and updating of accident progression analysis

PCV pressure(Actual measurement)
PCV pressure (Analyzed data)

Added local flow analysis
model (stratification) in PCV

Added molten lower RPV internal model

Main Achievements and Approaches

Close SRV (Estimation)

RPV pressureThis project is intended to upgrade accident progression analysis
codes (MAAP, SAMPSON) and promote decommissioning by
providing necessary information to identify conditions inside thereactor, conducting comprehensive analysis and evaluation utilizing
accident progression analysis and actual data. This is a joint projectwith the Institute of Applied Energy.

Fallen fuel debris into the bottomof RPV (Analysis result)

Improved primary thermalhydraulic system model

PCV pressure

It is essential to identify conditions inside reactors
for planning the method for fuel debris retrieval anddeveloping safety measures. However, it is difficult to
directly investigate the inside of Units 1-3 at the
Fukushima Daiichi NPS due to extremely highradiation levels.

RPV pressure(Actual measurement)
RPV pressure (Analyzed data)

Start core melt (Analysis result)

Improved core shroud/radiation
model between RPV wall

Molten core

Start water injection
by fire engine

Subsidized Projects

Aims

Open SRV

Jet pump
ShroudImproved items in FY 2015

Background

Decrease RCIC
function (Estimation)

RPV wall

Improved items in FY 2014

Main R&D Achievements

Upgrading for Identifying Conditions Inside the Reactor
through Accident Progression Analysis and Actual Data

Flange forconnecting VESTA
Simulated head mirror
plate lower RPV head
(SFVQ1A material)

The instrumentation piping was melted inside
a pot (equivalent to actual lower plenum).
Corium was leaked through inside theinstrumentation piping and a part of corium(approx.. 40 kg) was fallen; however, theinstrumentation piping was not penetrated.
After that, corium was solidified, leading to
the blocking of the pipe, and then stopped
falling for about 30 sec.. (almost the same
behavior as the analysis data).

Fig. 4: PCV penetration pipe molten failure test
(Simulated instrumentation piping inside the reactor test)
In the Korea Atomic Energy Research Institute (KAERI), a test was conductedusing a full-scale penetration pipe and actual corium, which is equivalent tofuel debris; the damaged situation for the instrumentation piping inside thereactor was also confirmed.

Future Developments
The reliability of comprehensive analysis and assessment will be advanced by improving the accuracy of prediction forprogressing events through sensitivity analysis based on actual data information. Additionally, BSAF Phase-2 project anddatabase continue to be managed and organized. Upgrading for identifying conditions inside the reactor will be promoted.

※Utilized joint research results by TOSHIBA and Nagoya University.

14

IRID Annual Research Report 201515

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

In order to select effective methods of fuel retrieval, it isimportant to understand conditions inside the reactors including
the distribution of fuel debris.
However, it is difficult to verify the conditions directly since theradiation level inside the RPV is too high for workers to access thedebris location. Therefore, the tomography using cosmic-ray muonis expected to provide us with information inside the reactor.

The Muon tomography consists of transmission and scattering
methods. The transmission method is implemented readily at thesite and enables to identify fuel debris at Unit 1 at the Fukushima
Daiichi NPS at the spatial resolution of about 1 m. Furthermore, ameasurement system that uses the Muon scattering method will be
developed to confirm the location of the remaining fuel debris in thereactor at the spatial resolution of about 30 cm.

a) Surface of
two-layered scintillator

Fig. 2: Overview of a small detector for the transmission method
Fig. 1: Detection system for the transmission method used for measurement in Unit 1
Three units of detectors are housed in an enclosure with iron walls 10 cm
thick. The shape is approx. 2.5 m in length, 2.0 m in width and 2.1 m in height.

Point 2
(north side)

Position resolution enables to quadruple by shifting a cubic foot of scintillator
elements 5 mm the direction toward both of x and y to make two-layer structure;
therefore, the resolution can be obtained same as the original even if size of adetector is half of its length and width and distance between detector units islessened by half.

Point 3(center of the north side)

Observation data (Expectancy)
No fuel and water exist

Point 1
(northwest side)

Only water exists

Main Achievements and Approaches

Small-scale verification test by the transmission method
at Unit 1 (at the spatial resolution of about 1 m)

a- Improvement of a detection system for the transmission method to be applied on-site
In order to measure muon even in high radioactive environment (0.4 mSv/h), the detection system that was used at Unit 1 wasstored in an enclosure with iron shielding. The detection system is comprised of three units to reduce the effects of gamma
radiation by conducting coincidence counting (Fig. 1). On the other hand, for easy handling and less spatial restriction on-site, thecompact system, which sizes a fourth of the current whole system, was developed by improving the surface of the detection
system without reducing the resolution (Fig. 2).

Fig. 3: Position of detection system
for transmission method

A north side PCV
wall of thereactor bldg.

Although the place to be measured
was limited due to on-site conditions,
measurement was carried out at threelocations in the northwestern area of
the Unit1 bldg.

Horizontal direction(m)Fig. 4: Image of observation data (left fig.) and distribution oftransmissivity (right fig.)

Fig. 5: Comparison with simulatedtransmissivity measurement

The black area shows a small transmission amount of Muon while
the white part has a large transmission amount. The observation
data represents a distribution of transmissivity inside the reactor.

The transmission amount of Muon, which
passed through inside the bldg., waspredicted by simulation and was comparedwith the observation data. It is concluded that
fuel hardly existed in the reactor core.

Measurment range
Muon tracking device system②

b- Development of an estimation technique of fuel debris distribution inside the reactor by muon tracking
Improved technique was developed as an estimation method of fuel debris distribution by adding a unique noise suppression
technology to 3-D reconstruction algorithm that is used for CT technology in the medical field. The improvement ensured thecapability to estimate fuel debris distribution with better detection rate (rate of determining the fuel debris in areas where fuel debris
is inexistent), even though measuring time was shortened (Fig. 7). In addition, configurations of fuel debris that were deposited in
the lower part of the RPV were clearly confirmed, applying the technique to remove scattered components of the building structure(Fig. 8).

16

IRID Annual Research Report 2015

Measuring for 23 days

Fuel debris (red)(Cross sec. 60 x 60 cm)
Stainless steel

Measuring for 3 months

Fuel debris(deposit thickness 60 cm)

Simulated fuel assemblies

Simulation model

Result fromthe 1 hr. measurement
Lead

Conventional method

Lead block

Wrong detection rate
= 45%
Concrete

Polyethylene

a- Development of a muon measurement system under high radiation (less than 50 μSv/h)Since the detectors for the scattering method are large, an electric circuit incorporating gamma ray-eliminating logic in signal
processing was developed to reduce the amount of shielding materials. The 7 x 7 m muon tracking device system that incorporates
this electric circuit was installed. A measurement test was conducted using two muon tracking device (upper/lower) placed tosurround the simulated core of the Fukushima Daiichi NPS – lead block (simulated fuel), concrete block (biological shield, building),
and polyethylene block (simulated water) (Fig. 6). As a result, it was confirmed that uranium can be visualized at the muon
scattering angle not affected by the lead structure. Additionally, functionality of the algorithm to adjust the position of the detector
was confirmed by the measurement results of the system.

Concreteblock

Simulated fuel assemblies(Filling in container)

Lead

Wrong detection rate
= 22%

Improved method
Measuring for 23 days

Measuring for 3 months

Wrong detection rate
= 15%

Wrong detection rate
= 14%

Simulation resultin the last FY

Data Sheet

Design and manufacturing for detection system throughscattering method (at the spatial resolution of about 30 cm)

Polyethylene
block

Others2

Muon tracking device system①

Fuel and water exist

Reactor core
RPVPCV

RPV and Fuel pool Isolation
biologicalcondenser (IC)shield wallb- Measurement and assessment by the transmission method at Unit 1
In order to verify the capability to detect the remaining fuel debris in the reactor or spent fuel stored in the spent fuel pool at the
spatial resolution of about 1 m, measurement was conducted for a period of three months at three locations in Unit 1 (Fig. 3). As aresult, spatial resolution of about 1 m was ensured through assessment of spent fuel in the pool. On the other hand, high density
structures such as the Reactor Pressure Vessel (RPV) and the Primary Containment Vessel (PCV) were identified throughmeasurement of data, however high density large materials were not identified in the reactor core where fuels were supposed toremain. Therefore, it is concluded that there are few fuels remaining in the reactor core (Fig. 4 & 5).

Only fuel exists

Main R&D Achievements In-house R&D Activities

1b) Assembled small detector

Transmissivity

Aims

Subsidized Projects

Background

Main R&D Achievements

Development of Technology for Detectionof Fuel Debris in the Reactor

Simulation resultin present FY

Simulated fuel assemblies(Empty in containe)

Simulated fuel assemblies(Simulated fuel made in lead)

Fig. 6: Measurement results of the transmission method by thedeveloped system
The figures are layout for two Muon tracking device systems(upper/lower) and simulated materials of the reactor core: a lead block(simulated fuel), a concrete block (biological shield and the bldg..) anda polyethylene block (simulated water) at the Fukushima Daiichi NPS.
The measurement of about 1 hr. reveals the distribution of scattering
angle in the layout.

Eliminating the effectsof the bldgs.

Fig. 7: Shortening of measuring time

Fig. 8: Effect reduction in the bldg.

The figures show the comparison among
simulation results of debris distribution
between the conventional method and theimproved method applied to ML-EM
method used in the medical field. The
measuring period is approx. 3 months
and 23 days, which is equivalent to1/4 of
3 months.

