Annual Research Report
2016

Printed on recycled paper.

5F, 3 Toyokaiji Building, 2-23-1 Nishi-Shimbashi,Minato-ku, Tokyo 105-0003,Japan
Phone Number:+81-3-6435-8601

URL http://www.irid.or.jp/en

Published: July 2017

IRID Research and Development Project, FY 2016 (Overview)

More than six years have passed 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 for decommissioning.
In the Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO s Fukushima
Daiichi Nuclear Power Station Units 1-4, (hereinafter, Mid-and-Long Term Roadmap ) revisedin June 2015, the latest milestones are clariﬁed. Also, "Technical Strategic Plan for
Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power
Company Holdings, Inc." which provides technical evidence for the Mid-and-Long Term
Roadmap was updated by NDF in July 2016. As R&D themes and major technical issues to besolved are getting apparent, further achievements are being required.
The International Research Institute for Nuclear Decommissioning (IRID) has been addressing
various themes in the following three R&D categories as the most urgent issues toward thedecommissioning of the Fukushima Daiichi NPS since its start in August 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

As a result, robotic technologies have been developed; remotely operated robots for thedecontamination of the reactor building and robots for the investigation inside the PrimaryContainment Vessel (PCV) have been developed and veriﬁed the feasibility of the technologies
on-site, which brought us eﬀective information inside the PCV. Furthermore, with the Muon
tomography using permeation method, we estimated the conditions of debris inside the Reactor
Pressure Vessel (RPV), and have achieved success in the development of technologies, which isessential for fuel debris retrieval.
This annual report is intended to introduce achievements of R&D which IRID had taken upon
as a challenge since FY 2016. We hope this report will help you understand the results of our
R&D.
IRID will continue to take responsibility for the R&D on the steady and eﬃcient
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.
March 2017

2. R&D for Fuel Debris Retrieval
Technology for Fuel Debris Retrieval

Technologies for Decontamination/ Radiation Dose Reduction〈Ensuring of Stability〉〈Ensuring of Work Environment〉

Improvement of
Suction/Blast
Decontamination
Equipment forHigh Places

Development ofSeismic-resistanceand Impact
Evaluation Method
for RPV / PCV

Development ofCorrosion Control
Technology for
RPV / PCV

Improvement of
Injection Functionof Dry Ice Blast
Decontamination
Equipment

Development ofTechnology for
Criticality Control
in Fuel Debris
Retrieval

R&D forTreatment andDisposal of Solid
Radioactive
Waste〈Fuel Debris Retrieval〉

Technology for Investigation /Analysis inside the Reactor〈Direct Research〉

Development ofTechnology for
Investigation inside PCVDevelopment ofTechnology for Investigation inside RPV〈Indirect Research and Analysis〉

Measurement and
Evaluation ofDistribution of Fuel
Debris in theReactor at the
Fukushima Daiichi
NPS Unit 2

Upgrading for
Identifying
Comprehensive
Conditions inside
the Reactor

Fuel Debris
Characterization

Repair Technology
for the Inside of PCVDevelopment of
Repair Technology
for Leakage Points
inside PCV

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

Upgrading ofApproach and
System for Retrievalof Fuel Debris andInternal Structures
Development ofFundamental Technologies
for Retrieval of Fuel Debris
and Internal Structures
Development ofTechnology forCollection, Transfer
and Storage of Fuel
DebrisMain R&D Achievements In-house R&D Activities

In order to promote the R&D activities, we committed to gathering expertise from around theworld under an integrated management which encouraged mutual coordination among projects,
overlooking the whole R&D projects, and conducted fourteen subsidized projects and three
in-house R&D projects in FY 2016.

3. R&D for
Treatment andDisposal ofRadioactive
Waste

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

Main R&D Achievements Subsidized Projects

Message from the President

1. R&D for Fuel Removal from Spent Fuel Pool

〈Legend〉： Subsidized
project： In-house R&D
activity

Contents
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 Fuel Debris Retrieval

Others Data Sheet

Development of Technology for Investigation inside the Primary Containment Vessel・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 4Development of Technology for Investigation inside the Reactor Pressure Vessel ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 6
Upgrading for Identifying Comprehensive Conditions inside the Reactor ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 8Fuel Debris Characterization ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 10
Development of Repair Technology for Leakage Points inside the Primary Containment Vessel ・・・・・・・・・・・・・・・・・・・・・・・ 12
Full-scale Test for Repair Technology for Leakage Points inside the Primary Containment Vessel ・・・・・・・・・・・・・・・・・・・・・・ 14
Development of Corrosion Control Technology for RPV / PCV ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 16
Development of Seismic-resistance and Impact Evaluation Method for RPV / PCV・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 18
Development of Technology for Criticality Control in Fuel Debris Retrieval ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 20
Upgrading Approach and System for Retrieval of Fuel Debris and Internal Structures ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 22
Development of Fundamental Technologies for Retrieval of Fuel Debris and Internal Structures ・・・・・・・・・・・・・・・・・・・・・・・ 24
Development of Technology for Collection, Transfer and Storage of Fuel Debris ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 26◆Key Challenge 3−R&D for Treatment and Disposal of Radioactive Waste

R&D for Treatment and Disposal of Solid Radioactive Waste ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 28

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

Hirofumi Kenda
President, International Research Institute for Nuclear Decommissioning

Measurement and Evaluation of Distribution of Fuel Debris inside the Reactor at the Fukushima Daiichi NPS Unit 2・・・・・・・・・・・・・ 30
Improvement of Suction/Blast Decontamination Equipment for High Places ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 31
Verification of Effect of Improved Injection Function of Dry Ice Blast Decontamination Equipment ・・・・・・・・・・・・・・・・・・・・・・ 31

Others − Data Sheet
Main Research Results in FY2016・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 32
Joint Research / Contract Research ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 35
Main R&D Installations / Equipment ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 36
1Main R&D Achievements
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
Aims

Fuel assemblies in the spent fuel pools at Units 1-4 at the
Fukushima Daiichi Nuclear Power Station (NPS) have been
stored in a water quality environment different from normal
conditions due to the injection of seawater and rubble falling
into the pool. To achieve long-term storage of fuel assemblies
in a common pool and dry storage system in future, it isnecessary to accurately evaluate the effects of these
conditions on fuel assembly component materials andoptimal storage environmental conditions.

In order to evaluate whether fuel assemblies retrieved fromthe spent fuel pools at Units 1-4 at the Fukushima Daiichi
NPS can be stored safely in the common pool for a long timeor not, we perform corrosion tests and investigate actual fuelin an environment that simulates real storage conditions. We
also conduct a simulated test to evaluate the impact on thefuel integrity during dry storage.

Unirradiated material (SUS304L),25℃
γ ray 500Gy/h Irradiated material (SUS304L),25℃
γ ray 5000Gy/h Irradiated material (SUS304L),25℃
Irradiated lock nut in actual use (SUS304L),50℃

Locknut at Unit 4

Mg

AlMain Achievements and Approaches
1 Technical development for the evaluation of the long-term integrity of theSiCl

No crevice
corrosion occurs.

Corrosion crevice re-passivation potential(mVvs, Standard hydrogen electrode)

Background

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Corrosion protection, data of ‘95 (SUS304),25℃

Measurement value of regular corrosion potential(Irradiated lock nut in actual use)

Ca

Chloride ion concentration (ppm)

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

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

fuel assemblies① Evaluation on the deposits on the surface of the fuel assemblies
Materials of fuel assemblies (lock nuts) of Units 4 at the Fukushima Daiichi NPS stored in the common pool were transported tothe post irradiation test facility, and component analysis for white deposits and measurement of corroded crevice re-passivation
potential were carried out. Among the constituents of the white deposit, the amount of Mg was the largest, and the amounts of Al
and Si were about half of it. The amount of Cl was below the detection limit. Since Mg (OH)2 was separated and Cl was not
captured, it is considered that there is no possibility of corrosion (Figure 1). In the electrochemical test, there was no crevice
corrosion sensitivity in the area where the chloride ion concentration was lower than 100 ppm, and it was confirmed that there wasalmost no possibility of corrosion in the common pool (Figure 2).

Figure 1: Component analysis result of white deposition area oflocknut at Unit 4

Figure 2: Re-passivation potential measurement result on corroded
crevice of lock nut

Some metal components in seawater are detected in the white
deposits, but Cl is not detected. This is considered to be a metaloxide rather than a chloride, and there is no possibility of corrosion.

When the chloride ion concentration is 100 ppm or lower, there is no
possibility of corrosion because the potential where crevice
corrosion occurs is in the region that is not corroded due to formationof oxide film (equal or lower than the corrosion crevice
re-passivation potential).

② Evaluation of integrity of fuel in dry storage conditions
Assuming dry storage of the fuel assemblies from the spent fuel pool of Units 1-4 at the Fukushima Daiichi NPS, in order to confirm
integrity of the fuel assemblies affected by rubble falling and seawater components, which were to be stored in the dry system, ahydride precipitation behavior verification test and a creep test were carried out, and the impact of the factors unique to Units 1-4
at the Fukushima Daiichi NPS on material properties was evaluated. It was confirmed that the impact on hydride precipitation
behavior and creep behavior was small even in the condition affected by both rubble damages and seawater injection (Figure 3
and 4).

Fuel cladding
Enlarged view

Damaged partA test to evaluate seawater components transfer behavior in the crevice structure of fuel components was carried out, and it was
found that the seawater components did not concentrate in the crevice structure and it changed according to the salt concentration
outside the crevice (completed in FY2015).

Future Developments

360℃, 170MPa, irradiated fuel cladding

0.50

0.40

0.30

0.20

0.10

Others Data Sheet

Others Data Sheet

From the viewpoint of long-term integrity evaluation of fuel assemblies, the fuel assemblies of the spent fuel pool of Units 1-4 at
the Fukushima Daiichi NPS were stored in the irregular water quality environment affected by seawater injection and rubble
contamination, the impact of such special background on wet storage and dry storage was evaluated, and it was confirmed thatthe impact was small.

HydrideStrain of circumferential direction (%)

2 Basic tests for long-term structural integrity

Main test (long-term integrity at 1F, 2015-2016),
irradiation 9×9, comparison
Main test (long-term integrity at 1F, 2015-2016),
irradiation 9×9, scratches + seawater
Main test (long-term integrity at 1F, 2015-2016),
irradiation of high burn-up 8×8, comparison
Main test (long-term integrity at 1F, 2015-2016),
irradiation of high burn-up 8×8, scratches + seawater
JNESH15, irradiation of high burn-up 8 x 8, No.64

0.00
0When a new issue is extracted at the time of spent fuel removal in the future, we will have to consider the necessity of new integrity
evaluation.

2

2000

4000

6000

Test hrs. (h)Figure 3: Hydride precipitation behavior verification test result

Figure 4: Creep speed test result(Irradiated test piece, 300℃, cooling velocity 0.04℃/h, stress in acircumferential direction 70MPa, attached scratches and seawater,
fixed rubble)(Irradiated test piece, 360℃, stress in a circumferential direction
170MPa, attached scratches and seawater injection, 5000 h)

There was no clear difference in the hydride precipitation state of thedamaged part even under the combined three conditions such as
fixed rubble, seawater injection, and damage.

Long time creep tests of 5000 hours were conducted for the cases
with/without seawater injection, and with/without damage. There
was no large difference in their results.

3

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

At Units 1-3 at the Fukushima Daiichi NPS, after the reactor
core has melted, nuclear fuel supposedly exists with someparts of reactor internals as fuel debris inside Reactor
Pressure Vessel (RPV) and Primary Containment Vessel(PCV). It is considered that the fuel debris that dropped fromthe bottom of the RPV into the pedestal supporting the RPV
were then spread from the opening at the bottom of thepedestal to the outside of the pedestal and distributed at thebottom of the PCV, but the condition of fuel debris is not still
identified.

By accessing the inside of the PCV from X-100B penetration
at Unit 1 and from X-53 penetration at Units 2 and 3, visual
images, radiation dose, temperature, and other data of theinside of the PCV were collected. In addition to the severe
environment with high dose and high humidity, it has been
confirmed that visibility is restricted due to vapor andstagnant water in the dark. It is also possible that interference
objects have been generated from the accident.
Therefore, it is necessary to develop technologies that canovercome these challenges and enable investigation inside
the PCV.

Appearance of B2 investigation device(outside of pedestal at Unit 1)

Image of Unit 3 investigation device(inside of pedestal at Unit 3)

Figure 1

Figure 3

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Technology for Investigation inside thePrimary Containment Vessel

Main Achievements and Approaches① Technology to access outside of the pedestal at Unit 1 (B2 investigation technology)

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

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

1 Development of technologies to access inside / outside pedestal
X-100BWe developed a B2 investigation device to access the inside of the PCV from Unit 1 X-100B penetration and investigate the
distribution of the fuel debris on the basement floor outside the pedestal, and carried out preparations for on-site verification at Unit
1 (Appearance of device: Figure 1).

② Technology to access inside of the pedestal at Unit 2 (A2 investigation technology)
We developed a device that could remotely drill Unit 2 X-6 penetration and an A2 investigation device to access the inside of the
PCV from the drilled X-6 penetration to investigate the condition inside the pedestal. We completed the on-site verification of thedrilling device in December 2016, and the verification test related to the A2 investigation in February 2017 (Appearance of device:
Figure 2).

X-53③ Technology to access inside of the pedestal at Unit 3 (Unit 3 investigation technology)
We have completed test manufacturing of an underwater swimming device which accesses inside the PCV from Unit 3 X-53
penetration and investigates the condition inside the pedestal (Image of device: Figure 3).

X-6

2 Development of technology for investigation inside PCV in next phase
In addition to the technologies to access inside / outside the pedestal mentioned in the previous section 1 , from necessity ofinvestigation aiming at obtaining more detailed information, we have prepared a plan of technology for investigation inside thePCV in the next phase.
In planning, we organized and analyzed the latest needs found in each project to establish targets of investigations inside the PCV
for the development of the fuel debris retrieval technology, extracted technical development tasks, and prepared for investigations
and development inside the PCV.

