decommissioning
while gathering knowledge
Our visions

The decommissioning of Fukushima Daiichi is an extremely difficult task that is unprecedented worldwide.

IRID is developing technologies based on accumulated knowledge and wisdom to overcome technologicalchallenges.

To conduct testing and research for the decommissioning of nuclear power stations and to
implement projects aimed at improving the technological prowess of IRID member
organizations with the goal of producing technologies for practical application.

From the standpoint of strengthening the foundation of nuclear decommissioning technology,

Greeting

we devote ourselves to research and development (R&D) of technology for the current andPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

IRID commits to R&D for the nuclear
by working together as one teamwisdom from around the world.

most urgent challenge: the decommissioning of the Fukushima Daiichi Nuclear Power Station(NPS).

The International Research Institute for Nuclear Decommissioning (IRID) wasestablished as a non-profit mutual benefit corporation in August 2013, withthe purpose of conducting tests study for the decommissioning of nuclear

We work on R&D projects eﬀectively and eﬃciently while advancing integrated projecttechnological standards in accordance with the Act on Technology Research

onsite deployment at the Fukushima Daiichi NPS as we face numerous extremely

power station (NPS), and improvement and practical application of

Toyoaki Yamauchi
President of IRID

management to develop and propose the best technologies and systems for rapid

Associations.

diﬃcult technological challenges.

Since its establishment, IRID has been engaged in research and development

We build the optimal R&D structure by cooperating with relevant organizations andgathering knowledge from Japan and abroad.

(R&D) of technologies necessary for decommissioning based on the needs ofTokyo Electric Power Company (TEPCO) Holdings, Inc., which is undertaking

We actively promote eﬀorts to develop and secure the human resources that will

onsite work in accordance with the Mid- and Long-Term Roadmap forcomprise the next generation of workers for nuclear decommissioning and related

Decommissioning Fukushima Daiichi NPS of TEPCO Holdings Inc.; thetechnologies, including eﬀorts to collaborate with universities and research institutions.

Technical Strategic Plan for Decommissioning Fukushima Daiichi NPS ofTEPCO Holdings, Inc.; and the R&D Mid-and-Long-Term Plan, which are

We strive to release information on our R&D activities and results to reassure and reachissued by the Nuclear Damage Compensation and Decommissioning

an understanding with Japanese citizens, including those in Fukushima, as well as the

Facilitation Corporation (NDF).

international community.

The reactors of Fukushima Daiichi NPS were damaged by the accident, andWe form an international research hub (center of excellence) through our R&D

To proceed with the decommissioning work in this environment, IRID has

rapid restoration of Fukushima and the improvement of technological capabilities in thethe surrounding area is a severe environment that humans cannot approach.

activities, thus contributing to the decommissioning of the Fukushima Daiichi NPS andmade it our mission to challenge technological development with the courageinternational community.

and wisdom necessary for this unprecedented decommissioning task.

So far, the technologies developed by IRID for use in the decommissioning

include confirming the conditions inside the primary containment vessel

Contents

cosmic-ray muons to identify the location of fuel debris. Currently, IRID is

Overview of IRID・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 03
Organizational Profile
Organizational Structure
Roles of IRID・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 05(PCV) through robotic investigation and a tomography technique uses

researching and developing the technologies needed to perform detailed
investigations within the PCV and retrieve fuel debris.

IRID is committed to the reconstruction of Fukushima as well as technologicalcontributions that transform Japan into a technology-driven nation. We also
would like to cultivate young talent for the next generation involved indecommissioning technology.
01

Activities of IRID
Roles of Organizations in the Decommissioningof Fukushima Daiichi NPS
Overview of Mid-and-Long-Term Roadmap
Scope of R&D Conducted by IRID

IRID' s R&D Projects・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 07
Three R&D Projects for Nuclear Decommissioning

Decommissioning Research and Development・ ・・・ 09
IRID R&D Progress in 10 years・・・・・・・・・・・・・・・・・・・・・・ 13
R&D Cooperation with Overseas Organizations・・ 115
Human Resource Cultivation in R&D・ ・・・・・・・・・・・ 117

02

Overview of IRID
Organizational Structure

1.Name6.Board of Directors

International Research Institute for Nuclear Decommissioning(IRID)

President: Toyoaki Yamauchi
Vice President: Tamio AraiManaging Director: Akihisa Heike
Directors: Shunji Yamamoto, Hiroshi Arima, Satoshi Ueda,
Satoshi Sekiguchi, Masaru Taniguchi, Kentaro Funaki,
Noritoshi Maeda, Junichi Matsumoto
Auditor: Masao Nakanishi

2.Location
5F, Toyokaiji Building No.3, 2-23-1 Nishishimbashi,
Minato-ku Tokyo 105-0003, Japan
Phone: +81 3 6435 8601

3.Date of Establishment
August 1, 2013
The establishment of IRID was approved by the Minister of
Economy, Trade and Industry in accordance with the
Research and Development Partnership Act.