The figure shows simulation result of the simulation
model which has arranged stainless steel to comparewith fuel debris in RPV. Compared with thesimulation result in the last FY, this FY’s simulation
result confirms the deposits of fuel debris in thelower RPV with higher contrast by reducing
effects of bldgs.

Future Developments
According to the measurement results at Unit 1, the transmission method confirmed its applicability for the measurement
technology of fuel debris distribution under high radiation environment. On the scattering method, a large detection system has
been completed for implementation and confirmed to identify fuel debris with high detection capability. Both systems of thetransmission method and scattering method will be applied to on-site use accordingly, considering the site situation and progressof identifying conditions inside the reactor.
17

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

In-core device

Development of Fundamental Technology
for Retrieval of Fuel Debris/Reactor Internals

Approx. 11 m
Internalof RPV
Individual device

It is assumed that fuel debris in the Reactor Pressure
Vessel (RPV) and the Primary Containment Vessel(PCV) at the Fukushima Daiichi NPS has not currently
reached criticality. However, reactor buildings, the RPV
and the PCV suffered damage at the accident, and thereactors are in unstable conditions. Therefore, it isnecessary that fuel debris be retrieved to maintainsubcritical conditions and become stable condition by
preventing diffusion of radioactive materials.

The feasible R&D continues toward the determination of the fueldebris retrieval methods (Summer 2017), finalization of retrieval
methods from the unit to be firstly retrieved (the first half of 2018) andcommencement of fuel debris retrieval from the unit to be firstly
retrieved (around December 2021). In order to verify the feasibility of
the fuel debris retrieval methods, element tests will be conducted toobtain the necessary information/data for three methods:
Submersion-top entry method, Partial submersion-top entry method
and Partial submersion-side entry method under this project.

12

Fig. 1: Conceptual diagram of a plan for the scale model testing facility
The study to prevent the spread of contamination will be proceeded by a scalemodel test.

Element tests required for the feasibility of the methods

Lower device sealing(Swivel drive)
Open/close port
Lower device sealing(Open/close port)

Rail
Robot arm

Hydraulic manipulator

Upper table

Electromagnetic valve

Lower table

Simulated pedestal
Image of the test

Fig. 3: Photo of test device for hydraulicmanipulator

Fig. 4: Diagram of the access device inside
RPV (plan)

Fig. 5: Mock-up test facility for the access device
inside the pedestal and test condition (Image)In accordance with a simulating actual remote
distance, the test device checks the operational
error.

Comparison study is performed for the supporting
method to a horizontal direction for a lower tablethat is hung with a wire.

The feasibility of Partial submersion-side entry
method will be verified by simulating an actual
arrangement.

Test plan was established for each element test in FY 2015 (includes tests under planning). Based on the test plan, feasibilityof the methods is verified through test manufacturing and element tests accordingly.

Equip. hatch

Personal air lockFig. 6: Remotely operated flexible
structure arm in actual use (Image)a- Prevention Technology against the spread of contamination in retrieving contaminated large structure

It is possible that flexible structure arm canestablish a remotely operated device with avariety of shapes/functions similar to toy blocks.

●In order to verify the prevention technology against the spread of contamination, a small model test is being conducted for each working step (Fig. 1).

b- Prevention technology against the spread of contamination in retrieving fuel debris in the Reactor Pressure Vessel (RPV)

Operating device

Closing deviceof canistere- Prevention technology against spread of contamination in retrieving fuel debris●Test plan of the platform and cell by Submersion method is formulated, and an element test is being planned (Fig. 8).
●For the purpose of remotely seal welding for the cell by partial submersion-side entry method, a test plan and an element testof the PCV welding device are being formulated (Fig. 9).

Cell

Closing deviceof airtight container
ShieldingRail

Mobile-type cell (Example)

Fixed-type cell (Example)

Fig. 8: Mobile-type/fixed-type cell (Image)
Mobile-type cell and fixed-type cell will be compared and evaluated from thepoint of required function and workability among others.
Partial modeld- Technology for remotely operated retrieval of fuel debris●Test of remotely operated flexible structure arm is being conducted (Fig. 6).
●Test plan of handling device of fuel debris canister is formulated, and an element test is being planned (Fig. 7).

Shielding port (floor door)

Data Sheet●A test is being conducted for the hydraulic manipulator (Fig. 3).
●Test plan of the access device inside the RPV is formulated, and an elemental test is being planned under submersion-top
entry method (Fig. 4).
●Test plan of the access device inside the pedestal is formulated and a mock-up test is being planned under partial
submersion-side entry method (Fig. 5).

Cover

Fixing mechanism
Turning mechanism
Inspection
Bolt tightening
mechanismmechanism

Others●Tests are being conducted on the sealing in the access device and in the lower part of the device inside the RPV undersubmersion-side entry method (Fig. 2).

c- Technology for accessing fuel debris

Reactionforce
receiving(6 points)

Swivel drive

Oil pressure source

Coordinating and analyzing results for each element test

Another proposer who partially engages in the operation joins in the R&D aside from IRID for this project. IRID will coordinate
the whole project including element testing that the proposer plans and conducts. Currently, IRID is organizing test plans including
plans by the proposer.

Issues will be clarified during
actual application suchas confirming the sealing
performance through apartial model device.

A-A cross-sec.

Main R&D Achievements In-house R&D Activities

Main Achievements and Approaches

Swiveling table

Subsidized Projects

Aims

Dryer

Fig. 2: Conceptual diagramof the access device inside
RPV (In-core device)

Main R&D Achievements

DS plug

Background

Approx3.5 m

Shield plug
PCV head
RPV head
RPV flange

Internalof RPVInternal sealingin RPV

Moving mechanism

Closing deviceof canister (plan)

Cart forairtight container

Simulated
X-6 penetration

Transport container

Simulated
PCV side

Simulated
access opening

Handling deviceof canister (plan)

Fig. 7: Handling device of fuel debris canister (Image)
Basic design of the device and elemental test items of the closing device will
be clarified.

Fig. 9: Mock test for PCV welding device (Image)
A welding test to be operated remotely will be conducted under the same conditions in
actual use; solutions and feasibility will be confirmed.

f- Technology for reducing exposure of workers in retrieving fuel debris●Test plan of the lightweight shielding and shape to be applied for top entry method is formulated, and an elemental test is being planned.

g- Cutting/dust collection for fuel debris retrieval●Performance test of cutting/dust collection technology for fuel debris retrieval is being conducted18

IRID Annual Research Report 2015

Future Developments
In order to evaluate the feasibility of the methods for the three fuel debris retrieval methods, Submersion-top entry method,
Partial submersion-top entry method and Partial submersion-side entry method, elemental tests will be conducted until the end of
FY 2016. Through testing, we will identify and examine further issues, and then discuss action plans.
19

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Upgrading Approach and System for the Retrievalof Fuel Debris and Internal Structures

Upper container

Operating house

Aims

It is assumed that fuel debris in the Reactor Pressure
Vessel (RPV) and the Primary Containment Vessel (PCV) at
the Fukushima Daiichi NPS has not currently reachedcriticality. However, the reactor buildings, the RPV and thePCV were damaged at the accident, and the reactors havebeen in unstable conditions. Therefore, it is necessary to
retrieve fuel debris to maintain subcriticality and become in astable condition by preventing diffusion of radioactive
materials.

In order to retrieve fuel debris, three retrieval methods arechosen; Submersion-top entry method, Partial submersion-top
entry method and Partial submersion-side entry method. Toward
finalizing the retrieval methods of fuel debris and reactor
internals, plant-related information will be examined, retrieval
methods for fuel debris and reactor internals, system andequipment studied, feasibility of the retrieval methods and thewhole system verified. Also, development plan of retrieval systemand equipment will be formulated.

Gate

Subsidized Projects

Background

Main R&D Achievements

Ceiling crane

Gate

1

Clarifying plant information to determine the retrieval policy
of fuel debris/reactor internals

Storage canister

Transport container
Fuel debris
retrieval equip.

Plant data and development results of related projects were organized, and vital information was examined to determine theretrieval methods for fuel debris.

2

Main R&D Achievements In-house R&D Activities

Main Achievements and Approaches

Fuel debris

Study of retrieval methods for fuel debris/reactor internals,
system and equipment

R/B Carry-in entrance
for large objects

a- Study of the feasibility of the methods
Fig. 1: Flow chart of fuel debris retrieval by Partial submersion-top entry method (Image)This chart shows the flow of Partial submersion-Top entry method; fuel debris is retrieved from the RPV bottom, collected in a canister andloaded into a vehicle to transfer outside of the reactor building (R/B).
The flow chart details how to deal with actual fuel debris and extracts technical challenges to be analyzed.

OthersCarry-in/disposal
container

Waste storage bldg.

In order to ensure safety for fuel debris retrieval, it is necessary to install systems such as cooling equipment to maintain thecirculating and cooling water for fuel debris and the equipment to prevent dust scattering including radioactive particles that would
generate from cutting fuel debris. Safety criteria and requirements for the system are reviewed.

3

Collection Container
Storage canister
Carry-in collection container/
Storage canister
Fuel debris retrieval cell

Storage canister handling cell

Carry-out cell

Head of transport
container

Manipulator

Powdery debris
Manipulator

Movable door
Mini-cask(Weighting control)

Storage canister(blue line)

Future Developments

IRID Annual Research Report 2015

Cell under an atmosphere of N2
Flow of debris (Storage canister, transport container)
Flow of waste
Flow of other equip.

additional building

In order to formulate development plans for the retrieval system of fuel debris and reactor internals and the equipment, technicalchallenges are being clarified.