Access route：Investigated part

Appearance of A2 investigation device(inside of pedestal at Unit 2)

Others Data Sheet

Others Data Sheet

Based on those plans, we examined the technology for establishing an access route to the inside of the PCV, access investigation
technology, and measurement technology, and conducted element tests to construct the concept.

Future Developments
We completed the B2 investigation at Unit 1 in FY2016 and continue the development for the investigation inside the pedestal
scheduled for Unit 3 in FY2017, aiming to complete the verification test.
Also, we will continue to develop technologies for investigations in the next phase to acquire further information about the insideof the PCV.

Figure 245

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

In order to retrieve the fuel debris and reactor internals,
information about their locations, shapes and conditions must
be understood in advance. However, since the inside of RPV
has a complicated structure with extremely high radiationlevel, it is difficult to acquire directly information about theinside of RPV.

In order to promote the future decommissioning including
retrieval of the fuel debris safely and smoothly, it is required
to collect information on conditions inside the reactor in which
are unclear now. Therefore, we clarify investigation targets,investigation methods and obtained results, and establish acomprehensive investigation method, to develop feasible
technologies.

Main Achievements and Approaches
1 Formulation and updating of plan for investigation/development

Top hole drilling investigation method

Primary Containment Vessel(PCV)

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Technology for Investigation inside theReactor Pressure Vessel

See Figure 2

Reactorbuilding(R/B)

Steam drier
Reactor Pressure Vessel(RPV)Side hole
drilling
investigation
methodSteam separator

See Figure 3
Upper grid plate

Reactorcore

2 Development of equipment to access the reactor core from the top

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

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

We investigated and organized relevant development projects and on-site needs, and extracted useful information for developinga debris retrieval method such as appearance and radiation dose at the reactor core and reactor bottom for investigation. Also,
regarding the access route for investigation of the reactor core, we selected not only the method allowing access from theoperating floor above the RPV but also the method allowing access by drilling from the side of the reactor building as an applicable
option, and evaluated feasibility of assessment of the building strength (Figure 1).

Shroud

① Development of drilling device to access through RPV head

We compared and evaluated several methods considered to be applicable at each work, concerning positioning of holes on thePrimary Containment Vessel (PCV) head and the RPV head, processing of workpieces (leaving or collecting), etc. and selected
applicable technologies and methods (Figure 2).

Figure 1: Concept of investigation method② Development of boundary function maintaining device / access device for work
We evaluated the exposure caused by dust scattering to air during processing of the reactor internals with a simple model. Also,
after cleaning of the PCV head surface, we conducted an element test to verify the method to seal the connecting part between
the PCV head and the guide pipe, with respect to maintenance of the boundary function to prevent contamination from spreading.
In addition, we compared and evaluated several methods considered to be applicable at each work, concerning procedures forexchanging resin packings used for the connecting parts, and selected applicable technologies and methods (Figure 2).

Gate valve③ Development of drilling equipment through upper grid plate

Shielding unit

We manufactured a prototype of small tool head of Abrasive Water Jet (AWJ), which was remotely operated for opening work in
the narrow part of reactor internals, and conducted an elemental test to confirm its drilling workability. Also, we evaluatedfragments generated by the air AWJ for a mechanism that allowed to descend while opening on a complex shaped structure, andconducted a conceptual design (Figure 3).

3 Development and selection of investigation method up to the reactor core

We have confirmed applicability to the environment (high dose and fog) with the elemental test for the technology to investigate
the internal situation and dose rate by passing through small diameter access routes. In addition, we conducted a conceptual
design of the mechanism that brought the investigation devices closer to the deeper part in the reactor core.

Tool box

Well cover
Guide pipe

Steam separator
simulated unit

PCV head
RPV insulator
Opening work of steam separator

Others Data Sheet

Since the inside of the reactor well, individual structures, and condition of the access route to the reactor core part, the dose rate
and others are unknown, we classified the work steps in confirming and establishing the access route and examined plans to be
implemented together with the investigation, in order to implement the safe and reliable method.

Future Developments
Concerning the top and side hole drilling investigation method, we will conduct a detailed design of the equipment related to theapplicable technologies and methods while ensuring safety, and verify the possibility of remote operation on site with a partial
mockup test. In addition, we will formulate a series of work steps from installing the devices on site, conducting investigations, toprocessing after the investigations.

6

Others Data Sheet

AWJ
small tool head

4 Design of total investigation device system and method planning

Guide pipe
simulated unit
Seal mechanism

PCV headsimulated unit

PCV head section seal verification test
Figure 2: Drilling through RPV head / boundary function toprevent contamination from spreading

Shroud head
simulated unit
Opening of shroud head
Figure 3: Verification test of hole drilling on reactor internals
with air AWJ

7

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

Background

Aims

Identifying the conditions inside the RPV andthe PCV is essential for planning the methodof fuel debris retrieval and developing safetymeasures. However, Units 1-3 at the
Fukushima Daiichi NPS have extremely highradiation dose, and it is difficult to investigate
or observe them directly.

This project is intended to steadily proceed with decommissioning of the
Fukushima Daiichi NPS to utilize results of the accident progression analysis
and other R&D projects, analysis results of the measured data such aspressure and temperature at the time of accident, and information obtained
on-site. We promote to estimate the conditions inside the RPV and the PCV
based on comprehensive analysis and estimation on these information with
the Institute of Applied Energy.

Main Achievements and Approaches
1 Comprehensive analysis and evaluation of condition inside the reactor① Comprehensive analysis and evaluation based on actual data and results of other projects② Establishment of database required for comprehensive analysis and evaluation
In order to effectively promote the activities shown in 1 ①, we established a database of totally collected and organized informationsuch as actual measurement data and results of the on-site investigation. In addition, as a computational evaluation function
contributing to comprehensive analysis and evaluation, we prepared the function to convert the FP mass calculated by the
analysis code into the dose rate at the actual measurement location.

2 Comprehensive analysis and evaluation on behavior and characteristics of fueldebris and FP① Uncertainty reduction by utilizing analysis method
Using the accident progression analysis code, we conducted sensitivity analysis and inverse analysis, etc., for the events
presumed to have occurred inside the reactor based on the boundary conditions and analytical models, and acquired knowledge
contributing to the comprehensive analysis and evaluation described in 1 ①. We also conducted a simulated fuel assemblies
plasma heating test (Figure 2) and obtained knowledge that could reduce the uncertainty of phenomena such as core damage,core melt, and core move in the BWR system.

② Evaluation of FP chemical properties
In evaluating chemical properties of the FP, we focused on Cs with a large contribution of dose during decommissioning, andinvestigated the distribution of Cs and its chemical properties such as identifying the chemical species to be considered in addition
to the standard chemical species such as CsI and CsOH, and the possibility of uneven distribution of poorly soluble Cs along withreaction with the structural materials.
We also started to analyze the samples taken at the site for identifying conditions inside the reactor (e.g. A curing sheet collectedon the operating floor at Unit 2 (Figure 3)).

③ Utilization of domestic and international knowledge through international joint research●Debris could have penetrated inside the
CRD due to CRGT damage (Generalestimation).

●Some fuel debris could have solidified
without
causing
MCCI(Generalestimation).
●Steam generation was confirmed during
investigation inside the PCV. Fuel debris
could have emerged from the water(Actual measurement).

●Since the temperature rise of the outer
fuel may not be high, the pellet may
remain the outer fuel debris (Generalestimation, test and analysis).

●When the heat transfer from fuel debris
is small, CRGT remains unmelted(General estimation).

●Particulate debris and pellet may
accumulate in the stagnant area
(General estimation, test and analysis).

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

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

In estimating the conditions inside the RPV and the PCV of each unit, we created an information summary map thatcomprehensively summarized various kinds of information in each section of the RPV, the PCV, and the reactor building. By
comprehensively analyzing and evaluating the information, we created estimate diagrams of the fuel debris distribution (Figure 1),
the FP (Fission Products) distribution, and the dose distribution.

●Muon measurement indicates a shadowof high-density substance, fuel debris, at
the RPV bottom. Fuel fallen the lower
plenum might remain at the RPV bottom(Actual measurement).

●Water level could be formed outside theshroud due to the PLR system pressure
rise during the FDW flow rate increase(Actual measurement).
●No damage of the shroud is estimateddue to temperature decrease by CS
water injection and water level riseoutside the shroud during the increase ofwater injection (Actual measurement).

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Upgrading for Identifying Comprehensive Conditions
inside the Reactor●Assuming the energy amount basedon the PCV pressure rise due tohydrogen generation, most of fuel isassumed to have melted (Actual
measurement/analysis).
●As the temperature drop was confirmed
during CS water injection, fuel isassumed to exist around the outerreactor core in which low flow CS water
was injected (Detailed positions of fueldebris cannot be estimated as it behaves
similar as a heat source, even whenmolten fuel drops, and solidifies on thefuel support and CRGT). (Actual
measurement)●The result of muon measurement
indicates the possible existence of fuelon the outer reactor core (Actual
measurement).
●Fuel rods are assumed to exist in a partof the outer reactor core (Generalestimation).
●Molten fuel is assumed to be generalsolidified oxide (General estimation).

●The outer CRD was confirmed during
investigation inside the PCV. The RPVholes are assumed to be in around thecenter of the RPV (Not large). (Actual
measurement)●Some debris fallen from the holes may
adhereon
the
CRD(Generalestimation).

Explanatory note●Particulate debris can be formed inaccumulated water on the PCV floor.
●Particulate debris could be accumulated
in the stagnation area (Generalestimation).

●Fuel debris caused MCCI is assumed tobe mixed with concrete (Generalestimation).
●MCCI is estimated to be limited due to notendency of the PCV damage (Noleakage from sand cushion drain pipes).
(Actual measurement)

Integral fuelBallooning fuel※Oxide debris (looseness)

Oxide debris※

Particulate debris

Heavy metal debris※

Concrete mixed debris

Particulate pellet※

Integral CRGT

Cladding residue※

Damaged CRGT

Molten reactor internals※

Integral CRD

Solidified B4C※

CRD (debris inside)

Control rod mixed melt※

Integral shroud
Pellet※Not used in the estimate diagram of Unit 2

Figure 1: Estimate diagram of fuel debris distribution (example of Unit 2)

As of March 31, 2017

Others Data Sheet

Others Data Sheet

We operated the international joint research project (OECD/NEA BSAF Phase 2) as a host and compared achievements
conduced by many organizations to identify the accident progress scenarios and the scope of uncertainty of analysis andevaluation, which were utilized for comprehensive analysis and evaluation shown in 1 ①.
We also conducted SAMPSON-MELCOR Crosswalk, and identified the common points and different points of both codes related tofuel temperature rise, melting progress, etc. As a result, we acquired knowledge to understand the accident progression analysis.

Future Developments
Following the activities in FY2016, we will promote comprehensive assessing of conditions inside the reactor, by conducting"improvement of the analysis code" as one of the approaches when needed, in addition to conducting "analysis and evaluation
based on the information acquired on site," "analysis and evaluation of the data collected at the time of and after the accident andinverse problem analysis," and "analysis and evaluation based on the analysis code."
We will effectively utilize information to be acquired in scheduled investigations, such as investigation inside the PCV at Units 1
and 2, and continue to identify comprehensive conditions inside the reactor.
8

Figure 2: Simulated fuel assemblies plasma heating test (preliminary test device,heating condition, and tested unit)

Figure 3: Curing sheet to be analyzed

9

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval
Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Fuel Debris Characterization
BackgroundAims

To ensure safe decommissioning process including retrieving,collecting, and storing fuel debris, it is necessary toinvestigate and estimate the fuel debris properties such as
hardness and drying characteristics that can be used fordevelopment of retrieval equipment, and also to analyze theactual debris after obtaining the fuel debris sample to identify.

In order to provide information on the fuel debris required forthe decommissioning works, we estimate properties of thefuel debris generated at the Fukushima Daiichi NPS withtests using simulated debris, and also develop elemental
technologies for fuel debris analysis, so that we can analyze
the actual fuel debris effectively for evaluating the fuel debris
properties.

1mm

50μm

1mm

①Status after heat treatment at 300°C for 15 hours
in a low oxygen atmosphere(Without crack generation)

50μm②Status after heat treatment at 600℃ for 11 hours
in a low oxygen atmosphere(Crack generation)

Figure 1: Appearance of simulated debris after heat treatment
With parameters such as temperature, time, and oxygen amount, we conducted a test using simulated debris. As a result, no crack was generated
even in the heat treatment for 15 hours at 300°C with the oxygen amount under the environment where the decompression with a vacuum pump
was assumed. Therefore, 300°C is one of rough targets in the drying temperature conditions.

Main Achievements and Approaches
1 Estimate of fuel debris properties
Based on the results (molten core concrete interaction (MCCI) product information) obtained in FY2016, we updated the "fueldebris characteristics list" (summary of FY2015) that had estimated the fuel debris properties.

Φ 25cmZrO2 cylinder

Air hole

Simulated fuel material, etc.
(UO2 + ZrO2 + Zr + Stainless steel)① Identifying fuel debris characteristics that affect collection and storage
We are developing technologies for drying characteristics, which are important for safety management, such as hydrogengeneration management during collection and storage of fuel debris. In order to investigate the drying characteristics of concretewith MCCI products, we continued the drying behavior test using cement (main component of concrete), and evaluated impact thatrehydratability of heat-deteriorated cement gave to the dry behavior in FY2016. Also, assuming powdered fuel debris that was
expected to be difficult to dehydrate, we evaluated the drying behavior of the powdered fuel debris using particle size and others
as parameters and acquired data on the powdering behavior of fuel debris to be used as the basic data for setting temperatureduring dehydration and heat treatment (Figure 1). In addition, we organized the existing knowledge on the dissolution behavior of
actinide elements, compiled and provided them to the collection, transportation and storage projects as fuel debris properties data
related to the storage canisters.

② Evaluation on unevenness of properties
In cooperation with the Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA), we used the MCCI test products
prepared under past rapid cooling conditions stored in CEA to acquire the generated compounds, hardness, and othercharacteristic data, and expanded the property data in the rapid cooling conditions. We also conducted a large-scale MCCI test,
considering the molten components and concrete components at Fukushima Daiichi NPS, and produced MCCI test products(Figure 2). As obtaining samples of different parts by disassembling the created test products, we obtained the knowledge thatwould be useful for retrieval, by confirming the erosion depth of the concrete and the state of the product, etc.