General Meeting
Board ofDirectors
Technology Advisory
Committee

R&D Strategy Planning Dept.

7.Personnel
681 people* (Excluding directors)

Planning & Management Group*Including members of the above membership organizations who engage inInternational
Advisors

R&D Management Dept.

Administration Dept.

Naraha Development
Center

Accounting GroupPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Organizational Profile

International &Academia Cooperation Group

research projects. (As of March 31, 2022)

Corporate Communications
Group

4.Scope of Work
■ R&D for nuclear decommissioning■ Cooperation with overseas organizations■ Human resource cultivation in R&D(As of April 1, 2022)

5.Members (18 organizations)Technology Research Associations are mutual aid organizations (non-proﬁt mutual beneﬁt corporation) that conduct joint
research on technologies for use in industrial activities, for the beneﬁt of association members. IRID was established as aTechnology Research Association to rapidly systemize its activities and take advantage of the transparency and ﬂexibility interms of running the organization.

National research and development agency:
Japan Atomic Energy Agency
National Institute of Advanced Industrial Science andTechnology
Plant manufacturers, etc.:
Toshiba Energy Systems & Solutions Corporation
Hitachi-GE Nuclear Energy, Ltd.
Mitsubishi Heavy Industries, Ltd.
ATOX Co., Ltd.
Electric utilities, etc.:
Hokkaido Electric Power Co.,Inc.
Tohoku Electric Power Co., Inc.
Tokyo Electric Power Company Holdings, Inc.
Chubu Electric Power Co., Inc.
Hokuriku Electric Power Company
The Kansai Electric Power Company, Inc.
The Chugoku Electric Power Co., Inc.
Shikoku Electric Power Company, Inc.
Kyushu Electric Power Company, Inc.
The Japan Atomic Power Company
Electric Power Development Co., Ltd.
Japan Nuclear Fuel Limited

Overview of the Technology Research Association Model

Project costs
180

148 143 149 143 145

150

168
142

120

120

Features of Technology Research Association

90

Each association member provides researchers, funds, and facilities for use in joint research. The result of the research are jointly managed andutilized by all members.
● Technology Research Associations are joint research organizations that have a legal identity independent of association members.
● Transparency and reliability of the association's management can be increased with the approval of the minister in charge and by holding
regular member meetings as well as meetings of the board of directors.
● Those who directly or indirectly utilize the results of joint research (including corporations, individuals, foreign companies, and foreign
nationals) can become association members.
● Since universities, R&D institutes, technical schools, local governments, and foundations primarily engaged in testing and research canparticipate as association members, the Technology Research Association serves as a conduit for collaboration between industry, academia,
and government.
●

60

46

30

0

03

(100 million yen)

(FY)

*Source: "Technology Research Association" on the Ministry of Economy, Trade and Industry website.

04

Roles of IRID
other organizations involved in the decommissioning of Fukushima
Activities of IRIDof the Japanese government while closely cooperating with

Daiichi NPS.

Overview of Mid-and-Long-Term Roadmap

IRID has a three-pronged strategy: R&D on decommissioning, cooperation with domestic andoverseas organizations, and the cultivation of human resources.
IRID is an organization composed of 18 corporations that play leading roles in

research and development (R&D) on the decommissioning of Fukushima
Daiichi Nuclear Power Station (NPS).

Currently, IRID conducts R&D on the urgent issue of decommissioning(Revised on December 27, 2019)

The decommissioning of Fukushima Daiichi Nuclear Power Station (NPS) is proceeding basedon the "Mid-and-Long-Term Roadmap for Decommissioning of Fukushima Daiichi NPS of
Tokyo Electric Power Company (TEPCO) Holdings, Inc." (the "Mid-and-Long-Term Roadmap")
formulated by the Japanese government.