20

Maintenance/
decontamination cell

Reactor bldg.

Formulation of development plans of device/system
for retrieval of fuel debris/reactor internals

Following FY 2015, we continue to study for promotion of plans in response to the changing situation, new challenges andrequirements.
This achievement will serve as useful information to determine the retrieval policy of fuel debris, finalize the retrieval method andcommence retrieval for the unit which will be firstly retrieved from, based on the Mid-and-Long Term Road Map (revised in June2015).

Canister storage bldg.

Carry-out/
in cellb- Conceptual study of the system

Data Sheet

The retrieval methods were classified into 12 types in combination with the water level that fuel debris would be retrieved from (roughly
divided into Submersion method and Partial submersion method) and the direction that fuel debris would be accessed from (entry from top,side and bottom). In order to prove technical challenges from the three representative methods (Submersion-top entry method, Partial
submersion-top entry method and Partial submersion-side entry method), a flow chart (Fig. 1 & 2) and a working step chart summarizing thechanging conditions of the plant, the flow of each working step and treatment of retrieved fuel debris and wastes.
After identifying the issues to be solved, assessment of the feasibility and development plans for resolution will be designed.
In alignment with related projects such as the development of canister, scope of each study was confirmed, latest information was shared,
the common challenges were clarified, and preconditions/hypothetical conditions were set up.

Storage canister transport system (Mini-cask)

Robot arm

Transport
container
Inside the additional bldg. (outside of cell)

Carry-intransport
container

Collection Container

Fig. 2: Flow chart of fuel debris retrieval by Partial submersion-side entry method (Image)
The chart shows a flow of fuel debris transfer by Partial submersion-side entry method; fuel debris in the PCV bottom is transferred to the side, stored
in the canister and loaded into vehicle for transferring. Fuel debris is safely transferred in the order shown inside the red square frames. In order torealize the flow, technical challenges and issues will be clarified and analyzed.

21

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Collection,
Transfer and Storage of Fuel Debris
Fuel debris retrieved from the Fukushima Daiichi NPS is
expected to be carried out in the reactor buildings and storedsafely until the time when the method of final disposal isdetermined. For this reason, system required for thecollection, transfer and storage of fuel debris must be
established.

Based on existing technologies for transportation andstorage of spent fuel, fuel debris storage canisters (hereinafter
called “storage canister”) and canister handling device for safe
collection will be developed. Materials and structure of storage
canisters will be studied, and basic specifications of thecanister will be created in FY 2015.

Main Achievements and Approaches
1

In order to investigate for the design of storage canisters, subcriticality evaluation technology, evaluation technology forhydrogen generation, sludge collection technology, drying technology for vacuum and technical information were obtained from avisit to Paks Nuclear Power Plant in Hungary, the Pacific North West National Laboratory (PNNL) and the Hanford Site in the
United States (Fig. 1 & 2).

2

HEPA filter assembly

A storage canister at Paks

Structure to release hydrogen（Compensator）
Installing at the topof a storage canisterwhen storing in a poolFig. 1: A storage canister used at Paks NPS
Technical information was obtained for collection, transfer andstorage of damaged fuel at Paks NPS.

Ref.：L. Szöke, Managing of the damaged fuel at Paks NPP, Meeting in
Vienna on Management of Severely Damaged Spent Fuel And Corium
2013

Ref.: 1)R.McCormack,Multi-Canister Overpack, Nuclear Waste Technical
Review Board, 2014
2)N Reactor (U-metal) Fuel Characteristics for Disposal Criticality
Analysis, DOE/SNF/REP-056, Rev0, 2000

Storage canister
Drain pipe

Development of safety evaluation methods
4000

OthersDevelopment of technology for transfer and storage of the canister

In order to study the equipment handling storage canisters, the report entitled “Flow of the Collection, Transfer and Storage for the
Fuel Debris Storage Canister” formulated in FY 2014 was updated based on the latest situation on-site and knowledge obtained fromthe project (Fig. 5).

Load cellFig. 3: Example of a storage canister drop test in the vertical direction
Posited on an occurrence of an accident during the handling of storage canister, a drop
test in the vertical direction was conducted by structure evaluation method.

Fig. 4: Basic design of a storage canister (Example)
Based on the information and requirements acquired, the basic design ofa storage canister was examined. The spec. and shape are continuously
updated depending on the results of the safety evaluation and the study
on the handling of storage canister when retrieving fuel debris.

Reactor bldg.
Maintenance Carry-out portcell(additional)

Future Developments

Storage cell forfuel debris retrieval

Transport cell

Storage facility

Fuel retrieval

IRID Annual Research Report 2015

Debris transport bldg.

Handling cell fora storage canister

The feasibility of basic specifications for the storage canister will be verified in cooperation with other R&D projects based onachievements in FY 2015. Furthermore, in order to solve clarified issues, the basic design of storage canisters will be performed
through analyses, evaluations and element tests.

22

Data Sheet

Φ220

Development of technology for fuel debris storage

Based on the knowledge obtained from R&D of the ongoing project, we have examined requirements for the design of storage
canisters and reviewed basic specifications of the storage canister. Future challenges were also clarified (Fig. 4).

5Filter preventing
the retentionof hydrogen

Head

Fuel debris in the Fukushima Daiichi NPS consists of products generated from the Molten Core Concrete Interaction (MCCI) and seawater
substances, unlike the actual fuel debris from the Three-Mile Island Unit 2 (TMI-2) Reactor. In addition, for the effective and safe collection, transportand storage of large amount of fuel debris, it is necessary to enlarge the internal diameter of storage canisters, and handle them through remote
operation. In order to reflect these matters in the design of storage canisters, reviewing evaluation methods through investigation of documents,
survey of overseas case, tentative analysis and conducting tests, items for detail study were determined on evaluation for the amount of hydrogengenerated (countermeasures for hydrogen generation), evaluation on material aging degradation (material selection of the storage canister), criticality
evaluation (subcriticality design) and structural strength evaluation of storage canisters (Fig. 3).

4

Head2）

Fig. 2: Storage canister used at Hanford Site in U.S. (Ref.)
Technical information was obtained for collection, transfer
and storage of metal uranium damaged fuel caused bydecommissioning of pool at Hanford Site.

Study on the fuel debris storage system

Based on the knowledge obtained from the latest situation on-site, information from other R&D projects and this project – concept of fueldebris storage that was studied in FY 2014 – was verified. Results confirmed that the current storage concept was not necessary to review.

3

Whole image1）

Main R&D Achievements In-house R&D Activities

Investigation and establishment of a research planfor the transfer and storage of damaged fuelSubsidized Projects

Aims

Main R&D Achievements

Background

Rupture disk

Storing in canister

Washinga canister

Storing canister ina transport container

Carrying out
a transport container

Fig. 5: Flow chart for thecollection, transfer and storage
for a storage canister (Fig.
shows non-submerged fromside method from the side of
the reactor.)
In preparation for the study
on required equipment fortransfer, the flow of handlingcanister in FY 2014 wasreviewed. More updates will
follow.

23

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Evaluation of submersion method feasibility based on the seismic
integrity of the PCV/RPV

The feasibility of dry/full submersion methods of fuel debris retrieval (entry from top) was examined through a seismic-resistanceevaluation for the PCV/RPV equipment, which reflected assumed reactor conditions. For areas that showed severe result on theevaluation, a detailed evaluation of seismic resistance was conducted. Additionally, assuming that the suppression chamber (S/C) support
is not reinforced, a detailed evaluation of seismic resistance was conducted (Fig. 1).

2

Simple evaluation of the seismic resistance of equipment considering repairs(water leakage stoppage) and water level rise in the PCV

Seismic response analysis was conducted using the selected parameters (water level in the PCV) that impact on seismic response analysis.
Response comparison of seismic load due to parameters variation was examined as well. Furthermore, a simple evaluation method was developed
for the seismic resistance of the equipment in combination with various methods. Results of the evaluations as compared with those of the simple
evaluation method and regular active analysis were almost same, and the validity of the simple evaluation method was confirmed.

Development of anti-corrosion measures

4③Seismic evaluation for equipment
under calculated loading conditions【 Equipment to be evaluated】

・Equipment-related function maintenance for PCV submersion boundary
and injection cooling facility
・Excluding equipment to be repaired for development of PCV repairing
technology

Sample material
Test temperature
Test time
Artificial seawater□：diluted 200 x○：diluted1,000 x△：diluted10,000 x

1,000 2,000 3,000 4,000
Concentration of the addition④Full evaluation of equipment whose tolerance is low【Full evaluation method】

・Evaluation of simulated reactor by finite element method【To be implemented】
Fig. 1: Flow of PCV submersion method feasibility evaluation

・Test specimen presented a metallic gloss by adding
3,500 ppm seawater diluted 200 x and 2,000 ppm
seawater diluted 1,000/10,000 x. Corrosion amount
was almost 0 mg/cm2（Confirmed anti-corrosion effects on carbon steel）

Fig. 2 Example of validation test for anti-corrosion effects
of rust-preventive agents（Test result of adding zinc / sodium molybdate mixing phosphate）

Data Sheet

Results from tests evaluating rust inhibiting under high levels of radiation exposure after selecting rust-preventive agents and sufficient
information was obtained to judge their applicability on-site. Evaluation tests of the rust-preventive agents on the effect on the function ofwater treatment equipment were conducted, and technical issues were clarified (Fig. 2).