Porous area

High frequency
induction heating coilHigh-density area
H 50cmLayer containing
large amount of metals

Concrete test bodyΦ 50cm①Cross sectional view of test equipment

A large MCCI test product was prepared by melting and reacting about 50 kg of simulated fuel material with the concrete using the test equipment①. As shown in ②, the porous area, metal layer, and areas of the prepared test products were observed in the heterogeneous state. The test
product was disassembled, which provided samples of individual parts. Detail analysis of each item including hardness is scheduled in FY2017.

Si288.1578nm
1.5

10

1

Emission intensity

U 100ppm
Pu 100ppm

0.5

Others Data Sheet

Others Data Sheet

In order to analyze fuel debris, we selected the elemental technologies for fuel debris analysis to be developed considering
introduction of them into the Okuma Analytical Research Center, and are promoting technology development. For dissolution ofpoorly soluble fuel debris, we continued technology development using the alkali melting method, and evaluated dissolution
conditions using oxides and various metal compounds which composed fuel debris (Figure 3). In addition, in developing themulti-element simultaneous analytical technology with inductively coupled plasma atomic emission spectrophotometry (ICP-AES),
as conducting evaluation of spectral interference of uranium contained in large amount (Figure 4) and investigation of its solution
method, we promoted development of technologies such as a quantitative evaluation method for porosity by X-ray CT, and arationalization analysis method of multinuclear by ICP-MS, and identified the technical issues while verifying the feasibility of eachtechnology. In addition, to study transport the fuel debris samples, we conducted safety analysis for transportation of fuel debris
and obtained technical prospect of sample transport.

We will analyze and measure the hardness and others for each sample of the MCCI test product produced in FY2016, andincorporate the result in the "fuel debris characteristics list." We will also start to evaluate FP release behavior, which is necessaryfor examining drying equipment. Furthermore, as preparation for analyzing the fuel debris samples in the Ibaraki area, we continue
development of analytical technology and investigation of transportation, and will prepare a guideline on the analysis method.

②Status of MCCI test product after the test

Figure 2: Large MCCI test equipment and status of test products generated in the test in cooperation with CEA

3 Development of elemental technologies for fuel debris analysis

Future Developments

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

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

2 Characterization with the use of simulated debris

Si 1ppm
0288.11

288.12288.13288.14

1ppm

Figure 3: Alkali melting test in hot cell using
simulated debris (status of solution)
Dissolution test of uranium simulated debris (sample prepared bymelting UO2 and Zircaloy) was conducted with the dissolution
procedure using the alkali melting method. As a result of the test, it
was confirmed that no residue was generated in the dissolution
liquid and the total amount could be completely dissolved.

288.15288.16

U100ppm

288.17288.18288.19Pu 100ppm

288.2

Wavelength (nm)Figure 4: Interference of U and Pu in Si element analysis
Since fuel debris contains a large amount of uranium, spectralinterference may be caused by uranium when other elements areanalyzed. The figure shows the interference during analysis ofsilicon (Si) as an example of test results. The interference by
uranium was evaluated for each element, and the solution method
was examined.

11

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

Equipment for water stoppage

BackgroundAims

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 PrimaryContainment Vessel (PCV). In order to retrieve fuel debris,
we plan to submerge them in the PCV. For that purpose, it isnecessary to prevent water leakage from the PCV.

In order to substantiate fuel debris retrieval method bysubmersion or running water, this project intends to establish
technology for repairing water leakage points in the PCV.

⑤●

Water flow into S/COuter diameter: φ2057mm, plate thickness: 9mm
Number of pipes: 8, Material: Carbon sttel

Component
Vent pipewater stoppage

④
●
③
●

●
②
①
●

Filling in S/C forwater stoppage

Main Achievements and Approaches
1 Development of PCV bottom repair technology (Figure 1)

Damaged part
(assumed)① Water stoppage technology inside of the vent pipe facility

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Repair Technology for Leakage Points
inside the Primary Containment Vessel

Vent pipe

Figure 1: Overview of PCV bottom repair (example of Unit 2)

Application

Inflatable① sealing bag

Building a temporary dike forinjecting water stoppage
materials

Sub-inflatable② sealing bag

Sealing gaps after installation of
the inflatable bag

Water stoppage③ material

Constructing a submersion
range④ Repairing
material

Repairing cracks caused bylong-term deterioration andexternal factors

Remote repairing⑤ device

Remotely operating for drilling in
the S/C and vent pipes, removalof intervention, and installationof materials

Procedure of vent pipe stoppage: Starting construction
using remote repairing device(⑤) from the 1st floor of thereactor building in the order, ①,②,③ and ④.

② Water stoppage technology injecting filling inside of the suppression chamber (S/C)(Figure 2)●We confirmed that the water stoppage plug developed in FY2015 had no leakage in the water-stoppage test at 0.45MPa, butsome tasks such as improvement of installability of the water stoppage plug in the piping were extracted. In FY2016, we
improved the developed water stoppage plug, and have been discussing how to improve on-site applicability including
downsizing and simplification.

m. 10

Stiffening ring
Damage hole: Φ50mm

Quencher

Simulated strainer

Strainer

2 Boundary construction technology for connector pipes●For water stoppage of the PCV connector pipes, we are developing water stoppage materials to be filled in the piping andremote devices. In FY2016, we have started designing each device according to the water stoppage method, and been
conducting specific studies.

3 Development of water stoppage technology for torus room wall pipe

Stiffening ringrox

③ Water stoppage technology injecting filling into vacuum break line (Particular application to Unit 1)

Approx. 6 m

App●We completed the stiffening ring surmounting test, long distance pumping test, and elemental test under the assumed actual
environment. In addition, we conducted a 1/1 scale concrete water stoppage test simulating the inside of S/C, and succeededin stopping running water simulating the on-site environment.

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

Main R&D Achievements In-house R&D Activities●Concerning inflatable sealing bags, sub-inflatable sealing bags, and waterproofing materials continuously developed until
FY2015, we determined an action policy for the issues, planned a test schedule, and carried out each test.
●In addition to the conventional cement-based water stoppage materials, we tested rubber-based water stoppage materials in
FY2016.

Filling in S/C for water stoppage:
Drilling of the upper part of the S/C
and injecting concrete from thathole to stop water.
Injecting filling into the damage
hole (φ 50), quencher, and strainerof the S/C to prevent water fromflowing out of the S/C.

Simulated quencher
Damage hole: Φ50mm

Figure 2: Water stoppage test by filling in S/C (1/1 scale)

penetrating part●We selected a water stoppage material that could be used for spraying construction and tested it to confirm its water leakstoppage capability. Rubber materials were added and tested in FY2016.

ox.

pr
Ap

4m

Test condition

4 Development of PCV top repair technology

Upward, side, downward

Silicon G + Coating + Heating + Laser + Auto welding

Upward, side, downward

Silicon G + Coating + Heating + Blast + Laser + Auto welding Upward, side, downward

Water stoppage technology for leakage from seal section (equipment hatch) (Figure 3)
Approx. 4 m

Silicon G + Coating + Heating + Blast or Laser + Water stoppage material Upward, side, downward

Others Data Sheet

Others Data Sheet●In order to prevent leakage from the flange area of the equipment hatch and secure airtightness, it is necessary to seal theflange joint, and we are examining a remote welding method. In FY2016, we removed rust of the welded part, manufactured a
prototype of welding equipment and carried out elemental tests.

PostureSilicon G + Coating + Heating + Blast + Auto welding

5 Study of environmental improvement concept for actual application of

repair method●We extracted high priority sections for repairing works, in which in the PCV bottom and top an environment improvement model
was created as a representative. We also started to evaluate dose rates and to consider dose reduction.

Future Developments
It is necessary to examine applicability in actual use to reflect in the required performance of the devices, and also to study
maintenance of the long-term water stoppage function.
12

Simulated platform insideequipment hatch

Blast treatment

Figure 3: Equipment hatch test device (1/1 scale)

Laser treatment

Welding

Filling of water
stoppage material

The PCV equipment hatch on the first floor of the reactor building is usuallytightened by bolts, however they may be loosened after the accident. We
are developing a construction method to weld the gap for water stoppage.

13

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

Background

Aims

The Fukushima Daiichi NPS is in a harsh environment withhigh dose, narrow spaces, etc. There are many places where
it is extremely difficult for people to approach and work fordecommissioning. Therefore, in order to retrieve fuel debris,
it is necessary to develop equipment/device for remote
operations, which are implemented for decontamination andinvestigation or repair of each part, and to steadily promote
decommissioning of the Fukushima Daiichi NPS.

To determine the policy for fuel debris retrieval, we have to
conduct full-scale tests of the technology developed in the
R&D of "Repair Technology for Leakage Points in the PrimaryContainment Vessel (PCV)" and to confirm its feasibility
including its workability by remote operations. We also
prepare a manual considering actual construction, andconfirm its feasibility of the procedures and repair for waterleak blockage capability. In this project, the test will be
conducted in the Naraha Remote Technology Development
Center of Japan Atomic Energy Agency.

Main Achievements and Approaches
1 Full-scale test of repair technology for lower part of PCV

Workability verification test for reinforcement of
S/C support columns

Outline of technology for reinforcement ofsuppression chamber support columns

Major implementation of workability verification test

Since weight increase due to filling a water stoppage material inside
S/C is expected, technology aiming at seismic reinforcement for the
support columns of the S/C is developed.
Filling of
reinforcing
material

Unit 1

Filling of
reinforcing
material

Unit 2 and Unit 3

S/C

S/C

●Verification of workability of installation and collection of theplacing hose with the device on the work floor●Verification of construction procedures through water flow andremote monitoring performance
Work floor (The firstfloor of the reactor
building is assumed.)

Simulated interference
objectives(two-stage grating)

Placing device

Simulated suppression
chamber

PlacinghoseSimulated torus room
[Material to be selected] Underwater inseparable mortar

Figure 1: Overview of technology for reinforcement of
S/C support columns

Figure 2: Overview of workability verification test

Photo 1: Implementation of reinforcing workability verification test
for S/C support columns

Photo 2: Supply equipment of reinforcing material for
S/C support columns (outdoor)

Photo 3: Appearance of test device

Photo 4: Inside of test device (inside of S/C)

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

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

As one of the criteria for verifying the applicability, we examined an evaluation method based on risk assessment. Also, we areconducting a full-scale test for the following items:

Full-scale Test for Repair Technology for Leakage
Points inside the Primary Containment Vessel

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Full-scale Test for Repair Technology for Leakage Points
inside the Primary Containment Vessel① Suppression chamber (S/C) support columns (Figure 1)●We conducted a workability verification test as a full-scale test of S/C support columns reinforcement, and confirmed that there
was no problem in works under high dose and remotely controlled operation, and applicability of equipment prepared for the
PCV repair development. We will also discuss how to improve the workability based on the expertise obtained through the tests(Figure 2, Photo 1 and Photo 2).
●In order to determine necessity of reinforcement of S/C support columns according to the S/C support columns elasticity
analysis result implemented in the seismic resistance / impact assessment project as a hold point for implementation of theplacing test, we revised the implementation schedule of FY2017.

② Vent pipe water stoppage●We began a workability verification test for vent pipe water stoppage such as setting of the remote operation, and drilling in the
S/C and the vent pipes by the remote operation (scheduled to be continued until FY2017).

③ Filling in S/C for water stoppage (downcomer water stoppage)●As sharing the information of the various technical development projects, we are revising a test plan.

④ Preparation for test (Photo 3 and Photo 4)●As preparation for the tests, we implemented interference objectives to be used for test device and verified integrity bysubmerging the test device. Also we maintain operation and provide operation training for the equipment that supplies hot waterand drains water after testing to simulate the test environment, in order to establish a system with which we can operate thetests thoroughly.

2 Preparation of VR data for preliminary simulation test
Others Data Sheet

Others Data Sheet

We are establishing an environment where a remote manipulator used for vent pipe water stoppage is simulated on a VR (virtual
reality) system, allowing operation trainings. In FY2016, based on the interview of the designers and the operators of the remotedevices, we decided required items for accuracy improvement simulated actual operations on the console, and implemented
additional operation functions to the remote operation devices.

Future Developments
In order to judge the applicability of the actual equipment, we think that it is necessary to extract risks and tasks on the actual
equipment construction not only from the R&D personnel but also the workers in site, and utilize them in improvement of theworkability and construction process.
Also, we would like to establish an environment where information can be fed back to the remote operation devices in combination
with the remote operation devices and VR technology effectively.
In FY2017, we will continuously conduct a full-scale test of each method and verify the test results.
14

15

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

Due to severe events caused by the Great East Japan
Earthquake, Reactor Pressure Vessel (RPV) and PrimaryContainment Vessel (PCV) at the Fukushima Daiichi NPS
were exposed to high temperature and sea water, and fall offuel debris and other issues are expected. Corrosion control
technology is needed to prevent progress of corrosion of thestructural materials and to maintain integrity for a long period
until fuel debris retrieval from the reactor core.

Corrosion of the RPV / PCV is supposed to be controlled
comparatively more than the normal environment due todeaeration treatment of water injection and nitrogen injection
into the PCV, which have been carried out right after theaccident. Since the PCV may be exposed to the atmosphereat the time of fuel debris retrieval, we are promoting
development of corrosion control technology (rust inhibitor)
instead of nitrogen injection, and the evaluation of its
applicability.

●Flag type test piece(ESP measurement)

Washer(titanium)

Referenceelectrode

Electrode rod(titanium)

Electrode rod(titanium)

Gasket(PEEK)

Potentiostat(electrometer)

CoolerAir, Ar/N2 gas

Referenceelectrode holder

ThermocoupleBolt(titanium)

Separable flask
Test solution

Pt counter electrodeLuggin capillary

Bubbler

Test pieceNut(titanium)

Figure 2: Configuration example of electrochemicalmeasurement apparatus (irradiation)●By using an electrochemical measuring device of three electrodes(specimen pole (test piece), counter electrode, reference electrode), a
measurement was conducted.
●During self-potential (ESP) measurement, we used two electrodes
such as a specimen electrode and a reference electrode without settinga Pt counter electrode.