For decommissioning

R&D

Fukushima Daiichi NPS based on the Mid- and Long-Term Roadmap
developed by the government while aiming to cultivate, accumulate, andDecember 2011
(Step 2* completed)

upgrade technologies necessar y for future nuclear decommissioning

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

IRID works on R&D for decommissioning based on a major policy

throughout Japan.

In addition, it is essential to amass knowledge from within
Japan and abroad to proceed with the decommissioning ofFukushima Daiichi NPS, a globally unprecedented ande x tre m e l y d if fi cult t a s k ; t h e refo re , I R I D i s p ro m otin gcollab o ratio n with re late d d o m e stic a n d inte rnatio nal
organizations. Moreover, IRID commits to the cultivation of the

For decommissioning

Cooperation with domestic
andinternational organizations

For R&D related to decommissioning

Human resource
cultivation

human resources needed to continue the decommissioning ofFukushima Daiichi NPS.

IRID has been engaged in various R&D projects under the Mid-and-Long-Term Roadmap. These R&D projects produced sucessful

results in investigating the inside of the primary containment vessel (PCV) by using remote-operated robots and imaging the insideof PCV using cosmic-ray muons radiography, while clarifying technical challenges that must be overcome.

IRID will continue to tackle these issues as it commits to the R&D needed to begin retrieving fuel debris from the initial unit.

Scope of R&D conducted by IRID

Steady implementation of decommissioning tasks

Decommissioning work

Overview of R&D■ Continue to maintain the cold shutdown state
Information provision andthe practical application of
the achievements ofdevelopmentof the reactor

Nuclear Damage Compensation
and Decommissioning Facilitation
Corporation (NDF)

On-site needs■ Process the accumulated water(countermeasures for contaminated water)■ Reduce radiation dosages throughout the plant

Decommissioning ofFukushima Daiichi 1F

and prevent the spread of contamination■ Remove fuel from spent fuel pool
■ Retrieve fuel debris

R&D ﬁeld
conducted by
IRID

Practical application ofdevelopment■ Plan for the storage, management,

and disposal of solid waste■ Plan for the decommissioning of

nuclear facilities

Funds for R&D

R&D field conducted by IRID

Applied research

Basic research
Fundamental reserach(development of common fundamental technology;
organization of basic facilities;
the acquisition of basic data)
Partially revised by reference to the NDF Technical Strategic Plan 2021

05

06

IRID' s R&D Projects
considering appropriate approaches and methods of reducing

Explanation of technical terms related torisks, focusing on the end state (the most appropriate state).

the reactor building

Located on the top ﬂoor of the reactor building, this tankholds the water used to store racks of spent fuel until the
Operating ﬂoor

Three R&D Projects for Nuclear Decommissioning

heat generated by the decay of ﬁssion products haslessened.

Shield plug

This subtance is formed when high-temperature fuel melts

IRID has three major R&D projects: R&D related to removing fuel from spent fuel pool, R&D related to preparing for thealong with control rods and structures inside PRV thenretrieval of fuel debris, and R&D on the treatment and disposal of solid radioactive waste. IRID will further advance its R&D

cools and resolidiﬁes.

based on the Fuel Debris Retrieval Policy issued in 2017.

A safety structure comprised of a ﬂask-shaped containerthat houses equipment such as RPV; contained

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

To advance the decommissioning strategies, IRID has beenradioactive material at the time of the accident.

Located in the basement of the reactor building, thisdoughnut-shaped device stores water. The S/C condenses
vapor generated if there is a break in the reactor's piping

R&D 1

R&D 2

R&D 3

R&D for removing fuel from

R&D for preparing to retrieve

R&D for treating and disposing

the spent fuel poolfuel debris

of solid radioactive waste

to prevent excess pressure from building up. It also serves

the important function of providing a water source for theEmergency Core Cooling System (ECCS) during incidentsof coolant loss.

■Two-dimensional ﬁgure of the reactor building

Policy for Fuel Debris Retrieval and the Current Approach
IRID will proceed with fuel debris retrieval based on the following policy, taking into account feasibility evaluations andproposals for fuel debris retrieval methods that were studied in the Technical Strategic Plan issued by the Nuclear Damage
Compensation and Decommissioning Facilitation Corporation (NDF).

Step-by-step approach
Start fuel debris retrieval at a small scale then gradually increase theretrieval scale of fuel debris while working ﬂexibly

Optimization of the entire decommissioning process
Consider a comprehensive plan aimed at total optimization of the entire
decommissioning process, including preparation work, and retrieval,transport, processing, storage of fuel debris until clean-up of the site

Combination of multiple methods
Study multiple methods of retrieving fuel debris under the assumptionthat the primary containment vessel (PCV) will be accessed from thebottom and the reactor pressure vessel (RPV) will be accessed from the
top.