・Two cases based on H25-2 and H25-3 in FY 2013
・Basis ground motion (Conventional Ss: Max. acceleration 600Gal)
・Seismic response analysis model of the reactor bldg.: A model used in FY 2013

Others3【Conditions for seismic response analysis】

Main R&D Achievements In-house R&D Activities

1②Formation and analysis of seismic response analysis model(bldg..-equipment) in addition to the new plant informationsuch as repairing (reinforcement), fuel debris facility and others

Corrosion amount

Main Achievements and Approaches

・1F-1: Column support
・1F-2/3: Reactor shielding wall, column support, seismic support

Artificial seawater diluted 200 x

In order to evaluate the feasibility of Submersion method,urgent seismic-resistance evaluations are required accordingly
in response to assumed various reactor situations. In addition,
considering seismic-resistance evaluations for corrosion thinning
amount for a long period of time, the integrity of the PCV/RPV will
be evaluated. Corrosion control measures will be also studied tomaintain the integrity.

Artificial seawater diluted1,000 x

Due to the devastating Great East Japan Earthquake, thedeterioration of the structural materials at the Reactor Pressure
Vessel (RPV) and the Primary Containment Vessel (PCV) at
the Fukushima Daiichi NPS, which was caused by hightemperatures, injection of seawater and fallen debris, became theconcern. Action plan is necessary to maintain the structural
integrity of the PCV and the RPV over the long period of retrieving
fuel debris from the reactor core.

①Organizing countermeasures for the repairing (reinforcement)of equipment that was not able to ensure tolerance at the timeof evaluation in FY 2013
【Equipment. that was not able to ensure tolerance orequipment whose tolerance was low.】

No corrosion

Artificial seawater diluted10,000 x

Aims【Completed】

General corrosion or local corrosion

Subsidized Projects

Background

Flow of feasibility evaluation
for PCV submersion method

Main R&D Achievements

Development of Technology for Integrity Evaluationof the Reactor Pressure Vessel/Primary Containment Vessel

Upgrading predictability of long-term corrosion thinning rate

In order to improve accuracy prediction of corrosion thinning rate, a long-term corrosion test for 10,000 hours was conducted. Additionally, newknowledge was obtained through investigation of corrosion impact and leaching components from fuel debris and concrete inside the reactor.

5

Evaluation of pedestal erosion impact

Tests such as concrete strength test and rebar corrosion test were conducted under high temperature heating and air/underwater
exposure conditions to obtain various data on cylindrical specimen, small-scaled mock-up test body and block specimen among others.
Through these tests, effective information was obtained for further study (Fig. 3).

Future Developments
In order to implement on-site, it is necessary to select a rust inhibitor that would not cause leakage due to local corrosion such as crevice corrosion.
Since an immersion test for 500 hours is not able to evaluate the applicability, it is necessary to evaluate a local corrosion occurrence and its progression
through electrochemical measurements that do not depend on the evaluation time. Furthermore, since there is a concern that precipitation film type ofrust inhibitor (phosphate) would adhere on high temperature area of the fuel debris surface, its impact requires to be evaluated.

24

IRID Annual Research Report 2015

PC steel rods foraxial force introduction(4 rods in total)

Axial force loading frame

Test device
Thermal insulation material isseen in white and a heater
is installed inside.

Fig. 3: Proof stress evaluation testof the RPV pedestal (Installation
loading frame)
A cylindrical test device simulating
the RPV pedestal, which was
injected with cooling water after
being exposed to high temperature
at the accident, was loaded toconfirm tolerance of concrete
structures, which deteriorated due
to high temperature, and theimpact on the destructiveness.
(After drying, the cylindrical test
device is heated at 400℃ and
800℃ for prolonged time andimmersed in water.)
Outside diameter: 1,240 mm, Inner
diameter: 834 mm, thickness: 200mm

25

Main R&D Achievements
Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Concentration degree Gd concentration
Unit 1

Criticality scenario and major risk
・No debris and criticality risk is min.
or none.

No debris in the reactor core.

Aims

It is estimated that fuel debris is not currently in a criticalstate. However, since the shape of fuel debris and water levelmay change during retrieval of fuel debris in the future, it isnecessary to develop criticality scenarios, criticality
evaluation, monitoring technology and criticality prevention
technology accordingly.

Criticality scenarios for multiple methods (i.e. Fullsubmersion, Partial submersion, Partial dry and Dry) will beclarified in FY 2015. In addition, preventing technology forcriticality using neutron absorption materials and monitoring
technology such as criticality detection will be developed;
criticality control will be established to maintain subcriticality,
and recriticality detection technologies will be developed toprevent excessive exposure, even if criticality occurs.

Although the RPV bottom isdamaged and most of debris
had fallen into the PCV bottom,
a part of debris may remain.

Criticality scenario and major risk
・At the time of PCV submersion:
Exposed debris submersion : Low
・At the time of debris retrieval:
Criticality risk s min..
・No criticality risk before PCV submersion
starts.
(Other risks are min..)

Almost total amount of debris may
have fallen into the PCV bottom.

Criticality scenario and major risk
・At the time of PCV submersion:
Shape change of debris: Low
Granular debris curling up: Low
・At the time of debris retrieval:
Shape change of debris: Low
Granular debris curling up: Low
・Criticality risk is min. before PCV
submersion starts.
(Other risks are min..)

MCCI may be caused.

Current water level of the PCV
bottom is approx. 2.8 m.

Min. burnup fuelUnit 2
assemblies ave.
（Unit 3） (evaluated in Unit 2)

Technology for reactor recriticality detection

Technology for criticality prevention“Non-soluble neutron absorbing material” is a technology that prevents criticality by supporting soluble neutron absorbing materials
through direct application on the debris. Continuing with FY 2014, in order to select candidate materials, irradiation tests for new candidate
materials and improved candidate materials were conducted. In addition, required input of “non-soluble neutron absorbing material” wasevaluated while the conceptual design of applicable equipment conducted.
“Soluble neutron absorbing material” will be widely used to prevent criticality between the PCV submersion and retrieval of fuel debris. In
order to confirm the effect the application of this material will have on the integrity of the PCV and the Reactor Pressure Vessel (RPV),
corrosion testing using sodium pentaborate in the low concentration region was performed. Result confirmed that significant corrosion was
not observed with a concentration of over 2,000 ppmB, and scope of applicability was clarified with tests in the high concentration region(Fig. 3). Additionally, regarding the applicability of the “soluble neutron absorbing material” to an actual plant, the conceptual study for water
quality control facility, taking into account the impact on the nuclide removal system, was performed.

Fig. 1: Result of criticality risk evaluation at Unit
The criticality risk was evaluated based on the estimatedconditions inside the reactor of Unit 1.
As a result, it proved that the criticality risk is low at the time of
PCV submersion when there is almost no fuel debris in the reactor
core and the RPV bottom.
Measuring neutron released from fuel debris
Subcriticality: Shallow

Passive method(noise)Subcriticality: Deep
Neutron sources areapplicable at the site.

Not applicable

Active method(neutron source multiplication)

Virtual neutron
capturing method

Estimating fuel debris multiplication factor before detection
Monitoring neutron counting rate and the reactor period during detection.
Phenomenon in which
the reactor period is small.

5,150
2,550

Criticality
controlmethods

Pure water
Submerging graduallywhile confirming
subcriticality condition

Debris retrieval

Boric acid solution
Same as left

・Criticality
prevention due
to boric acid

Criticality
prevention
system(PS)

Submerging graduallyat limited speed while
confirming subcriticality
condition

Mitigation systemof the impact
at the time ofcriticality(MS)Detection by gas sampling fission productsystem FP gamma radiation detector systemand terminating criticality by injecting boric
acid solution

Retrieving undersubcriticality

・Monitoring subcriticalityof criticality approachdetection system nearby
debris・Criticality prevention bysoluble/insoluble neutron
absorbing material

Same as lef

Table 2: Basic idea of criticality control
Two cases are considered for the submersing of PCV: one is to submerge
gradually injecting pure water while confirming subcriticality condition and theother is to submerge using boric acid solution while prevent criticality.
Neutron absorbent will be used accordingly as criticality approachdetection system is operated during fuel debris retrieval.

1-34Left side

Data Sheet

3

10,100

Others“Gas sampling fission product gamma radiation detector system,” which detect recriticality, was confirmed to be applicable inmultiple methods.
Additionally, in order to confirm the feasibility of the system, a verification test plan for subcriticality estimation algorithm in actual
use was designed.
“Neutron detection system” was studied for its application as a backup method.
In regards to the “in-core criticality approach detection system,” which detect abnormalities before criticality is reached, conceptof the detection system was established in combination of several methods (Fig. 2). In order to confirm the feasibility of thedetection system, a test plan that used a storage facility of spent fuel and critical assembly was developed.