Main Achievements and Approaches
1 Evaluation of corrosion control effect and impact (Figure 1, 2, 3)

Data logger

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Corrosion Control Technology for RPV / PCV●Crevice corrosion test piece(ER,CREV, constant potential crevice
corrosion test)

Figure 1: Electrode composition of corrosion test piece① Evaluation of local corrosion resistance of rust inhibitor by electrochemical measurement
ReferenceelectrodeLow dose (0.2kGy/h)② Evaluation of fixation effect of phosphate rust inhibitor at high-temperature part

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

Main R&D Achievements In-house R&D Activities●We conducted electrochemical measurements for evaluating local corrosion resistance of carbon steel, the main structural
material, (corrosion crevice re-passivation potential measurement, self-potential measurement, constant potential crevice
corrosion test) under gamma irradiation and non-irradiation environment, and selected rust inhibitors that would not cause
localized corrosion.

Cooling pipe
Specimenelectrode●We conducted a batch test and a water flow test to evaluate the fixation effect of a phosphate rust inhibitor at thehigh-temperature part, and checked the temperature and other conditions of fixation. We also conducted a combined impact
assessment test with the use of the phosphate corrosion inhibitor and a sterilizing agent together to evaluate the impact ofcorrosion on carbon steel, and confirmed that there was no adverse effect.

Specimenelectrode③ Impact assessment on water treatment facility●We conducted a test to verify impact of the rust inhibitor on the water treatment facility, and confirmed that it would be necessary
to dilute collected water or remove the rust inhibitor from water in advance to reduce the impact on the water treatment facility,
and to make the concentration of the rust inhibitor in the water to be treated lower than the concentration at the time of injection
into the PCV.

High dose (4kGy/h)

Thermocouple
Gamma ray source

2 Conceptual design of corrosion control system (Figure 4)

Figure 3: Test condition under irradiation●We implemented conceptual design of a corrosion control system to apply the rust inhibitor (corrosion control measure) to actual
equipment. In addition, we formulated management guidelines for applying the rust inhibitor.

Future Developments●Development of corrosion control measures in the atmosphere based on the fuel debris retrieval method

Before start offuel debris
retrievalSP

In case that exposed parts to the air are required corrosion control by the fuel debris retrieval method, we will resume the study.

During fueldebris retrieval

Circulating
pump

RPV

CST

Monitoring facility

Others Data Sheet

Others Data Sheet

Cooler

CST reactor
injection pump

Corrosion inhibitor
supply device

SP
D/W

SP
To water treatment facility

Ground water

Corrosion inhibitor
removal device

Figure 4: Example of conceptual design of corrosion control system

16

The corrosion control system consists of a supply
system to supply rust inhibitor, a removal system toremove the rust inhibitor before transfer to thewater treatment facility, and a sampling / monitoring
facility to observe the water quality.

17

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

Due to severe events caused by the Great East Japan
Earthquake, Reactor Pressure Vessel (RPV) and PrimaryContainment Vessel (PCV) at the Fukushima Daiichi NPS
were exposed to high temperature and sea water. It isestimated that fuel debris fall and other issues occurred.
Some measures are necessary to maintain integrity of the
PCV / RPV structures for a long period until fuel debris
retrieval from the reactor core.

We clarify damage of major equipment in the RPV / PCV andthe spreading impact at the time of occurrence of alarge-scale earthquake, which affects the water level in the
PCV assumed during fuel debris retrieval and the installation
status of the critical equipment in the building, etc. We devise
a countermeasure that can prevent or control such impact
and confirm its effectiveness by assessment of seismic
resistance.

Main Achievements and Approaches
1 Formulation of safety scenario for large earthquake (Figure 1)

Clarification of safety
requirements (safety
review WG)

Selection of damaged parts(S/C support columns, RPV
pedestal)

Evaluation of spreading
impact caused by damaged
part (maximum assumption)

Countermeasure for spreading
impact [safety scenario(preventive measure, mitigation
measure)] (primary)

Extraction of damaged partand damage mode to be
critical on safety scenario

Extraction of damaged
parts (evaluation resultof integrity PJ in 2015)

Estimation of damaged
components to be damaged at thetime of accident (PJ of identifying
conditions inside the reactor)

Water level in PCV
during debris retrieval(PCV repair PJ)

System configuration
during debris retrieval(debris retrieval PJ)

Countermeasure for spreading
impact [consideration of the safety
scenario (preventive measure,mitigation measure)] (secondary)2 Development of seismic resistance / impact assessment method for safety scenario formulation
Extraction of evaluation items to
be required for safety scenario
formulation

Evaluation of detail required
for safety scenario
formulation

Implementation of analysis andtest required for upgrading ofsafety scenario：Implemented in this fiscal year

Regarding seismic resistance of the RPV pedestal, we will conduct a detailed assessment on the seismic resistance underdamage prediction conditions based on the latest knowledge.

We examined the high temperature corrosion test condition for estimating the strength deterioration of the reinforcing bar in theconcrete by the temperature history of the latest analysis, and fabricated a test body. In addition, we constructed a large-scaleequipment coupled analysis model, which RPV, and a stabilizer are integrated.

●Future assumption: After 15 years●Building damage model●Building damage model●D/W water level: OP.9000●D/W water level: OP.9600●Inside S/C: Concrete OP.2960●Inside S/C: Concrete OP.4400●Torus room: OP.−930●Torus room: OP.−2000●Inside vent pipe: Filled with water●Inside vent pipe: Repair to be●Vacuum break pipe: Repair●Strainer + Quencher water stoppage●Without reinforcement of S/Csupport columns

building up)●Seismic wave: Current design basis
seismic motion

Steel material 1% (for design in●Downcomer + Vacuum breaker water stoppage●Without reinforcement of S/C support columns

Unit 1

building up)●Seismic wave: Current design basis
seismic motion

Unit 2 and Unit 3

Figure 2: Example of seismic resistance assessment condition for S/C support columns●Modeling of assessment target(time history seismic response analysis)

Reinforcing ring

Torus body

Vent header●Calculation of seismic load(time history elastic seismic response analysis)

Vent pipe

In case where: Primary stress > Su
●Stress evaluation based on applied standard ①(supply condition: Ds or Su)

HeadersupportColumnsupportDowncomer●Modeling of assessment target(elastic-plastic analysis (double gradient method))

Header partof vent pipe

Downcomer
supportOthers Data Sheet

Others Data Sheet

In order to upgrade the evaluation method mentioned in the section 2 above, we verify the confirmation measures such as
analysis and tests based on seismic load and constraint conditions that can be supposed actually. As a method of evaluating andverifying resistance of the S/C support columns, we began studying a support column material test.

●Attenuation coefficient: Concrete 5%,

Steel material 1% (for design in

Bracing

3 Safety scenario upgrading

considered●Attenuation coefficient: Concrete 5%,

We created a seismic resistance assessment condition based on the vent pipes and the S/C water stoppage conditions which aremeasures in the safety scenario (Figure 2), formulated a time history seismic response analysis model (elastic analysis) and a limit
analytical model (elastic-plastic analysis) for the S/C support columns based on a total model where the both are coupled. We are
analyzing with these models now (Figure 3).
We studied the pedestal temperature history and distribution, and the influence of the range eroded by fuel debris, based on thelatest PCV accident progression analysis data. Considering these situations, FEM analysis model and analysis case draft were
formulated to start analysis.

●Future assumption: After 15 years

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

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

Figure 1: Total flow of study on seismic impact assessment PJ

As a primary idea of the safety scenario against the risk of rise in the water level of the torus room due to water leakage of thePCV, we indicated measures to reduce the load by the PCV inner water level restriction without reinforcing the S/C support
columns and the optimum water stoppage conditions of the S/C and the vent pipes.

formulating safety scenario

Extraction of analysis andtests required for upgradingof safety scenario

3 Upgrading of safety scenario

We selected the suppression chamber (S/C) support columns and the RPV pedestal as damaged parts which may have spreading
impact, based on the integrity evaluation and the analysis results of the PCV inside temperature during the accident and fueldebris erosion situation, and evaluated the damage and its spreading impact on the maximum assumption.

2 Development of seismic resistance / impact assessment method forMain R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Seismic-resistance and Impact
Assessment Method for RPV / PCV

1 Formulation of safety scenario for large scale earthquake

Restraint●Calculation of allowable collapse load
●Elastic-plastic analysis based on applied standard ②
In case where:
Seismic load > Allowable load

●Calculation of collapse load
●Elastic-plastic analysis based on applied standard ③

Future Developments
In order to formulate and upgrade the safety scenario, we will study the expanded material test data and the analysis method(model, elastoplastic analysis, etc.), and establish the seismic resistance / impact assessment method.

Implemented inthis fiscal year

IIn case where:
Seismic load > Collapse load

●Consideration based on time history
elastic-plastic seismic response analysis and test
(consideration of residual plastic displacement
amount)(implementation of material test)

Applied standard①JEAC 4601-2008: Guideline for seismic design of nuclear
power plants②JSME SNC1-2005: Rules on Design and Construction for
Nuclear Power Plants③JSME SNA1-2012: Maintenance standard

Figure 3: Development flow of seismic resistance impact assessment method, and total analysis model of S/C support columns

1819

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

Aims

It is estimated that fuel debris is not currently in a critical state. Since
the shape of fuel debris and water level may change during retrievalof fuel debris in the future, we secure the prevention of criticality andalso promote development of criticality monitoring technologies andcriticality prevention technologies so that even if criticality should
occur, it would be safely terminated.

In the FY2016 projects, we examine and evaluate the criticality
scenario based on the latest knowledge of condition inside thereactor for multiple retrieval methods to review criticality risks foreach unit. In order to confirm feasibility of the technologies, thecriticality monitoring technologies such as critical approach detection
technology, and the criticality prevention technologies using soluble
and non-soluble neutron absorbing materials was verified.

Unit 1

Unit 2

Unit 3MediumLow(Fuel may remain (Fuel may remainin the periphery) in the periphery)

Reactorcore● Submersion ofremaining fuelExtremely low(almost no
remaining fuel)Lowerpart of
RPV● Submersion ofdebris● Changes during
retrievalDuring submersion: During submersion: During submersion:
MediumLow
MediumDuring retrieval:During retrieval:During retrieval:
Low
Low
Extremely low(RV: Large and(RV: Large and(RV: Small)
exposed to the air) exposed to the air)

CRD
housing● Submersion of
Low - Extremely low
accumulated debris(Risk: low)

PCVbottomLow - Extremely low(Risk: low)

TargetLevel 2Control of abnormal operation andtermination of failures (PS system)

Monitoring of criticality approachto prevent criticality

Detection and controlof criticality

Criticality
controlParameter
monitoring

Prevention ofabnormaloperation● Restriction ofdebris retrieval● Criticality
approach

Low-Extremely low(Risk: low)

Specificmonitoring withamount● Application of
measures the system(Primary ● Monitoring water Boric acidsolution/level / boric acidissues)non-soluble
solution● Submersion ofSubmersion: Low Submersion: Medium Submersion: Low
exposed debris
Retrieval: Low
Retrieval: Small
Retrieval: Low● Change of state during(RV: Large,(RV: Large,(RV: Large,
retrieval(incl. curing)
exposure: small) exposure: large) exposure: small)
Residual volume=RV

concentration.

Table 1: Critical scenario and critical risk of each unit

We evaluated re-critical risks based on the latest prediction about the amount of fuel andfuel debris. The critical approach detection and criticality prevention technologies areapplied to the debris retrieval works.

Level 1Prevention of abnormal
operation (PS system)

Detection of
failures

Mitigation of
impact● Criticality
detection by
neutronassemblies or
FP gasconcentration● Termination ofcriticality byinjection of
boric acidsolution or
non-soluble
absorbent

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Development of Technology for Criticality Control in Fuel
Debris Retrieval
Background

Criticality risk

Sections

absorbent

Table 2: Idea of criticality control method based on defense in depth

To prevent criticality, and to detect to terminate it promptly even if it occurs, we summarize
ideas of safety measures that have multiple criticality prevention systems and impact
mitigation systems in the table above.

Main Achievements and Approaches
Category

1 Establishment of criticality evaluation methods

We reviewed the critical scenario of Units 1 to 3 based on the latest information on remaining fuel and fuel debris distribution acquired in the project
identifying conditions inside the reactor and the muon measurement of Unit 2 conducted in 2015, and re-evaluated critical risks (Table 1).

Liquid(viscousmaterial)

2 Development of critical approach monitoring method

To verify operability of the B-10 neutron detector under high radiation environment, we installed the detector near the spent fuel assemblies (within
the facilities of Nippon Nuclear Fuel Development Co., Ltd.) and prepared for verification tests for radiation dose and neutron detectability (Figure3). We confirmed that neutrons could be detected under proper shielding. In addition, we prepared a verification test the principle of each method
using a critical assembly (Kyoto University Critical Assembly: KUCA) that could actually simulate reactor core systems with different sub-criticality
levels (Figure 5).

3 Development of re-criticality detection technology

We are promoting development of advanced gas sampling system technology which can measure Kr-87 and Kr-88 in addition to conventionalXe-135 in the small amount of FP gas existing in Primary Containment Vessel as a method to detect re-criticality at an early stage. This technology
is based on gamma ray measurement with a germanium (Ge) detector.
For this reason, we installed an additional analyzer in the existing gas management system at Unit 1 equipped with a Ge detector, expanded themeasurement range of gamma ray energy up to 3 MeV, and started collecting gamma ray data (Figure 6). We have confirmed the peaks of Kr-87
and Kr-88 in the gamma-ray energy spectrum. In the future, we plan to analyze the data and use a result for studying delay in critical detection
time and for incorporating into the re-criticality detection model.

B4C / Metal sintered material◎○

Applicable to nuclear characterization

Glass material incl. B-Gd○○

Applicable to nuclear characterization (advanced)

Hollow boron

ーー

Withdrawal

Gd2O3 particle○○

Applicable to nuclear characterization (advanced)

Cement / Gd2O3 granulated powder material△△

Withdrawal

Water glass / Gd2O3 granulated powder material○△

Applicable to nuclear characterization (inorganic)

Underwater curing resin / Gd2O3 powder○○

Applicable to nuclear characterization (organic)

Underwater curing resin / B4C powder material○○

Applicable water stoppage material

Slurry / Gd2O3 granulated
powder material△○

Second option of nuclear
characterization organic system
of (Liquid to Solid) since it issolidified at 720 kGy or higher.

B4C gel material△ーLeaching characteristics of irradiatedmaterial must be improved.

At the time of accident
Nomal management

Gas discharge rate from PCV to environment 1/2-timeStandard caseGas discharge rate from PCV to environment 2-timeCase where the amount ofdebris is approx. 27 tonTime(s)Future Developments
Based on the latest knowledge about the condition inside the reactor, we will continuously refine the critical scenario and the critical risks, andexamine a criticality control method for each work process of the multiple construction methods to decide a fuel debris retrieval policy.
We will also carry out a feasibility verification test and a practical applicability test in order to indicate how to operate the actual equipment,
concerning element technologies such as the criticality monitoring technologies and the criticality prevention technologies.
20

Gas discharge rate from PCV
to environment
Critical detection delay due
to FP gas

Optical fiber cable（30m）

Crane
Retention / Water sealing jig(around 10 m)

Water injection stoppage delay
Boron transportation delay
Boron reactivity input rate
Gas discharge rate from PCV to operation floorGas discharge rate from PCV to outside of building
Building space volume
Operation floor space volume

Under
water

Spent fuel assemblyDistance L

Co-axial cable（30m）

Shield(with or without)
WeightFigure 3: Detector operation verification
test system under high radiation

Changed value of each input (multiple number)

We examined the equipment specifications with sensitivity analysis by changing major parameters (critical detection
delay time, gas leak rate, boric acid injection delay time, etc.) that would affect exposure evaluation.

Measuring system

The radiation dose is high in the vicinity of fuel debris. Therefore,
we conducted a test in a pool of facilities with spent fuel to confirm
whether the B-10 neutron detector could detect neutrons evenunder such circumstances.

Neutron multiplication factor（a）Prototype of measurement circuit

Criticality reference
0.95

Nitrogen

Δk:Maximum reactivity
addition amount at one
time of retrievalno!

k-meas(Measurement value)

Retrieval
ok

ok

k-alm
Alarm level
Operation stop

Measurement
delay

Rare gas, etc.
Ge（b）Neutron detector
PCV
Rare gas generation

Time

Figure 4: Image of monitoring critical approach

Others Data Sheet

Others Data Sheet

Furthermore, regarding the soluble neutron absorbing material sodium pentaborate (containing boron B) to be directly injected into cooling water,
the previous research results indicate that when boron alone is used to ensure sub-criticality, the required B concentration is about 6,000 ppm.
Since this value exceeds the actual concentration in the power generation reactors, we decided to prepare the nuclear characteristics verification
test with KUCA. We also confirmed feasibility of boron concentration maintenance equipment of sodium pentaborate water.

Figure 1: Image of exposure evaluation model

Figure 2: Evaluation example of exposure amount at the time of assumed
criticality occurrence at PCV submerging

4 Development of criticality prevention technology

We are promoting development of a non-soluble neutron absorbing material as applicable technology to prevent criticality during fuel retrieval.
Targeting the applicable materials extracted so far in the fundamental physical properties and radiation resistance performance test (B4C metal
sintered material, B-Gd glass material, cement/Gd granulated powder material, etc. (Table 3)), we prepared for nuclear characteristics verification
test with KUCA. Also, we are studying applied construction methods and equipment when implementing basic workability test to select applicablematerials.

PublicTo examine the specifications of the safety equipment, we develop a model that can
detect, terminate criticality and allows evaluation of the exposure amount of workers
and the public. It will be incorporated in the design so that exposure will not occur
even in case of criticality.

Water level increase rate

Control target
in normal

Workeroutside building
Ground shine(b) Example of sensitivity evaluation ofequipment parameters
At the time of
accident: 5mSv

Sky shine

Exposure caused
Partial submersion by discharge tomethodenvironment

Submerging method

We decided to conduct a nuclear characterization test on materials with stable characteristics inwater and air, among applicable non-soluble materials (solid, solidified material, viscous material).

(a) Example of sensitivity evaluation ofcritical detection delay time

Direct ray
Gas
management
systemOperation floor
workerTable 3: Evaluated applicable non-soluble neutron absorbing material

Public exposure dose (mSv)

We are developing a method to estimate sub-criticality and to monitor critical approach (Figure 4) with the reactor noise method, the neutron
source multiplication method, the period method, and the virtual neutron source multiplication method based on neutron measurements. We
prepared and tested to confirm feasibility of these methods on site.

In water

Public exposure dose (mSv)

Based on the idea of defense in depth, we summarized an idea of the criticality control method where each element technology was applied to theprotection system (PS) and the mitigation system (MS) for each fuel debris retrieval method (Table 2). Also, we summarized an idea of the method
to consider uncertainty of the critical calculation.

Liquidto Solid(solidified
material)

Exposure caused by leakFuture policy

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

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

We developed a criticality behavior evaluation model (Figure 1) with an additional exposure evaluation model, evaluated sensitivity of equipment
parameters related to criticality detection and critical shutdown (Figure 2), and examined equipment requirement specifications such as an FP gas
leakage rate, which would be effective for exposure impact mitigation.

Solid

Evaluation

Material

Figure 5: Preparation forfeasibility verification test

We constructed several reactor cores with different sub-criticality levels with KUCA. We prepared for a test withwhich we would measure neutrons with the B-10 neutron detector, and examine how accurately the sub-criticality
level could be predicted based on the obtained data.

Shield (lead)
Gas chamber

Figure 6: Image of gamma ray measurement test
for FP gas inside PCV
Gas in the containment at Unit 1 into the gas chamber was
introduced and measure gamma ray with a Ge detector to check
whether Kr-87 and Kr-88 can be be measured. Measurement iscurrently underway.

21

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

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. The reactor
buildings, the RPV and the PCV were damaged at the accident, andthe reactors have been in unstable conditions. We aim to stabilize
the reactors without diffusing radioactive materials, as retrieving fueldebris to maintain sub-criticality.

In order to retrieve fuel debris, we choose three retrieval methods;
Submersion-Top entry method, Partial submersion-Top entry
method, and Partial submersion-Side entry method. Toward
finalizing the policy for retrieval of fuel debris and reactor internals,
we clarify plant information, study methods, systems, and equipment
for retrieval of fuel debris and reactor internals. To confirm feasibilityof the retrieval methods and the entire system, we also formulate
development plans of the retrieval system and equipment.

：Confinement：Shielding

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

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

Shielding at the operation floor level to reduce shielding of
retrieval cell → Weight saving → Reduction of load on R/B
Fuel debris
retrieval cell
Reduction of direct radiationfrom fuel debris
Shielding at the operation floor level

Main Achievements and Approaches
1 Clarifying plant information to determine policy for retrieval of fuel debris

and internal structures

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

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

We organized plant data and development results of related projects, and clarified necessary information to determine the policy for retrieval offuel debris.

2 Study of methods, systems, and equipment for retrieval of fuel debris andinternal structures① Study of feasibility methods：Fuel debris retrieval route

Concerning the three representative methods; Submersion-Top entry method, Partial submersion-Top entry method, and Partial
submersion-Side entry method, we prepared a process flow showing a series of processes, and details of each work unit as a step diagram forfuel debris retrieval, and extracted technical issues (Figure 1).
In addition, we shared the information on latest situation of the other related projects, extracted common issues, and reviewed prerequisites.

② Conceptual study of system
In order to safely ensure fuel debris retrieval, we examined safety requirements during retrieving fuel debris, adopted the concept of defense indepth to organize functional requirements for the system, and considered the system configuration (Figure 2).

：Direct radiation from the inside of the reactor[Route A][Route B]■Shielding by the fuel debris retrieval cell on the operation floor (top floorof the reactor building)▶Simple flow line
▶Comparably easy correspondence in emergency▶To reduce load on the reactor building, the cell and equipment are
downsized.

■Shielding at the operation floor (top floor of the reactor building) level▶Reduction of load on the damaged reactor building is necessary.
▶Shielding at the operation floor level allows reduction of the shielding
function of the fuel debris retrieval cell. → Weight saving▶Shielding of direct radiation from fuel debris is allowed.
▶DSP (dryer separator pool) is utilized as a work space (for collecting inthe storage canister, etc.)

Furthermore, we assumed conditions such as a dispersion ratio at the time of fuel debris processing and a leakage rate from the boundary,carried out the exposure assessment, and examined required specifications of the main system.

Figure 1: Image of top-entry method③ Design of retrieval system
We studied a design of equipment for retrieval of fuel debris and reactor internals.

（C-3）
Discharge control ofradioactive materials

3 Formulation of development plans of systems and devices for retrieval offuel debris and internal structures

Cold
water

We implemented technical investigations concerning remote operation, cutting, shielding, prevention of dust scattering, etc. We also organized
the development plans formulated in the study of the methods, systems, and equipment for retrieval of fuel debris and reactor internals.

Future Developments
We will standardize the multiple ideas of fuel debris retrieval method, which are currently under consideration. We will confirm feasibility ofcommon technology required for all methods and individual technologies identified in consideration of standardization, and sharing withelemental tests, etc. For the safety system, we will confirm feasibility of the safety requirements and system function requirements examined inthis fiscal year. Also for fuel debris sampling, we will prepare a scenario and verify feasibility of the technology.
22（C-1）Negative pressure
maintained inprimary boundary

P
P

F

N

Condenser
Electric heater
Demister
PC

Cleaning tower

S
HEPA filter(remote maintenance required)
Rare gas monitor
Hydrogen concentration densitometer
Oxygen concentration densitometerTo building cell
Exhauster ventilation equipment(approx. 1000m3/h）

Cutting andcollecting device

1.5mH

X-6

(arm

P
T

P

P N

0.8mW

1.5mW
1.4mH

ion
penetrat

Arm

le

modu

1.4mH

6mL

3.2mW

PCV gas management
system (spare)

PC

odule

tion m

Opera

ule

r mod

ulato

Manip

Sampling equipment（C-8）
Monitoring of
released radiation

We additionally examined a safety system and an intermediate cell, considering reduction of working dose and various risks and benefits.

We examined improvement of accuracy of the measuring devices, countermeasures against debris with various shapes such as pebbles andsand, and also a concept of sampling equipment for fuel debris inside the RPV.

P

T

T

P

fuel debris

Based on the elemental test of the equipment to collect fuel debris after cutting, we examined the cutting conditions and actual structures (Figure 3).

P

Others Data Sheet

Others Data Sheet

4 Formulation of development plans of system and device for sampling of

PCV gasmanagement systemar)edevic t fiscal yescesspe ac t in the latya
Arm l to thequa
ucturestr

F

To building cell
ventilation equipment
ExhausterElectric heater HEPA filter(remote maintenance required) (approx. 1000m3/h）

Figure 2: Gas leakage prevention system configuration (example)
The system can keep negative pressure inside the PCV, and has ahigh performance particle filter in the exhaust line.

Figure 3: Example of sampling access device (arm type)The figure above shows an example of combination of a cutting andcollecting device and an access device based on sampling cutting
conditions.

23

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

BackgroundAims

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, the reactor buildings, the RPV and the PCV weredamaged at the accident, and the reactors have been inunstable conditions. Therefore, it is necessary to retrieve fueldebris to maintain sub-criticality, and stabilize the reactors
without diffusing radioactive materials.

We promote specific R&D for determination of fuel debris
retrieval policy (around the summer of 2017), establishmentof the retrieval policy for the initial unit (FY2018), and start of
retrieval at the initial unit (2021). In this project, targeting
three methods of fuel debris retrieval such as
Submersion-Top entry method, Partial submersion-Top entry
method and Partial submersion-Side entry method, we
conducted an elemental test to obtain the data information
required for evaluating feasibility of these methods.

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

DSP

Development of Fundamental Technologies for Retrievalof Fuel Debris and Internal Structures

Well
Work floor

SFP lid

Figure 1: Scale model test facility

Figure 2: Access device in RPV (device inside the reactor), full-scale modelA scale model test was performed to study on the prevention of spreading
contamination.

Sealing performance was verified by full-scale device
model.

Main Achievements and Approaches
Hydraulic manipulator

1 Comprehensive

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

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

adjustment of each elemental test and analysis ofelemental test result

In addition to IRID, another proposer who partially engages in the project joined in the R&D. IRID examined the achievements ofthe entire project, including the elemental tests planned and implemented by the partial proposer.

2 Elemental test required to determine feasibility of method
In FY2016, we conducted necessary confirmation through partial trial production and partial elemental tests according to eachelemental test plan formulated in FY2015.

① Prevention technology against the spread of contamination in removal of large structures●In order to verify prevention technology against the spread of contamination, we conducted a scale model test for each work
step (Figure 1).

Three-dimensional
measuring instrument
Figure 3: Hydraulic manipulator test
Pressure feedback was additionally installed to improve accuracy ofcontrol.

Figure 4: Access device in RPV test
The reaction force holding performance of the access device in the
RPV was verified.

② Prevention technology against the spread of contamination in retrieval of fuel debris inside RPV●We conducted a test on the RPV inner sealing and lower part sealing of the access device in the RPV under the Partial
submersion-Top entry method (Figure 2).

Robot arm

Access rail

③ Technology for access to fuel debris●A test on the hydraulic manipulator (Figure 3) was conducted.
●A test on the in-RPV access device was conducted under the Submersion-Top entry method (Figure 4).
●A test on the in-pedestal access device was conducted under the Partial submersion-Side entry method (Figure 5).

④ Remote operation technology for fuel debris retrieval●A test on the flexible structure arm for remote operation (Figure 6) was conducted.
●A test on the handling device for the fuel debris storage canister was conducted.

Figure 5: In-pedestal access device test
A test for access of the robot arm and the access rail was conducted to verify basic feasibility.

Figure 6: Prototype of flexible structurearm for remote operation
We manufactured a prototype of robot witha flexible structure arm and conducted a
test of operability inside the PCV.