No Gd
Ave. in the reactor
core (Unit 3)
Ave. in the reactor
core ((Unit 2)Main R&D Achievements In-house R&D Activities

2

6,400

PCV submersion

Evaluation of Criticality

Criticality scenarios, which reflected the latest information concerning various construction methods, were reviewed. Criticality risks were
evaluated for each unit and part (Fig. 1). Criticality behavior evaluation and exposure assessment on the remaining fuel in the reactor core,
which has an increasing criticality risk, were conducted by submerging the upper level of the Primary Containment Vessel (PCV). During theevaluation, a practical assessment was conducted using detailed fuel compositions of the fuel rod unit after considering the operating history
for each unit. In the result, it showed that submerging method using pure water is deemed applicable, if remaining fuel in the reactor core is
within the expected range. In addition, criticality analysis was conducted considering the proportion of Gadolinium (Gd) in fuel debris by using
the actual data and was studied to streamline criticality evaluation (Table 1).
Basic idea of criticality control for the fuel retrieval methods was studied. The idea of criticality safety control was studied based on defense
in depth at the time of incident/abnormal events. Requirements for the system of fuel debris retrieval were examined (Table 2).

8,500

Table 1: Required boron concentration considering Gd
When analysis was conducted for fuel with low burnup in both cases
where Gd did not remain on fuel originally containing Gd, and where Gd
remains average, it proved that required boron concentration candecrease approx. 6,000 ppmB in order to maintain subcriticality.
Besides, fuel in Unit 2 was evaluated as a representative as fuelconcentration in Unit 3 is lower than in Unit 2.

Main Achievements and Approaches
1

No Gd
Ave. in the reactor core

Subsidized Projects

BackgroundMin. burnup fuelassemblies ave.

Required boron conc. (ppmB)

Main R&D Achievements

Development of Technology for Criticality Control
in Fuel Debris Retrieval

Phenomenon in which
the reactor period is stable.

Rapid phenomenon

Quiet phenomenon

Rapid phenomenon

Quiet phenomenon

Period detection

Counting rate detection

Counting rate detection

Counting rate detection

Fig. 2: Detection process of criticality approach detection system inside the reactor core
Fuel debris multiplication factor is measured by Passive method based on themeasurement of neutron released from fuel debris.
When difficult to measure by Passive method due to a deep subcriticality, a multiplication
factor is estimated by virtual neutron capturing method. Active method is optional in accord
to the current situation. Based on estimated multiplication factor, neutron counting rate andthe limited value of the reactor period (equivalent to the change rate of the counting rate)
are defined, and the counting rate and the period are monitored during detection.

Fig. 3: Soluble neutron absorption material low concentration corrosion test result (example)
Left photo shows results of a 100-hr. test on 1,000 ppmB of sodium
pentaborate for non-coated base material and seawater diluted
1,000x.
Right photo shows test results on 2,000ppmB of sodium
pentaborate; other conditions were same.
Low concentration of 1,000 ppmB of sodium pentaborate partly
caused corrosion; however, it was confirmed that 2,000 ppmB did
not cause corrosion.

Future Developments
In regard to criticality evaluation for the multiple methods of fuel debris retrieval examined in FY 2015, criticality control methods during
the retrieval of fuel debris will be studied while formulating updates and elaboration showing the latest expertise.
Verification tests of the feasibility for the essential technologies such as the “in-core criticality approach detector system” and “non-soluble
neutron absorbing materials” will be conducted. Criticality control methods for the retrieval methods continue to develop, in addition to theestablishment of the essential technology for criticality control. The technologies will be utilized for the finalization of fuel debris retrieval
policy in the summer of FY 2017.
26

IRID Annual Research Report 201527

Main R&D Achievements（B）Macro-property items requiring evaluation

Key Challenge 2 – R&D for Preparation of Fuel Debris Retrieval

Property items required for each project on fuelretrieval

Each part of the（A）reactor internal/
PCV

Fuel Debris Characterization

Upper coreinternals…

Aims

To ensure safe and steady implementation for retrieval,collection, transfer and storage of fuel debris toward
decommissioning, fuel debris mechanical properties such as
hardness, thermal properties and moisture characteristics
must be fully understood through analysis, investigation,
assumption and evaluation.

In order to obtain and share the necessary information for the study offuel debris retrieval methods, development of collection, transfer andstorage of fuel debris, and fuel debris characteristics will be estimated by
testing and analysis using simulated debris. Furthermore, analysis
scenario will be studied, and elemental analysis technologies will be
developed to achieve fuel debris analysis.

Property Mixing rate…

…

（C）Macroproperty for each part…
PedestalbottomAs a result of R&D, macroproperty of fuel debris
distributed in the reactor/PCV was estimatedand clarified as an identified list.

① Homogeneous molten part in deposits

Surfacesolidification

Melt

Melt

Arc melting of material mixture

Based on the needs for each project on the fueldebris retrieval, required property item (B) was set
up. Items based on the property data for estimatedarea of debris distribution (A) will be completed.
Property data is estimated from the data obtainedin experiment, thermodynamic analysis results,
existent information, etc.

② Layered structure of interface layerCore materials are locally heatedby light collection on the concrete(under the temperature gradient).

ConcretecomponentCore component

Main Achievements and Approaches
1a- Understanding the data on debris characteristics

In cooperation with France’s Alternative Energies and Atomic Commission (CEA), we continued to collect data on thecharacteristics of chemical compositions, hardness, and porosity by using MCCI test products that CEA had produced in the past.
Furthermore, metal ceramic solidified products containing UO2 were produced in cooperation with National Nuclear Center (NNC) of
the Republic of Kazakhstan (Fig. 4),obtaining knowledge on hardness, interfacial morphology between metals and ceramics andparticle distribution, which contribute to criticality control and development of fuel debris retrieval equipment.

3

Development of elemental technologies for fuel debris analysis

In regards to analysis work of fuel debris, a work flow from sample collection to analysis was examined, and a development plan
was formulated to extract factors of technological development. In addition to the study for quantitative analysis methods of element
through ICP-AES (Inductively Coupled Plasma-Atomic Emission Spectrometry) for fuel debris that was melted through alkali fusion
technique, the effect of spectral interference and ionization interference by a coexistent element was evaluated. A transport study
including fuel debris derived from MOX fuel was performed.

28

IRID Annual Research Report 2015

1,500〜1,800 K(<1,500 K)

Obtaining the data of the layer structure under the temperature gradientFig. 2: Production method for two kinds of simulated MCCI products according to the test purpose
MCCI products are generated from reaction of molten core fallen to the RCV bottom and concrete of the reactor. The property data of formation layer (Fig.①) produced the simulated
MCCI product through homogeneous melting by using arc melting and obtained the data such as a structure and hardness. In addition, the data of inhomogeneous layer structure
under the temperature gradient (Fig.②) produced the simulated MCCI product by the local collecting/heating and obtained the behavior data such as Gd and FP elements.
Rock shape (slow-cooling condition)（１）
Rock-shaped
solidification

Fuel debris

A part in which
the stainless 1 0 0 m m
steel is melting

Loose debris inupper core internals

Loose debris inlower core internals

Hard debris in the lower head

Grain boundary precipitates

Parent phase
Stoma

200 m

Indentationin Vickers
hardness
Hardness ofparent phase
12〜15 GPa
Same asthe upper crust【Sample in
FY 2015】
Upper
crust
moltenpool(2 types)

20 m

SEM image of hard debris in the lower headFig. 3: SEM observation results of TMI-2 debris appearance/ cross-section
Three kinds of debris among TMI-2 debris stored in JAEA; appearance of loose debris in upper core internals,
loose debris in lower core internals and hard debris in the lower head were observed, their hardness measured
through SEM/EPMA analysis. According to the results of SEM/EPMA analysis, it was verified that parent phaseof the lower head debris was mainly formed cubic (U, Zr, Fe, Cr, O2), Fe-,Cr-,AI-,Ni- and O-based oxide wasdeposited on the grain boundary and hardness was almost same as the upper crust.

50m m

Powdery/granular shape (water-cooling condition)（２）
Powdery/granular
Fixed ring shaped solidification
Fuel debris

Stainlesssteel plate

100mm

Data Sheetc- Evaluation on unevenness of properties

Thermal degradation part

Obtaining data for formation layer, structure and hardness

b- Comparison with TMI-2 debris
In order to compare and examine the data between the characteristics of the obtained fuel debris mentioned in the above 2.a andthe characteristics of fuel debris collected from the Three Mile Island Unit 2 (TMI-2) accident, metal structure was observed, andVickers hardness was measured using three kinds of TMI-2 debris including accumulated debris on the lower head (Fig. 3). In
addition, it is confirmed that TMI-2 debris in the pool can be melted through alkali fusion technique and ammonium salt fusion
technique that are considered to be effective melting methods for fuel debris.

Image of core melt deposits on the concrete

OthersIn order to expand the data on mechanical fuel debris properties, Zr (O) (solid solution of oxygen and zirconium), using the amount ofoxygen as a parameter, was fabricated. Mechanical properties (hardness, elasticity ratio and fracture toughness) were measured andevaluated. In addition, (U, Zr) O2, which was mixed into the Stainless Steel (SUS) components and fission product (FP) elements to form asolid solution, was fabricated; the formation layer and its mechanical properties were measured and evaluated. Then, the impact of theelements such as FP that can be dissolved to create solid solution was clarified, and necessary data on oxide-based fuel debris
characteristics were compiled. Furthermore, in order to evaluate element behavior like FP elements of molten core concrete interaction(MCCI) products, simulated MCCI products in homogeneous melting were produced by arc melting. To evaluate the formation layer,
structure and mechanical properties, simulated MCCI products, having heterogeneous (layered) structure under temperature gradientthrough locally focused radiation melting, were also produced to clarify the characteristics of the behavior of FP elements (Fig. 2).
Test focusing on the hydrous/dry properties that impact on hydrogen generation at the time of collection/storage of fuel debris, wasconducted using UO2, ZrO2 and cement paste. Additionally, property change due to oxidizing velocity/oxidation of fuel debris under theconditions of the atmosphere, which easily progress oxidation, was evaluated by using simulated MOX debris.

Inhomogeneous melting

Concrete melt
(molten glass)

Main R&D Achievements In-house R&D Activities

Evaluation of the characteristics using simulated debris-2,800 K
Pyrolysis gas(mainly H2O)

Thermal degradation(embrittlement)

Based on the survey of the project needs for the preparation of fuel debris retrieval, fuel debris properties for evaluation were clarified;micro-properties/macro-properties of fuel debris were estimated, and the data compiled (Fig. 1).

2

Completely melted
and mixed

Concrete

Assumption of fuel debris characterization inside reactor

Subsidized Projects

Background

Qualities of materials
or phase

Main R&D Achievements

Estimated area of debris
distribution

Mass

Fig. 1: Diagram of fuel debris characterization(sectional)

50mm

Fig. 4: Production of metal ceramics
melt-solidification using UO2 in the NNC,
Kazakhstan.
Large-scale melt solidification test for UO2+Zr+B4C,
which simulated fuel debris components in the
Fukushima Daiichi NPS, was conducted in NNC
Kazakhstan. Rock-shaped melt solidification, mixing
with metal and ceramics, and powdery/granularsolidification were firstly produced. In addition, theobtained physical property data will be utilized for thedevelopment of device and the criticality control forfuel debris retrieval.

Future Developments
We continue our international cooperative efforts with CEA to evaluate the characteristics of MCCI products and expand thecharacteristics data on MCCI products for which we have little information. Hydrous/dry properties of fuel debris required for thedevelopment of technologies for the collection, transfer and storage are further evaluated by tests and investigation, and property
data of fuel debris related to storage canister are continuously compiled until the end of FY 2016. Furthermore, analysis elements
technology for fuel debris will be developed based on the development plan formulated in FY 2015.
29

Main R&D Achievements
Key Challenge 3 – R&D for Treatment and Disposal of Radioactive Waste

R&D for Treatment and Disposal of Radioactive Waste

Existing ALPS: Coprecipitated iron
Existing ALPS: Carbonate

As fuel debris will be retrieved at the Fukushima Daiichi
NPS, a large amount of radioactive wastes are expected togenerate in the future, in addition to debris and fallen treesthat had already been stored. In this situation, it is necessary
to continue characterizing the radioactive wastes in parallelwith the development of methods for storage, treatment anddisposal to be implemented.

Technology for safe disposal and treatment of accident-generated
solid wastes will be developed through a study on a series of storage
management strategies for main radioactive wastes (waste
management stream), analysis of wastes, characterization ofwastes through inventory assessment based on waste productanalysis, fundamental test of treatments, study of long-term storage
strategy for secondary wastes stemming from treatment ofcontaminated water, understanding of disposal concept and thestudy of disposal classification for accident-generated waste.

Activity concentration

Aims

Subsidized Projects

Background

Main R&D Achievements

Additional ALPS: Carbonate

Fig. 1: Analysis data of radioactive waste (slurry)
generated by ALPS operation

1

Main R&D Achievements In-house R&D Activities

Main Achievements and Approaches

Integration of R&D achievement(Suggestion of basic concepts on disposal/treatment)

After gathering the current knowledge and information for the waste management stream, change of the amount of materials and inventory for
waste management stream will be examined in accordance with the waste management stream and evaluation items at a turning point. In
addition, information management tool to support R&D was studied; and the relationship between the data and workflow about characterization,
disposal and treatment was organized. Furthermore, we participated in the investigation committee by a specialist group established by
OECD/NEA and cooperated to organize a report on the management of accident-generated wastes.

2

Characterization

Along with the formulation of a mid-and-long-term plan on waste analysis, debris, secondary waste stemming from treatment of contaminatedwater treatment and contaminated water were transported to an off-site facility, then radioactivity analysis was carried out. Results showed thatthe slurry generated from the Advanced Liquid Processing System (ALPS) mainly consists of 90Sr and also includes Pu (Fig. 1). Analysis model
for inventory assessment was improved to reduce uncertainty according to the analysis result. Waste products generated from the secondary
cesium adsorption apparatus were evaluated using the analysis data such as treated water.

4

Study on disposal of the waste

With the establishment of a provisional disposal classification based on the inventory data setting for safety assessment of the waste,safety assessment techniques (scenarios, models, parameters and analysis case) were reviewed (Fig. 4). As safety assessment for theprovisional disposal classification was being conducted, information for safety improvement was also examined.

Future Developments
Aside from the continuous analyzing of accident-generated waste characteristics, inventory data setting will also be reviewed.
In regards to the study of long-term storage strategies, processing equipment for the stabilization of slurry generated from ALPS will
be selected and studied for implementation on-site.
A plan to expand the technical information and data on fundamental tests for the processing study is being formulated, and treatment
and disposal technologies will be evaluated.
Candidate methods of applicable disposal concept and safety evaluation will be narrowed down for the study of waste disposal.
The above results for processing and disposal methods will be reflected to further promote waste management.

30

IRID Annual Research Report 2015

Fig. 3: Appearance of a full-scale absorber vessel to be used forvarious tests for absorbing, etc.

・Farmers

・Freshwater fishermen
・Seawater fishermen

Trench water disposal

Surface water
systemIntra-pit
disposal〜100m

Trench, pit and marginal depth disposal

Data Sheet

Study on treatment of radioactive wastes treatment and long-term storage methods

In order to narrow down the packaging technologies for radioactive waste, flow of technology assessment and assessment
requirements were studied and examined. In addition, applicability tests on cement solidification for secondary waste stemming fromthe treatment of contaminated water such as the slurry generated from ALPS and absorbent, geopolymer solidification, sintering, andcompaction (Fig. 2) were conducted while selection requirements for the treatment equipment were examined.
In regards to the long-term storage of cesium absorption vessels, a measurement test for the distribution of Cesium absorption using
full-scale absorption vessel of the secondary cesium absorption apparatus (Fig. 3) was conducted, and applicability for absorption
code of nuclides was verified based on the test results.

Saltwater tank

Others3

Absorber vessel
Fig. 2: Pressure filtering test device to examine the dehydration
treatment for the slurry from ALPS

Soil

Sedimentary rock・Nuclide which goes through the
bedrock or aquifer is discharged into
the ground surface.
・Exposure dose is assessed forfarmers and fishermen (freshwater/seawater).

Marginal
depth disposal
Geological disposal
・Nuclide which goes through the
bedrock via fault is discharged into
aquifer and the surface.
・Exposure dose is assessed forfarmers and fishermen (freshwater/seawater).

Fault

Crystalline rock

Geological disposal
300m 〜
1,000mFig. 4: Conceptual model for a transition of the groundwater

31

Main study results:
In-house R&D activities

Other – Data Sheet

Key Challenge 3: R&D for Treatment and Disposal of Radioactive Waste

Background

Upon conducting the decontamination work for high places on
the first floor of the southwest area in Unit 3 at the Fukushima
Daiichi NPS, dry ice blast decontamination equipment, which wasdeveloped in FY 2014-2015 for high places, will be mounted to thedecontamination work carts and support carts, used since FY 2013
for low places. Dependable capacity and decontamination
performance for the equipment will be confirmed as well.

№

Presented at/by

Date

Details
Reactor imaging through scattering method of cosmic ray muon

Radiation Science Society

Apr. 17, 2015

2Young Researchers Conference, Atomic Energy Society ofJapan

Apr. 17, 2015

Reaction products at interface between molten core andconcrete

3

Japanese-German Symposium on Technological and Educational
Resources for the Decommissioning of Nuclear Facilities

Apr. 21, 2015

R&D on Treatment and Disposal of Radioactive Waste resulting
from Accident at Fukushima Daiichi NPS

42015 International Congress on Advances in Nuclear
Power Plants (ICAPP 2015)

May 3, 2015

Improvement of Molten Core-Concrete Interaction Model of the Debris
Spreading Analysis Module in the SAMPSON Code5

Non-Destructive Inspection International Conference

May 19, 2015

Fukushima Daiichi Muon Imaging

6

23rd International Conference on Nuclear Engineering(ICONE)May 19, 2015

Fukushima Daiichi Muon Imaging

7

23rd International Conference on Nuclear Engineering(ICONE)May 19, 2015

Mechanical properties of fuel debris for defueling toward
decommissioning

8

23rd International Conference on Nuclear Engineering(ICONE)May 19, 2015

Study of Treatment Scenario for Fuel Debris Removed from
Fukushima Daiichi NPS

9

5th International Conference on the Chemistry and Physics of
the Transactinide Elements

May 25, 2015

Current Status and JAEA's Challenges for
Fukushima Dai-ichi Nucleaer Power Stations

10

EPRI D&D Workshop and International Low-level Waste
Conference and Exhibition

Jun. 14, 2015

Fukushima Inspection Manipulator (FIM)

11

Water Chemistry Division, Atomic Energy Society of Japan

Jun. 15, 2015

Environmental remediation at the Fukushima Daiichi NPS anddevelopment of technologies

12

National Symposium on Power and Energy Systems

Jun. 18, 2015

Approaches andtechnologies

Decontamination for high places (suction and dry ice blast decontamination)

13

Summer School in Radiation Detection and Measurements

Jun. 18, 2015

Issues of Radiation Measurement for Fukushima-Daiichi Restoration
Activities-Visualization and Non-Invasive Inspection

Based on the measurement results on the high radiation dose rate, the decontamination (suction/dry ice blast) was performed
in priority to high radiation rate area (β ray) in the southwest of the first floor in the reactor building at Unit 3.