⑤ Prevention technology against the spread of contamination in fuel debris retrieval⑥ Dose reduction technology for workers during fuel debris retrieval

Welding head

Others Data Sheet

Others Data Sheet●A test on the Submersion method platform and cell (Figure 7) was conducted.
●A test on the PCV welding device used for remote seal welding related to the cell of Partial submersion-Side entry method(Figure 8) was conducted.
Large shielding door

●Tests on shape following and lightweight shielding for the application of Top-entry methods were conducted.

⑦ Cutting and duct collection in fuel debris retrieval●A test on performance of the fuel debris cutting and collection technology was conducted.

Future Developments
Among the elemental tests conducted in FY2016, we will plan and implement not only the elemental tests to be carried out
continuously but also additional elemental tests to solve issues newly extracted in implementing the elemental test. We also plan
to implement elemental tests to solve issues extracted from the feasibility evaluation of the methods.
24

Figure 7: Test device for large shielding door of platform and cell
We manufactured a large shielding door inside the cell for theSubmersion method, and conducted a verification test on its
operability, etc.

Figure 8: Test for PCV welding device
A remotely-operate welding test was conducted to verify basic feasibility.

25

Main R&D Achievements
Key Challenge 2

R&D for Fuel Debris Retrieval

Development of Technology for Collection, Transfer andStorage of Fuel Debris
Background

Aims

According to the Mid-and-long-Term Roadmap towards theDecommissioning of Fukushima Daiichi Nuclear Power Units
1-4 (TEPCO), fuel debris retrieved from the Fukushima
Daiichi NPS is expected to be moved from the reactor
buildings and stored safely until the time when treatment anddisposal methods are determined. For this reason, a system
required for the collection, transfer, and storage of fuel debris
must be established.

Based on the experience of Three Mile Island Nuclear
Generating Station Unit 2 (TMI-2) in the US, and existing
technologies for transportation and storage of spent fuel, wedevelop a fuel debris storage canister (hereinafter called“storage canister”) and canister handling devices for safe andeffective collection, transfer, and storage. In FY2016, wesummarized storage canister design mainly for safety design.

Main Achievements and Approaches
1 Investigation and establishment of research plans for transfer and storage
In order to investigate sub-criticality control technology, hydrogen management technology, damaged fuel drying technology, andcharacteristics of products generated by reaction with molten core and concrete (MCCI product), which were considered to
5 below. We visited the Sellafield facility in the UK and the Argonne National
contribute to the discussion of sections 22 to 5
Laboratory in the US to obtain relevant information.

2 Study on fuel debris storage system
Following FY2015, we examined the storage concept of fuel debris that was considered in FY2014 based on the latest situationof the site, information from the related projects, and the knowledge obtained from this project. We have confirmed that it is notnecessary to review the storage system concept.

GravimetricdeviceMesh(Immersion of concrete
test sample)

Simulatedfuel debris

Test container

Unit canister

Unit
canisterHeight adjustment platform

Irradiation
sourceWater

Irradiation platform
Water tank(Conceptual diagram of irradiation test)

Figure 1: Elemental test for detailed investigation on hydrogenamount generated

Water tank

Figure 2: Drain test for mitigation of sub-criticality condition[Example of buffering structure]
Skirt type

Storagecanister

Body

Storagecanister

Bottomplate

Storage canister

Bottomplate

Body

Bottomplate

Buffering
structure

Buffering
structure

Buffering
structure
Collision surface

Collision surface

Load cell

Unlike the experience of TMI-2 fuel debris, the Fukushima Daiichi NPS requires countermeasures for safe collection, transfer, andstorage of fuel debris and MCCT products which are supposed to contain sea water components. In addition, for safe and effective
collection, transfer, and storage of fuel debris, expansion of the inner diameter of storage canisters and remote operation are
required.

Load cell(Before collision)(After collision)

Figure 3: Elemental test for buffering structure
We performed a verification test on the function to cushion the shock applied to the storage canister in the case the storage canister would be dropped.
We will reflect the result obtained into the buffering structure design and improvement of accuracy for simulated behavior analysis methods.

(Lid structure)

Storage canister
hanging tool andlid hanging tool(common use)

4 Development of technology for fuel debris collection

Mesh(side andbottom)5 Development of technology for transfer and storage of storage canister

Future Developments
For development of the technology, we will reflect the study results of fuel debris retrieval and collection to be conducted
especially in the projects for the fuel debris retrieval method and system upgrading, and the fundamental technology in the
future. Also, the final specifications of the storage canister and its device will be verified to optimize from both aspects of safety
and operability.
26

Others Data Sheet

Others Data Sheet

Based on the basic specification of storage canister designed in FY2015, we summarized an idea of storage canister for fueldebris retrieval mockup tests (Figure 4) based on the results of examination of the safety assessment method described in the
3 above and the requirements from the related projects (lid structure for remote operation, etc.)
section 3

We formulated basic specification drafts of a storage canister hanging tool for remote operation, a storage canister lid hanging tool
for remote operation and others, which would be basic equipment for the storage canister in the section 4 above (Figure 5).

(After test)

Increasing the inner diameter of the storage canister contributes to improved
effectiveness of fuel debris retrieval and storage. We set a water amount limit
value as a scenario to maintain sub-criticality even if the inner diameter isenlarged. We are conducting a test for water content reduction assessment bydraining a mesh structure using oxide simulating fuel debris.

In order to assume the amount of hydrogen generated by radiation decompositionof water remaining in fuel debris, we conducted a test to confirm the impact of theelution of concrete components into water. We will reflect this data to improve theaccuracy of prediction of hydrogen amount generated.

3 Development of safety evaluation methods

Continuously from FY2015, concerning evaluation on hydrogen amount generated (countermeasure against hydrogengeneration), aging degradation evaluation on materials (material selection), criticality evaluation (study on mitigation of
sub-criticality conditions with water amount restriction) to reflect these requirements in the storage canister design, and study on
the storage canister structure to satisfy these requirements (a buffer structure for ensuring structural integrity at the time of falling),
we conducted investigation of domestic and overseas case studies, trial analysis and tests, to summarize an assessment method
plan (Figures 1, 2, 3).

(Before test)

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

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

of damaged fuelUnit canister

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Hangingdevice(Buffering(Unit canister) structure)(Unit canister stored)

Figure 4: Idea of basic plan of storage canister
This figure shows a sample idea of storage canister form for mockup test. We
will optimize the form in cooperation with the fuel debris retrieval method andsystem PJ.

Tab

Tab

Figure 5: Example of basic design of storage canister hanging tooland lid hanging tool (common use)
To use the storage canister by remote operation, we formulated the basic
specification drafts of the storage canister hanging tool and the lid hanging
tool (common use). From now, we will optimize the easier handling form incooperation with PJ of the fuel debris retrieval method and system.

27

Main R&D Achievements

R&D for Treatment and Disposal of Radioactive Waste

Aims

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 for implementation.

We develop technology for safe disposal and treatment ofsolid wastes generated in the accident through study on aseries of storage management strategies for main radioactive
wastes, update of waste management stream, analysis ofwastes, characterization of wastes through inventory
assessment based on the waste analysis, fundamental testof treatments, study of long-term storage strategy forsecondary waste generated from contaminated water
treatment, identifying of disposal concept, and study ofdisposal classification for accident wastes.

Main Achievements and Approaches
1 Integration of R&D achievements

Sr activity concentration (Bq/g)

Background104

Rubble inlower part

102
100

101
137

103

3 Study on treatment of radioactive waste treatment and long-term storage

method105

Sintered solidification body107

Cs activity concentration (Bq/g)

Figure 2: Various solidified body of simulated waste

Figure 1: Ratio of 90Sr/137Cs in rubble
The ratio of 90Sr/137Cs in the rubble in Unit 1 R/B was consistent with thedata obtained so far (*data from FY2014 to early FY2016).

Fixing platform forhigh-pressurecleaning device

We conducted a solidification fundamental test of the secondary waste
generated from contaminated water treatment based on the existing
technology, and collected basic data such as compressive strength,
advance rate, hydrogen gas generation amount of the solidified sample.

High-pressurecleaning device
A test piece (HIC cut piece) applied with simulated slurry on the
inner wall of the HIC was installed to verify the impact on removalof the simulated slurry after cleaning.

SimulatedslurryCleaning
nozzleHIC
Cleaning
water

Before cleaning

We implemented a solidification fundamental test (Figure 2) based on the existing technology for secondary waste generated fromcontaminated water treatment without any experience in solidification treatment, and obtained data of availability of solidification,and integrity verification data of solidified material. We also compared the data with the requirements for technical evaluation andevaluated applicable technologies for each waste.

An accelerated test of remaining water evaporation behavior accompanying heat generation of the cesium absorption vessel andevaluation of salinity concentration behavior were conducted to propose measures and a verification method required forlong-term storage.

Future Developments
For characterization, we will feed back the result of study on treatment and disposal, update the analysis plan, and reflect it into
improvement of the inventory estimation model. For study on the long-term storage policy, we will investigate and examine theevaluation method of hydrogen generation and the handling method of generated hydrogen for safe and rational storage of thesecondary waste generated from contaminated water treatment. When examining treatment, we will investigate possibility oflong-term change in physical properties of solidified materials and formulate a test plan. When examining disposal, we will
investigate the overseas disposal concepts, evaluation methods, evidence for setting, and background in detail. With the results
above, we will integrate research progress and achievements comprehensively based on the waste management stream.
28

A three-layered cover system is installed at the
top of the waste layer.
(Furthermore, each cover system is configured
with multiple layers with water blockage,drainage, and other functions.)

EVAPOTRANSPIRATION
COVER SYSTEMBIOBARRIER
COVER SYSTEM

Compressed clay layer, rebar blown concrete
layer, and drainage layer are placed on thebottom and sides of the waste layer.
(At the bottom, a layer for withstand load isplaced.)PERFORMANCE COVER SYSTEM
Geocomposite
Drain

Reinforced Concrete
Barrier (Shotcrete)

WASTE PLACEMENT AREA

Sand Drainage
Layer

Compacted
Clay Liner

Not to Scale

Four-stage reinforced concrete canister(Slits are filled with sand.)Reinforced Concrete Barrier(Shotcrete)

Cover System
Shotcrete (high strength reinforced concrete)

Natural Red Bed Clay Soil

Others Data Sheet

Others Data Sheet

Based on the case studies on the overseas disposal concept (Figure 4), we organized information for formulating a disposal
concept and summarized features of the existing disposal concept. In addition, we improved a safety assessment method toevaluate the disposal classification considering uncertainty.

After cleaning

Figure 3: HIC container inner surface cleaning verification test

In order to stabilize the Advanced Liquid Processing System (ALPS) slurry, we studied a concept of device based on theoperational aspects for introduction of the slurry stabilizing device on site and the verification test results (Figure 3).

4 Study on disposal of waste

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

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

Rubble, secondary waste generated from contaminated water treatment, and contaminated water were transported to an off-sitefacility, and then radioactivity analysis was conducted. The rubble (concrete) of Unit 1 reactor building (R/B) showed results
consistent with the analysis data of the rubble obtained so far (Figure 1). We examined a method of collecting cesium adsorption
vessel zeolite, which was a sample of high dose, conducted conceptual design of a sampling device of the absorption vessel ofthe cesium adsorption apparatus, and proposed a sampling method. We also examined a method to reduce the uncertainty ofevaluation for analytical inventory estimation methods, found a set of inventory data with an improved method, and made it an
input for safety assessment of disposal.

Compressed solidification body

10-3 10-4

Concerning the waste management stream, we established a method to narrow down waste management streams with multiple
options and extracted issues to apply the method based on case studies. We also participated in a workshop for the report on the
management of accident waste by the organized by OECD/NEA expert group, and cooperated in publishing the report.

2 Characterization

Glass solidification body

Rubble inupper part

10-1 10-2

100

Cement solidification body

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

R&D for Treatment and Disposal of Solid Radioactive Waste

Rubble in Unit 1 R/B
Sludge in Unit 1 T/B
Rubble in Unit 1 R/B*
Rubble in Unit 2 R/B*
Rubble in Unit 3 R/B*
Sand in Unit 1 T/B*

106

90

Key Challenge 3

Geocomposite Drain
Compacted
Note 1Clay Liner

Clay Layer

This example shows the waste facilities (CWF) of the agreed states. In
addition to this, there is a compartment of the Federal Waste Facility(FWF), and it is required not only to comply with the state regulation
requirements but also to the Resource Conservation Recovery Act(RCRA), so the facility specifications and installation concept (a reinforced
concrete canister of wastes solidified to the container has seven stages)
are slightly different.
Note 2 The figure above shows an image created for the purpose of indication of
the facility functions, and actual scale is different. The actual scaled imageis shown below.
Source Application form of WCS

Figure 4: Case study example of overseas disposal concept [Conceptual diagram of the US Texas WCS (trench with liner) disposal facility]29

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

Improvement of suction and blast decontaminationequipment for high places

Background

Aims

Background and aims

In order to select an efficient fuel removal method, it is important to
identify fuel debris distribution in the reactor. For this reason, we
have developed technology to see through the inside of the nuclear
reactor using cosmic-ray muon with the cooperation of domestic andforeign organizations. We are currently promoting a plan to utilize
this technology for investigation of each unit.

Based on the muon transmission method technology developed by
the High Energy Accelerator Research Organization, we intend to
confirm the fuel debris distribution in the Reactor Pressure Vessel(RPV) using the improved muon transmission method for the
Fukushima Daiichi NPS.

Since the cylinder part of the suction and blast decontaminationequipment for high places developed on the supplementary budget
in FY2013 was damaged during the mock-up demonstration test, wechanged the design, and derived applicability of the technology. In
this research, we decided to conduct a verification test for
operability, which confirmed applicability of the technology.

Main Achievements and Approaches

Main Achievements and Approaches

1 Measurement with compact transmission method measuring device

1 Countermeasures and design change

In the measurements of Unit 1 conducted from FY2014 to FY2015, the site operations were so congested that we could not install the equipment
near the reactor building. Accordingly, we could not secure the elevation angle required to measure the lower part of the RPV. Therefore, wedeveloped a compact device of which ground contact area was about one fifth, allowing less interference with the on-site operations, and used
it in this project for the first time. In addition, in order to improve handling performance at the site, we adopted a surrounding system with leadplates at the site to shield environmental gamma rays, instead of the conventional iron box shielding body (Photo 1, 2). Since this system could
be installed even in places where large cranes could not approach, it contributed to completion of the measurements.