14

Nuclear Fuel Division, Materials Science and Technology Division,
Water Chemistry Division, Atomic Energy Society of Japan

Jul. 9, 2015

Data acquisition for estimating property of fuel debris retrieval

15

Annual Conference, Japan Society of Maintenology

Jul. 14, 2015

Development of the decontaminating equipment for low places, Suction / blast decontaminationequipment, Dry ice blast decontamination equipment, High pressure water jet decontamination equipment

Future Developments

16

Annual Conference, Japan Society of Maintenology

Jul. 14, 2015

Investigation of the reactor through the Muon transmission
method at the Fukushima Daiichi NPS

We will evaluate and study the results of the decontamination work and consult with the Tokyo Electric Power (TEPCO) for
future on-site implementation.

17

Annual Conference, Japan Society of Maintenology

Jul. 14, 2015

Improvement for severe accident code MAAP and accident
analysis18

Annual Conference, Japan Society of Maintenology

Jul. 14, 2015

Overview of the research on fuel debris characterization

19

Back-end Summer Seminar, Atomic Energy Society ofJapan

Aug. 5, 2015

R&D for treatment and disposal of radioactive waste

20

17th International Conference on Environmental Degradation of Materials
in Nuclear Power Systems-Water Reactors

Aug. 9-13, 2015

Research results for corrosion evaluation technology contributing
evaluation for the integrity of the RPV/PCV structure

21

Japan-Korea Joint Summer School, Atomic Energy Societyof JapanAug.t 16, 2015

Overview of development of criticality control technology for fuelretrieval

22

imPACT Tough Robotics ChallengeAug. 23, 2015

Status of development of robotics technology for decommissioning

23

10th International Conference NUCLEAR AND RADIATIONPHYSICS 2015

Sep. 8, 2015

Characterization of fuel debris by large-scale simulated debris examination
for Fukushima daiichi nuclear Power stations

24

TIA-ACCELERATE Symposium

Sep. 9, 2015

Development of non-destructive detection system of large-scale
structure through cosmic-ray muons

252015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Evaluation of long-term structural integrity of fuel assemblies
removed from the spent fuel pool26

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Verification test of decontamination equipment for high places: Suction/blast decontamination equipment,
Dry ice blast decontamination equipment, High pressure water jet decontamination equipment27

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Research for identifying conditions inside of the reactor throughthe MAAP severe accident analysis code

28

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Impact assessment on seawater for identifying conditions insideof the reactor during severe accident

292015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Development of jet breakup phenomenon prediction method for thecomplicated structure of the BWR lower plenum

Main Achievements and Approaches
1

Combination test on dry ice blast decontamination equipment for high/low places

In addition to the maintenance of dry ice blast decontamination equipment for low places which had been stored at the
Fukushima Daiichi NPS, a combination test for high places will be conducted. Result confirmed that there will be no problem on
the decontamination of high places in the reactor building at Unit 3 (Photo 1).

2

Measurement of radiation dose rate for high places

Using a dry ice blast decontamination equipment for high places that is connected to support cart, in which a dry ice blast
decontamination equipment for low places was mounted, the radiation dose (β ray and γ ray) was measured in 19 areas in thesouthwest of the first floor in the reactor building at Unit 3 from December 23, 2015 to January 22, 2016 (Photo 2).

3

IRID Annual Research Report 2015

remotely

operated

Data Sheet

32

Photo 2: Scene of radiation dose rate measurement for high places (Thedecontamination equipment comprises two of the equipment for highplaces (left) and the equipment for low places (right).

for

OthersPhoto 1: Scene of a combination test for the decontamination equipment
for high places and low places

achievements

Main R&D Achievements In-house R&D Activities

1

Subsidized Projects

The decontamination work was firstly conducted for low
places on the first floor of the reactor building at the
Fukushima Daiichi NPS. On the other hand, according to theinvestigation of the contamination level in the reactor
building, the radiation dose in high places (duct, cable tray,
piping, etc.) is the highest – around 70%. Reduction ofradiation dose in high places is expected.

Aims

Main Research Results Announced/Published in 2015
Main R&D Achievements

Reliability Evaluation of Technologies
for Remotely Operated Decontamination Inside Reactor Buildings33

Other – Data Sheet

Other – Data Sheet

Presented at/by

No.

Details

No.

Presented at/by

Date

DetailsJan. 28, 2016

Development of robotic technology for decommissioning andon-site application

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Development of criticality control technology for fuel debris

60

312015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Investigation of prosperity for the secondary wastes generated from contaminatedwater treatment at the accident at the Fukushima Daiichi NPS

61

Briefing session on Fukushima, Japan Atomic Energy Agency
Sector of Fukushima Research and Development

Jan. 27, 2016

Fuel debris retrieval and the further approach – R&D policy for fueldebris retrieval and response to the policy – and others

32

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Study of analysis method for the secondary wastes generated
from multi-nuclide removal system

62

Society of Engineering and Research in Nuclear System

Feb. 16, 2016

Overview of investigation inside the Reactor Containment Vessel

33

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Development of inventory evaluation method for the wastes
generated at the accident at the Fukushima Daiichi NPS

63

National Meeting, The Institute of Electrical Engineers of Japan

Mar. 16-18, 2016

Nuclear technology and its trends for decommissioning of the Fukushima Daiichi NPS– Technology trends for measurement/analysis technology

34

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Study of conditioning technology for the secondary wastes
generated from contaminated water treatment

64

National Meeting, The Institute of Electrical Engineers of Japan

Mar. 16-18, 2016

Development of robotic technology for decommissioning andon-site application

35

2015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Structural integrity of the Reactor Pressure Vessel / PrimaryContainment Vessel

65

Annual Spring Meeting 2016, Atomic Energy Society of
Japan

Mar. 26-28, 2016

Improvement of robustness for the muon scattering method
based on EM algorithm362015 Fall Meeting, Atomic Energy Society of Japan

Sep. 9-11, 2015

Characterization of fuel debris

66

Annual Spring Meeting 2016, Atomic Energy Society of
Japan

Mar. 26-28, 2016

Study on long-term storage method for spent zeolite vessels(Reports no.10 & 11)

37

TopFuel 2015

Sep. 13-17, 2015

Evaluation of long-term integrity structural integrity of fuelassemblies removed from the spent fuel pool

67

Annual Spring Meeting 2016, Atomic Energy Society of
Japan

Mar. 26-28, 2016

Investigation of properties for secondary waste generated from contaminatedwater treatment at the Fukushima Daiichi NPS accident

38Open symposium, Japan Energy Association

Sep. 17, 2015

High pressure water jet decontamination equipment / Investigation of the torus
room in Unit 2 / Investigation inside of the Primary Containment Vessel in Unit 1

68

Annual Spring Meeting 2016, Atomic Energy Society of
Japan

Mar. 26-28, 2016

Hydrogen generation due to radiolysis of carbonate slurry
generated from the multi-nuclide removal system

39

Global2015

Sep. 20-24, 2015

Comparative study of Sr adsorbents for radioactive
contaminated water on severe accident

69

Annual Spring Meeting 2016, Atomic Energy Society of
Japan

Mar. 26-28, 2016

Development of the eroded concrete advection-diffusion modeldue to MCCI analysis

40

Safety seminar for light water reactor, Tohoku University

Sep. 28, 2015

Development of robotics technology for decommissioning and thestatus of application on site

41

Institute of Nuclear Materials Management

Sep. 29, 2015

Application of Muon Transmission method to reactor/
nuclear security

42

Takasaki Advanced Radiation Research Symposium

Oct. 8-9, 2015

Evaluation of hydrogen gas for inorganic solidification which is simulated
wastes generated from ALPS during gamma radiation

43

Society of Muon and Meson Science of Japan

Autumn 2015

Imaging measurementtransmission method

44

Academic meeting, Laser Society of Japan

Oct. 15, 2015

Application of laser for fuel debris retrieval at the Fukushima
Daiichi NPS

45

Workshop on Radiation Measurement

Oct. 17, 2015

Development of visualization technique inside the reactor using
cosmic-ray muon

46

American Physical Society

Oct. 29, 2015

Muon Reactor Imaging

47

Workshop on the environment and materials, Japan Societyof Corrosion Engineering

for

the

reactorthroughmuon

48

Annual Meeting of the International Network of Laboratories for
Nuclear Waste Characterization (LABONET)Nov. 10-13, 2015

Inventory evaluation based on the analysis data of contaminated water generated from thecontaminated water treatment system at the Fukushima Daiichi NPS

49

Workshop on dose reduction cases for decommissioning work at
the Fukushima Daiichi NPS

Nov. 10, 2015

Radiation dose reduction due to the decommissioning work at thereactor building at Unit 2, Fukushima Daiichi NPS

Data Sheet

Study results of corrosion evaluation technology contributing structural
integrity evaluation of the RPV/PCVOthersNov. 26-28, 2015

List of Joint Researches/Contract Researches
No.