We installed a measuring device on the ground at the west side of the Unit 2 reactor building (Figure 1) and conducted measurements fromMarch to July, 2016. Then we evaluated distribution of the amount of substance based on the measurement results (Figure 2) and confirmed ashadow which seemed to be fuel debris at the bottom of RPV.

3 Summary①We eliminated a stress concentration point and changed theprevious material to the new one that supported stress generated
by excessive load (acceleration) (Extra super duralumin ⇒Carbon steel), and improved components.
②We installed an upper limiter for operation speed against
excessive load (acceleration), and restricted its use to the range
from 1.71 to 2.42 G.

2 Operation verification test result

We conducted an operation verification test with the mockup
equipment after the improvement, and confirmed that it satisfied thedevice requirement.

●We measured Unit 2 using the more compact muon transmission method measuring device and confirmed the shadows of the main structures(high density materials).
▶We confirmed the shadow of shielding concrete for the outer circumference of the containment.
▶We confirmed the shadow at the position of the spent fuel pool.
▶We confirmed the shadows of structures such as walls and floors of the reactor building.
●As a result of evaluating the obtained data, we have confirmed that high-density substances considered to be fuel debris exist at the bottomof the RPV.
●As a result of the evaluation by comparison with the simulation, it was estimated that there might be some high-density substances which were
supposedly fuel in the lower part of the core and the core outer circumference. However, uncertainty still remains in the evaluation due to the
influence of the structures of the reactor building.

We will install it according to needs at the site.

Future Developments

Since an event where the dry ice blast decontamination equipment
for high places developed on the supplementary budget in FY2013
terminated ejecting dry ice during application to Unit 3 at the
Fukushima Daiichi NPS, we confirmed reproducibility and discussed
countermeasures, and derived applicability of the function. In thisresearch, we decided to further identify a cause for countermeasures,and verified effectiveness of those countermeasures with an injection
test.

Using the muon transmission method, we could investigate presence or absence of the high-density substances in the reactor in two measurements; themeasurement at Unit 1 and Unit 2. In the future, we will consider application of this method to acquire information in a short period of time, according to
the site condition and investigation needs.

※ Size of one pixel: Equivalent to approx. 25 cm
at the cross section of the nuclear reactor

Background and aims

1 Identification of cause and countermeasures

Height OP (m)Measuring
deviceMuon

Verification of effect of improved injection function of dry
ice blast decontamination equipment

We identified the cause that dry ice supply system and the injection
route were clogged by intruded dry ice. We decided to confirm thesituation inside the dry ice storage and clean it before and afterusing the equipment, and also to blow the injection route by air.

Rotary blade

Others Data Sheet

Others Data Sheet(Size: Approx. 1 m × 1 m × Height 1.3 m)

Photo 1: Muon transmission methodmeasuring device (compact type)

Improvement of inclined shaft hydraulic cylinder fixing bracket(both ends of cylinder) and lift part rotating shaft

Main Achievements and Approaches

N
Photography
direction

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

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

2 Measurement results and distribution of amount of substance inside RPV

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Measurement and Evaluation of Distribution of Fuel Debris
inside the Reactor at the Fukushima Daiichi NPS Unit 2

Dry ice storage

Hopper
Rotor
Compressed air

Unit 2 reactor building(cross sectional drawing of the 1st floor)

Photo 2: Measuring device surroundedwith lead plates
30

Figure 1: Measuring device
installation position

North

Horizontal distance (m)SouthStructure of lower
part of RPV

Figure 2: Distribution of amount of substance at lower part of RPV(Measurement result: As of July 22, 2016)

2 Injection verification test result

After taking the countermeasures mentioned above, we conducted
an injection verification test and confirmed applicability of equipment
without clogging of dry ice to satisfy the device requirement.

To decontamination
head

Structural diagram of dry ice supply unit outline

We will introduce it according to needs at the site.
31

Others − Data Sheet

Main Research Results in 2016
Presented at/by

Date

1

"Report on Decommissioning Acceleration Research Program for the Center of WorldIntelligence Project for Nuclear S&T and Human Resource Development (Japan-UK
Collaboration Project), FY2015 "Apr. 1, 20162

The 1st International Forum on the Decommissioning of the
Fukushima Daiichi Nuclear Power Station

3

Lecture at Kyoto University "Nuclear power plant engineering"4

Details

Presented at/by

Date

Details

Present status in Fukushima - Robot and sensing technologies fordecommissioning of Fukushima Daiichi -31

Research Conference on Radiation Measurement for Decommissioning of the Fukushima Daiichi NPP

Aug. 4, 2016

Technology development for investigation of inside of PCV"Development light - section method"

Apr. 10-11, 2016

Overview of A2 investigation / B2 investigation (panel), display of static A2
investigation device, demonstration of B1 investigation device prototype

32

Research Conference on Radiation Measurements for Decommissioning of the
Fukushima Daiichi NPP (Radiation summer school for decommissioning)

Aug. 4, 2016

Overview of fuel debris measurement device (presentation)

Apr. 14, 2016

Current status of Fukushima Daiichi Nuclear Power Station andefforts for revitalization

33

IRID Symposium

Aug. 4, 2016

A2 investigation / B2 investigation (panel), Demonstration of A2, B1 investigation device prototypes
Debris retrieval method and fundamental technology development (panel), exhibition of muscle robot
prototypeSpecial Lecture on Reactor Decommissioning (Tokyo Instituteof Technology)

Apr. 22, 2016

Current status of fuel debris and retrieval technology

34

The 13th Nuclear Power Generation Technology Summer
SeminarAug. 5, 2016

Approach to domestic and international nuclear power (1)5International Congress on Advances in Nuclear Power Plants(ICAPP2017)

Apr. 24, 2016

Characterization of Carbonate Slurry generated from Multiple Radio-nuclides
Removal System in Fukushima Daiichi Nuclear Power Station

35

Seminar at Advanced Course Program of National Institute ofTechnology, Fukushima College

Aug. 5, 2016

Overview of molten reaction and introduction of investigation forfuel debris characterization research

6

Nuclear Materials and Energy (magazine)

Apr. 2016

Radioactive Contamination of Several Materials in the Accident ofFukushima Daiichi Nuclear Power Station36

2016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ Development of critical management technology for fuel debris- Critical management policy of debris retrieval, Critical behavior evaluation, Critical
approach monitor, Insoluble neutron absorber, and other ten subjects

7

Japan Society of Mechanical Engineers

May 1, 2016

Introduction of application of muon scattering method to nuclear
power

37

2016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ Development of remote decontamination technology in reactor building- Result of verification test of upper floor decontamination equipment (blast system, highpressure water system), and other two subjects

8

Journal of the Japanese Society for Non-Destructive
Inspection

May 1, 2016

Development of pinhole type gamma camera to visualize gamma
ray intensity distribution

38

2016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ PCV integrity- Resistance evaluation of reinforced concrete structure after severe accident, and otherthree subjects

9

Conference of "Conditioning and Geological Disposal of
Radwastes"

May 23, 2016

JAEA R&D activities on fuel debris characterization and related
research for 1F decommissioning, and other two subjects

392016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ Long-term integrity assessment of fuel assemblies removed from spent fuel pool
- Three issues such as Evaluation of fuel integrity during dry storage, Evaluation of seawatercomponent transfer by tracer, etc.

10Japan Society of Maintenology

May 25, 2016

Development of Remote Decontamination Technologies improving
Internal Environment of Reactor Building at Fukushima Daiichi NPS

40

2016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7, 2016

◆ Analysis and evaluation of assessing conditions inside reactor- Four subjects related to MAAP, six subjects related to SAMPSON, and five other subjects

11

Exchange of opinions with UK Sellafield Ltd.

May 25, 2016

Current status of Fuel Debris Retrieval Technology Development
for 1F, and other six subjects

412016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ Fuel debris characterization- Test and evaluation on behavior of product with simulated debris, and other eight subjects

12

Special Lecture on Reactor Decommissioning (Tokyo Instituteof Technology)

May 27, 2016

Remote technology

422016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ R&D on treatment and disposal of solid waste- Development of treatment and disposal technology for Fukushima Daiichi NPS accident
waste, Development of inventory evaluation methods, and other twelve subjects

13

Joint research group of Institute of Nuclear Materials Management / Nuclear
Nonproliferation Liaison Committee of Atomic Energy Society of Japan

Jun. 1, 2016

R&D and challenges on decommissioning and fuel debris at
TEPCO's Fukushima Daiichi Nuclear Power Station

43

2016 Fall Meeting, Atomic Energy Society of Japan

Sep. 7-9, 2016

◆ Development of repair and water leakage stoppage technology for leakage points inside the PCV- Development of water stoppage technology in S/C by filling underwater insoluble concrete

14

Central Research Institute of Electric Power Industry "Annual
Report 2015"

Jun. 1, 2016

Enhancement of technology base to clarify damaged and molten
fuel behavior during severe accident

44

The 34th Annual Conference of the Robotics Society of Japan

Sep. 7, 2016

Robot for investigating inside different diameter pipes

45

EUROCORR 2016

Sep. 11, 2016

Corrosion behavior of SUS316L in nitic acid solution containing seawater components-Effect of metal ions in high active liquid waste-

The 27th regular meeting of "Division of Water Chemistry,"
Atomic Energy Society of Japan

Jun. 3, 201616

Lecture at Nuclear Power Generation Division "Interchange
seminar between professional people and students"

Jun. 4, 2016

Approach to domestic and international nuclear power

46

Open Innovation Symposium, Kanagawa version

Sep. 16, 2016

Open Innovation Symposium, Technology interchange panel

17

Nondestructive Inspection Symposium, FY2016

Jun. 7, 2016

"Measurement and visualization technology by radiation corresponding to
nuclear accident" Development of pinhole type gamma camera forenvironmental radiation monitor

47

Advanced Course for Engineering of Reactor Decommissioning Measures

Sep. 17, 2016

Characterization and treatment of fuel debris

18

The 26th Joint Symposium of the Institution of Professional Engineers of Railway
Technical Research Institute / the Institution of Hitachi Professional EngineersJun. 20, 2016

Development of pinhole type gamma camera for environmental
radiation monitor

48

Advanced Course for Engineering of Reactor Decommissioning Measures (Tohoku University)

Sep. 17, 2016

Development of robot technology for decommissioning activities
and status of application on site

19

ICONE24 (The 24th International Conference on Nuclear
Engineering)Jun. 26, 2016

IN PLANT 3D POSITIONING SYSTEM USING POINT CLOUDDATA FOR REMOTE DECONTAMINATION MACHINE

49

ISTP (The 27th International Symposium on Transport
Phenomena)

Sep. 20, 2016

Coupled analysis on fluid structure for assisting piping closure20

WNE (World Nuclear Exhibition)

Jun. 28, 2016

Introduction of MEISTeR II and Super Giraffe50

7th International Scientific and Practical Conference

Sep. 21, 2016

CHARACTERIZATION OF FUEL DEBRIS BY LARGE-SCALE SIMULATED DEBRIS EXAMINATION
AT NNC KAZAKHSTAN FOR FUKUSHIMA DAIICHI NUCLEAR POWER STATIONS

21

The 29th "Summer Seminar of Nuclear Fuel Division," Atomic
Energy Society of Japan

Jul. 6, 2016

Challenges for fuel debris retrieval/ Chemical forms of seawater
salt, FP in MCCI products

51

Journal of RANDEC (Decommissioning technique) No.54

Sep. 30, 2016

Present status of solidification technology of radioactive waste
using alkali active material - geopolymer -22

Advanced Special Lecture at Department of Sustainable
Energy and Environmental Engineering, Osaka University

Jul. 8, 2016

Identifying the condition of molten fuel and criticality management
at Fukushima Daiichi Nuclear Power Station

52

13th International Conference on Probabilistic Safety Assessment and Management(PSAM13)

Oct. 2, 2016

Water Injection Influence for Accident Progression in Fukushima
Daiichi Unit 1

23

3rd Workshop of the OECD/NEA BSAF Project Phase 2

Jul. 8, 2016

Introduction of Japanese national projects on debris characteristics

53

HOTLAB 2016 Karlsruhe, Germany

Oct. 2-3, 2016

Approach to estimating fuel debris properties generated in
Fukushima Daiichi NPS

24"HOZENGAKU (Maintenology)" Japan Society of Maintenology

Jul. 10, 2016

Verification test of upper floor decontamination equipment

54

PSAM 13

Oct. 2-7, 2016

Water Injection Influence for Accident Progression in Fukushima
Dai-ichi Unit1

25

Journal of Nuclear Science and Technology

Jul. 12, 2016

Estimation of the Inventory of the Radioactive Wastes in Fukushima Daiichi
NPS with a Radionuclide Transport Model in the Contaminated Water

55

Fuel debris research committee, Atomic Energy Society ofJapan

Oct. 4, 2016

Estimation of fuel debris distribution by analysis and evaluation

26

The Social Infrastructure Maintenance Show

Jul. 20, 2016

Introduction of MEISTeR II and Super Giraffe

56

12th International Conference on NDE in Relation to Structural
Integrity for Nuclear and Pressurized Components

Oct. 6, 2016

Overview of Revitalization for Fukushima Nuclear Accident and Studyon Robotics and Inspection Technologies for Decommissioning

27

Nuclear and Radiation Subcommittee, the Institute ofProfessional Engineers, Japan

Jul. 22, 2016

Challenges of waste treatment and disposal for decommissioning of
TEPCO Fukushima Daiichi Nuclear Power Station, and R&D status

57

NUTHOS-11

Oct. 9-13, 2016

Analysis for the TEPCO Fukushima Daiichi unit 2 by the
SAMPSON code with core support plate model

28

The 13th academic lecture meeting of Japan Society ofMaintenology

Jul. 25-27, 2016

Technology development for fuel removal towards decommissioning ofFukushima NPS, accident waste treatment and disposal technology, andother six subjects

58

NUTHOS-11

Oct. 9-13, 2016

Validation of severe accident code SAMPSON debris cooling analysis module(DCA) against OLHF experiments and development of creep models

29

Civil engineering magazine "Civil engineering construction"

Jul. 2016

Development of technology for investigation inside PCV- Response to challenges in B1 investigation for Unit 1 and investigation results -59

Japan Society of Corrosion Engineering

Oct. 17, 2016

Impact assessment on mixed phosphate of zinc and sodium
carbonate affecting PCV materials in diluted artificial seawater (2)30

Analytical Chemistry

Jul. 2016

Development of an Extraction Chromatography Method for the Analysis of 93Zr,94Nb and93Mo in Radioactive Contaminated Water Generated at the Fukushima Daiichi NPS

60

Institute of Nuclear Materials Management

Oct. 17-18, 2016

Others Data Sheet

Others Data Sheet

15

Impact assessment on rust preventives for RPV / PCV on water treatment facility
Overview of development of Fukushima Daiichi NPS accident waste treatment anddisposal technology

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

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

No.