Project Name

Category

Subject

Partner

PeriodContract research

Measurement of fuel debris in the reactor through muontransmission method at the Fukushima Daiichi NPSHigh Energy Accelerator Research
Organization (KEK)

Jun. 2014Dec. 2015

50

Japan Joint Automatic Control Conference

Nov. 14, 2015

R&D on technology for improving remotely operability of robots

1

Development of technology for thedetection of fuel debris in the reactor

51

International Robot Exhibition

Dec. 2-5, 2015

Decommissioning system for high places, “Super Giraffe”

2

Development of technology for
detection of fuel debris in the reactor

Contract research

Development of supporting algorithm for identifying
fuel debris through muon transmission method

Los Alamos National
Laboratory (LANL), U.S.A

Jul. 2014Nov. 2015

52

Radiation Measurement Forum in Fukushima

Dec. 7, 2015

Announcement of the investigation results at the southern area on the
1st floor in Unit 1, Fukushima Daiichi NPS

3

R&D on treatment/disposal ofsolid radioactive waste

Contract research

Research on estimation/assessment
technology of radioactive waste inventory

Central Research Institute ofElectric Power Industry

May 22, 2015Jan. 29, 2016

53

Human Resource Development Seminar in Kobe

Dec. 12, 2015

A robot working for decommissioning at the Fukushima Daiichi
NPS

4

R&D on treatment/disposal ofsolid radioactive waste

Contract research

Study on properties of criticality ofaccident-generated waste

University of California, Berkeley,
U.S.A.

Sep. 30, 2015Jan. 29, 2016

54

Radiation Measurement Forum in Fukushima

Dec. 15, 2015

Development of radiation measurement technologies by HITACHI that
support the restoration and revitalization of Fukushima

5

R&D on treatment/disposal ofsolid radioactive waste

Contract research

R&D on radioactive waste management methods
using gamma ray measurement

Tokyo Institute of Technology

Nov. 26, 2015Feb. 29, 201655

Power-generating technology, Japan Power Engineering andInspection Corporation

Dec. 15, 2015

Status of development of a robotic
decommissioning and on-site applicationfor

6

Characterization of fuel debris

Contract researchPhysical evaluation of melt-solidification fueldebris (2)Central Research Institute ofElectric Power Industry

Oct. 2015Feb. 201656

Basic Education Series Seminar, Nuclear Workshop, Tokyo
Institute of Technology

Dec. 22, 2015

Development of decommissioning equipment for high places, Dose measurement by gamma camera,
Development and investigation robots: water swimming robot / floor surface traveling robot / Shape change robot

7

Development of the remotely operated
decontamination equipment

Contract research

Research on camera calibration method for simulated overhead
image generation system installed in a remotely operated robot

University of Tokyo

Dec. 15, 2014Aug. 31, 2015

57

2015 International Chemical Congress of Pacific Basin
Societies (Pacifichem)

Dec. 2015

Basic study on the solidification of wastes generated from the contaminated water
treatment facility at the Fukushima Daiichi NPS

8Development of the remotely operated
decontamination equipment

Contract research

Research on technology on robots for
identifying three dimensions

University of Tsukuba

Dec. 15, 2014Aug. 31, 2015

58

Pioneering R&D Committee, Japan Society for the
Promotion of ScienceJan. 15, 2016

Development of robotics technology for decommissioning work andon-site application

9

Development of the remotely operated
decontamination equipment

Contract research

Research on improvement for remotely
operability of manipulator

Kobe University

Dec. 15, 2014Aug. 31, 2015

59

Session on Structural performance of the reactor building, Decommissioning Investigation
Committee, Atomic Energy Society of Japan

Jan. 15, 2016

Progress report of the integrity evaluation test / Analysis of 1
FRPV pedestal reinforced concrete

10

Development of the remotely operated
decontamination equipment

Contract research

Research on coordinate movement control
of multiple carts

Shibaura Institute of Technology

Jan. 20, 2016Mar. 9, 2016

IRID Annual Research Report 2015

Main R&D Achievements In-house R&D Activities

30

Nuclear Technologies Seminar, Japan Power Engineering andInspection Corporation

Subsidized Projects

34

Date

Main Research Results Announced/Published in 2015

Main R&D Achievements

Main Research Results Announced/Published in 2015

technology

35

Other – Data Sheet

Other – Data Sheet

Main R&D Installations / Equipment
Over 1 million Yen

Project Name

32

Development of technology for investigation inside the Primary Containment Vessel

Set of previously confirming equipment inside the pedestal

33

Development of technology for investigation inside the Primary Containment Vessel

Set of simulated mock-up structure in the PCV

34

Development of technology for investigation inside the Primary Containment Vessel

Set of elemental testing device for A3 investigation

Details

Development of remotely operated decontamination technology

Suction/blast decontamination equipment for high places2

Development of remotely operated decontamination technology

Suction/blast decontamination equipment for high places3

Development of remotely operated decontamination technology

Dry ice blast decontamination equipment for high places

35

Development of technology for investigation inside the Primary Containment Vessel

Set of relevant equipment for a hatch opening device

4

Development of remotely operated decontamination technology

Dry ice blast decontamination equipment for high places Testing device

36

Development of technology for detection of fuel debris in the reactor

Shielding materials of the measurement equipment for transmission method

5

Development of remotely operated decontamination technology

High-pressure water jet decontamination equipment

37

Development of technology for detection of fuel debris in the reactor

Small measurement equipment for transmission method

6

Development of remotely operated decontamination technology

High-pressure water jet decontamination equipment for high places Testing device

38

Development of technology for detection of fuel debris in the reactor

Small muon tracking device system for scattering method

7

Development of remotely operated decontamination technology

Decontamination equipment for upper floors (Work cart, Suction/Blast decontamination unit)

39

Development of technology for detection of fuel debris in the reactor

Muon tracking device for scattering method to be used at the Fukushima Daiichi NPS

8Development of remotely operated decontamination technology

Decontamination equipment for upper floors (Relay cart, Cable winder, Dry/Ice blast decontamination unit)

40

Fuel debris characterization

Large capacity of thermogravimetric balance and simultaneous thermal analysis equipment

9

Development of remotely operated decontamination technology

Decontamination equipment for upper floors (Transport cart, Support cart, High-pressure water jet decontamination unit)

41

Fuel debris characterization

Piezoelectric crystal four-component cutting dynamometer

10

Development of remotely operated decontamination technology

Decontamination equipment for upper floors Testing device

42

Fuel debris characterization

Elemental analysis system for SEM

11

Development of remotely operated decontamination technology

Cart position measuring software

43

Fuel debris characterization

Automatic hydraulic embedding machine

12

Development of remotely operated decontamination technology

Interference verification software

44

Fuel debris characterization

Inverted metallurgical microscope

13

Development of remotely operated decontamination technology

Crawler cart for cooperation control verification

45

Fuel debris characterization

Carbon coater

14

Development of remotely operated decontamination technology

Durability test device for cable hose

46

Fuel debris characterization

Vacuum displacement arc melting furnace

15

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

Heating/Feed water equipment

47

Fuel debris characterization

Fuel debris compression test device

16

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

Turbid water treatment equipment

48

Fuel debris characterization

Fuel debris sonic speed measuring device

17

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

Work floor

49

Fuel debris characterization

Metallographic image analysis device

18

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

Mock-up transfer rail

50

Fuel debris characterization

Dynamic Micro Hardness tester

19

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

Full-scale mock-up

51

Fuel debris characterization

Simultaneous thermal analysis system

20

Development of technology for investigation inside the Primary Containment Vessel

B1 investigation device

52

Fuel debris characterization

Gas piping valve heater

21

Development of technology for investigation inside the Primary Containment Vessel

Dispersion prevention equipment for B1 investigation device

53

Fuel debris characterization

Sample cutting machine

22

Development of technology for investigation inside the Primary Containment Vessel

Subsidiary equipment for B1 investigation device

54

Fuel debris characterization

Sample polisher

23

Development of technology for investigation inside the Primary Containment Vessel

B1 investigation device Simulated device for mock-up

55

Fuel debris characterization

Core collecting device

24

Development of technology for investigation inside the Primary Containment Vessel

Shielding block removal equipment

56

R&D on treatment and disposal of solid waste

Chamber for alpha nuclide analysis

25

Development of technology for investigation inside the Primary Containment Vessel

Fuel debris measurement equipment

57

R&D on treatment and disposal of solid waste

Digital spectrometer

26

Development of technology for investigation inside the Primary Containment Vessel

Fuel debris measurement equipment Equipment for an elemental test

58

R&D on treatment and disposal of solid waste

Gamma-efficiency calculation program

27

Development of technology for investigation inside the Primary Containment Vessel

A2 investigation equipment (Including chambers and guide pipes)

59

R&D on treatment and disposal of solid waste

Aerosol transition observing device

28

Development of technology for investigation inside the Primary Containment Vessel

Set of X-6 penetration hole boring device

29

Development of technology for investigation inside the Primary Containment Vessel

Set of previously confirming equipment inside the penetration

30

Development of technology for investigation inside the Primary Containment Vessel

Set of deposits removal equipment (Including chambers)

31

Development of technology for investigation inside the Primary Containment Vessel

Set of subsidiary equipment for A2 investigation

IRID Annual Research Report 2015

Testing device

Data Sheet

1

Details

Others

Project Name

Main R&D Achievements In-house R&D Activities№

Subsidized Projects

36

Over 1 million Yen№

Main R&D Achievements

Main R&D Installations / Equipment37