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects
32

No.

Limit of detection of a bare nuclear material debris lump33

Others − Data Sheet

Main Research Results in 2016
Date

Details

Oct. 19-21, 2016

Development of hydraulic manipulator simulator for control system design andevaluation of cotton type model identification method and other tow issues.

RADIEX 2016

Oct. 20, 2016

Efforts of the Japan Atomic Energy Agency for decommissioning

63

RADIEX 2016

Oct. 20, 2016

Overview of A2 investigation / B2 investigation (panel)64

Monthly OPTRONICS, December 2016

Oct. 27, 2016

A Robot Developed to Investigate Primary Containment Vessel(PCV) Interiors - Up to now and future65

INES-5Late Oct. 2016

66

ICMST 2016

67

Science AGORA

61

Japan Fluid Power System Society

62

No.

Project Nameresearch

Category

Subject

1

Upgrading for Identifying ComprehenContract research
sive Conditions inside the Reactor

Inversereactor2

Upgrading for Identifying ComprehenContract research
sive Conditions inside the Reactor

3

PHITS Benchmarking for HPGe Detector Efficiency

Nov. 2, 2016

Development of the Remote Decontamination Robot "MHI-MEISTeR
II" for an Upper Floor of Reactor Building in Fukushima Daiichi NPP

Nov. 5-6, 2016

A2 investigation (panel), exhibition of static device

Nov. 7, 2016

Sic issues such as Phases and morphology in the simulated MCCI
products prepared by arc melting method, etc.

problemevaluation

PartnerwithvirtualPeriodUniversity of Tokyo

Aug. 1, 2016 Feb. 28, 2017

Analysis of pipe deformation during corematerial slumping

University of Tokyo

Aug. 1, 2016 Feb. 28, 2017

Upgrading for Identifying ComprehenContract research
sive Conditions inside the Reactor

MELCOR analysis related to transition of
events during core slumping

Waseda University

Aug. 1, 2016 Feb. 28, 2017

4

Upgrading for Identifying ComprehenContract research
sive Conditions inside the Reactor

Segregation analysis of MCCI reactants atthe time of melting and solidification

Tohoku University

Oct. 17, 2016 Feb. 28, 2017

5Upgrading for Identifying ComprehenContract research
sive Conditions inside the Reactor

Research on supplementary reaction anddiffusion behavior of cesium

Osaka University

Oct. 17, 2016 Feb. 28, 2017

6

Development of Technology forContract research
Criticality Control in Fuel Debris

Verification test of critical approach detection
system and neutron absorber to be applied tosystems containing fuel debris

Kyoto University

Dec. 26, 2016 Feb. 28, 20177

Development of Fundamental
Technologies for Retrieval of Fuel Contract research
Debris and Reactor Internals

Investigation on discontinuous large load tip
control

Osaka University

Jun. 24, 2016 Feb. 28, 2017

68

Nuclear Materials Conference (NuMat2016)69

Annual Meeting of the International Network of Laboratories
for Nuclear Waste Characterization (LABONET)Nov. 7, 2016

Characterization of Carbonate Slurry generated from Multi-Radionuclide
Removal System in Fukushima Daiichi Nuclear Power Station

70

Annual Meeting of the International Network of Laboratories
for Nuclear Waste Characterization (LABONET)Nov. 7, 2016

Study on H2 gas produced by radiolysis of the waste of carbonate
slurry generated from Multi-Radionuclide Removal System

8

R&D for Treatment and Disposal
Contract researchof Solid Radioactive Waste

Research on criticality safety in geologicaldisposal of damaged fuel at Fukushima
Daiichi Nuclear Power Plant

University of California, Berkeley,
U.S.A

Apr. 1, 2016 Mar. 31, 2017

71

LS-DYNA & JSTAMP Forum 2016

Nov. 8, 2016

Coupled analysis on fluid structure for assisting piping closure

9

R&D for Treatment and Disposal
Contract researchof Solid Radioactive Waste

R&D on radioactive waste management
methods using gamma ray measurement (2)Tokyo Institute of Technology

Apr. 21, 2016 Feb. 28, 201772

The 2nd ITA-ACCELERATE Symposium

Nov. 10, 2016

Large structural imaging using cosmic rays

10

R&D for Treatment and Disposal
Contract researchof Solid Radioactive Waste

Research on estimation/assessment technology of radioactive waste inventory (Phase 2)

Central Research InstituteElectric Power Industry

Fuel debris research committee, Atomic Energy Society ofJapan

Nov. 16, 2016

74

ROBOT FESTA FUKUSHIMA

Nov. 19, 2016

Overview of A2 investigation / B2 investigation (panel), display of static A2
investigation device, demonstration of B1 investigation device prototype

75

Electronic journal of Japan Society of Maintenology

Nov. 25, 2016

High-pressure Water Jet Decontamination Apparatus for Upper Part of 1st Floor of
Reactor Buildings for Fukushima Daiichi Nuclear Power Station, and other two subjects

76International Conference on the Safety of Radioactive Waste
Management

Nov. 25, 2016

Inventory Estimation for Accident Waste Generated at the
Fukushima Daiichi NPS

77

International Conference on the Safety of Radioactive Waste
Management

Nov. 25, 2016

Approaches of Selection of Adequate Conditioning Methods for
Various Radioactive Wastes in Fukushima Daiichi NPS

78

NTHAS 10

Nov. 27-30, 2016

Validation of Debris Freezing Model in a Penetration Tube for the
SAMPSON Code against the GEYSER experiments

79

Technical Committee on Laser Processing Technology, theInstitute of Electrical Engineers of Japan

Nov. 30, 2016

Laser application for fuel debris retrieval at Fukushima Daiichi
NPS

80

The Society of Instrument and Control Engineers

Dec. 15, 2016

Construction and basic experiment of RSNP remote control
system of multiple crawler robots

81"Fuel debris research committee," Division of Fuel Debris,
Atomic Energy Society of Japan

Jan. 26, 2017

Overview of development of fuel debris criticality management
technology

82

Health Physics

Feb. 2, 2017

Radiochemical analysis of debris samples collected at Fukushima
Daiichi NPS

83

IAEA "First Coordinated Research Meeting and Consultancy
Meeting on Management of Severely Damaged Spent Fuel

Feb. 14, 2017

Characterization to Estimate Fuel Debris properties Generated in
Fukushima Daiichi NPS

84

WM2017(Waste Management Symposia)

Mar. 5, 2017

Long Reach Manipulator for PCV Repair at Fukushima Daiichi andother two issues.

85

Annual Spring Meeting 2017, Atomic Energy Society of Japan

Mar. 27-29, 2017

Development of analysis method for Pd-107 in radioactive waste
collected at Fukushima Daiichi NPS

86

Annual Spring Meeting 2017, Atomic Energy Society of Japan

Mar. 27-29, 2017

Analysis and evaluation of assessing conditions inside reactor of the TEPCO Fukushima
Daiichi NPS, seven subjects related to SAMPSON, and another one subjectofMay 20, 2016 Jan. 31, 2017

Others Data Sheet

Others Data Sheet

73

Investigation result on fuel debris characterization- Large-scale MCCI test, MOX simulated debris, U-containing simulated
debris, comparison with TMI-2

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

Main R&D Achievements In-house R&D Activities
34

Presented at/by

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

No.

List of Joint Researches / Contract Researches

35

Others − Data Sheet

Main R&D Installations / Equipment
Project Name

Details

Development of remotely operated decontamination technology

Suction/blast decontamination equipment for high places

51

Fuel debris characterization

Carbon coater

2

Development of remotely operated decontamination technology

Suction/blast decontamination equipment for high places Testing device

52

Fuel debris characterization

Vacuum displacement arc melting furnace3

Development of remotely operated decontamination technology

Dry ice blast decontamination equipment for high places

53

Fuel debris characterization

Fuel debris compression test device

4

Development of remotely operated decontamination technology

Dry ice blast decontamination equipment for high places Testing device

54

Fuel debris characterization

Fuel debris sonic speed measuring device

5

Development of remotely operated decontamination technology

High-pressure water jet decontamination equipment

55

Fuel debris characterization

Metallographic image analysis device

6

Development of remotely operated decontamination technology

High-pressure water jet decontamination equipment Testing device

56

Fuel debris characterization

Dynamic micro hardness tester

7

Development of remotely operated decontamination technology

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

57

Fuel debris characterization

Simultaneous thermal analysis system

8

Development of remotely operated decontamination technology

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

58

Fuel debris characterization

Gas piping valve heater

9

Development of remotely operated decontamination technology

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

59

Fuel debris characterization

Sample cutting machine

10

Development of remotely operated decontamination technology

Decontamination equipment for upper floors Testing device

60

Fuel debris characterization

Sample polisher

11

Development of remotely operated decontamination technology

Cart positioning measurement software

61

Fuel debris characterization

Core collecting device

12

Development of remotely operated decontamination technology

Interference verification software

62

Fuel debris characterization

Laser diffraction/scattering particle size distribution measuring device

13

Development of remotely operated decontamination technology

Crawler cart for cooperation control verification

63

Fuel debris characterization

Dry automatic density meter

14

Development of remotely operated decontamination technology

Durability test device for cable hose

64

Fuel debris characterization

Heating furnace for thermal analyzer

15

Heating / Feed water equipment

65

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

Full-scale test device

Turbid water treatment equipment

66

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

Full-scale test equipment

Work floor

67

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

1/4-scale test device

Mock-up transfer rail

68

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

1/4-scale test equipment

19

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

Full-scale mock-up

69

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

1/4-model test equipment

20

Development of technology for investigation inside the PCV

B1 investigation device

70

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

1/4-scale reaction force holding mechanism combination test body

21

Development of technology for investigation inside the PCVDispersion prevention equipment for B1 investigation device

71

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

Test body for water shielding of upper part of RPV

22

Development of technology for investigation inside the PCV

Subsidiary equipment for B1 investigation device

72

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

Set of flexible structure arm

23

Development of technology for investigation inside the PCV

B1 investigation device Simulated device for mock-up

73

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

Flexible structure arm control device

24

Development of technology for investigation inside the PCVShielding block removal equipment

74

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

Equipment for mock-up in PCV

25

Development of technology for investigation inside the PCV

Fuel debris measurement equipment Equipment for elemental test75

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

Equipment for hatch carrying-in test equipment

26

Development of technology for investigation inside the PCV

Fuel debris measurement equipment Equipment for elemental test76

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

1/1-scale hydraulic reaction force holding mechanism

27

Development of technology for investigation inside the PCV

A2 investigation equipment (including chambers and guide pipes)

77

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

1/1-scale electric motor-driven reaction force holding mechanism

28

Development of technology for investigation inside the PCV

Set of X-6 penetration hole boring device

78

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

Laser gouging power measurement unit

29

Development of technology for investigation inside the PCV

Set of previously confirming equipment inside the penetration

79

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

Laser gouging head30

Development of technology for investigation inside the PCV

Set of deposits removal equipment (including chambers)

80

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

Equipment for access device elemental test

31

Development of technology for investigation inside the PCV

Set of subsidiary equipment for A2 investigation

81

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

Robot arm

32

Development of technology for investigation inside the PCV

Set of pre-investigation equipment inside the pedestal

82

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

Access rail

33

Development of technology for investigation inside the PCV

Set of simulated mock-up structure in the PCV

83

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

PCV welding device

34

Development of technology for investigation inside the PCV

Set of elemental test device for A3 investigation

84

R&D on treatment and disposal of solid water

Chamber for alpha nuclide analysis

35

Development of technology for investigation inside the PCV

Set of relevant equipment for hatch opening device

85

R&D on treatment and disposal of solid water

Digital spectrometer

36

Development of technology for investigation inside the PCV

B1 investigation device

86

R&D on treatment and disposal of solid water

Efficiency calculation program for gamma-ray measurement

37

Development of technology for investigation inside the PCVDispersion prevention equipment for B1 investigation device

87

R&D on treatment and disposal of solid water

Aerosol transition observing device

38

Development of technology for investigation inside the PCV

Subsidiary equipment for B1 investigation device

88

R&D on treatment and disposal of solid water

Well type Ge detector

39

Development of technology for investigation inside the PCV

B1 investigation device Simulated device for mock-up

89

R&D on treatment and disposal of solid waterCore sampling device

40

Development of technology for investigation inside the PCV

Set of prototype underwater swimming device

41

Development of technology for investigation inside the PCV

Set of remote X-6 penetration hole boring device

42

Development of technology for detection of duel debris in the reactor

Shielding material of the measurement equipment for transmission method

43

Development of technology for detection of duel debris in the reactor

Small measurement equipment for transmission method

44

Development of technology for detection of duel debris in the reactor

Small muon tracking system for scattering method

45

Development of technology for detection of duel debris in the reactor

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

46

Fuel debris characterization

Large capacity of thermogravimetric balance and simultaneous thermal analysis equipment

47

Fuel debris characterization

Piezoelectric crystal four-component cutting dynamometer48

Fuel debris characterization

Elemental analysis system for SEM

49

Fuel debris characterization

Automatic hydraulic embedding machine

50

Fuel debris characterization

Inverted metallurgical microscope

Main R&D Achievements In-house R&D Activities
Others Data Sheet

Others Data Sheet

1

18

No.

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

Project Name

17

36

Over 1 million yen

No.

16

Details

Main R&D Achievements Subsidized Projects

Main R&D Achievements Subsidized Projects

Over 1 million yen

37