07

Initiative focused on the partial-submersion method
Currently, the submersion method is problematic due to the diﬃculty inpreventing leaks as well as the amount of radiation exposure during
retrieval work.
* The submersion method does have some advantages such as its shielding eﬀect,
so it may be reexamined in the future.

These connecting pipes transport vapor generated in the D/W by abreak in a reactor pipe to the S/C. Eight vent pipes are installed in
PCVs of Units 1-3 at Fukushima Daiichi NPS.

A room containing the torus-shaped (doughnut-shaped) S/C in thebasement of the reactor building.

A cylindrical steel container that houses the fuel assemblies. This

container can withstand the high-temperature water and high-pressure
steam generated by nuclear-ﬁssion energy inside the RPV, which ishoused within the PCV along with cooling equipment etc.

Prioritizing the method of accessing
the bottom of the PCV from the side
Fuel debris is located both at the bottom of the PCV and within the RPVof each reactor.
To swiftly reduce risks and minimize the increased danger associated withretrieval, the method for accessing the bottom of the PCV from the sideis prioritized in consideration of the following:① it oﬀers the best accessibility to the bottom of the PCV, which helps
build knowledge through investigations of the interior;② retrieval can begin at an earlier stage; and③ the work can proceed simultaneously with the removal of spent fuel.

A steel container that houses RPV, cooling equipment, and otherapparatus that perform important functions. This preventsradioactive material from being released outside under abnormal
plant conditions, such as when an accident occurs or the coolingequipment breaks down. Each of PCVs in Units 1-3 of Fukushima

Daiichi NPS consists of a ﬂask-shaped drywell, a doughnut-shaped

suppression chamber, and vent pipes connecting the drywell to the
suppression chamber.

08

31 Upgrading for identifying
comprehensive conditions
inside the reactor

Ⅰ Spent Fuel Management

19 Evaluation of long-term integrityof fuel assemblies removed
from spent fuel pool andstudy of methods to process
damaged fuelWhite sediment

Analysis and evaluation techniques for inferring the conditions
inside the RPV and PCV, which is essential for studying technology forfuel debris retrieval.
Understanding a reactorʼs internal conditions

Appearance of spent fuel from Unit 4Integrity assessments for the long-term storage of removed fuelas well as studies on the processing of damaged fuel.

Ⅱ Decontamination and Dose Reduction

23 Development of technology
for remotely operationin the reactor buildings

Suction/Blasting decontamination equipment

Ⅲ Investigating and Analyzing inside PCV/RPV

43 Development of technology
for detailed investigation inside
the primary containment vessel(PCV)

fuel debris

The Unit-1 submersible robot

Investigations of the inside of the PCV toobtain detailed information for use in fuel debris retrieval.
Muon

RPV
ReactorTechnique of measuring muons tomore promptly estimate the distribution offuel debris in the reactor.

Muon
Detector

Muon measurement

53 Development of technology for investigation
inside the reactor pressure vessel (RPV)
Technology used to investigate inside the PCV in preparation
for fuel debris retrieval.
Investigating inside the RPV

29 Development of technology for
non-destructive detectionof radioactive materials accumulated
in the suppression chamber

58 Integrity evaluation of PCV/RPV
and developmentof corrosion control technology

Methods of estimating and measuring the accumulationof radioactive material within the suppression chamber (S/C) etc.
Distribution of fuel-originating radiation in
the vicinity of Unit 1ʼs S/C

09

Example ofDevelopment of analysis technology and the acquisition of data
for ascertaining the properties of fuel debris,which is needed to study fuel debris retrieval technology.

Technology for decontaminating reactor buildings
with remote-operated equipment in preparation
for the retrieval of fuel debris.

27 Development of technology
for detection of fuel debris
in the reactor

35 Development of analysis anddetection technology
for characterization of fuel debris

Development of anti-corrosion technology as well as
Testing the durability of the RPV pedestal
integrity assessments of the PCV and RPV based onan earthquake-resistance assessment that considers corrosion and other factors.

10

95 Development of technology
for criticality control of fuel debris

Ⅳ The Water Circulation System and Repairing the PCV

63 Development of repair technology
for PCV and full-scale test

Research and development for detecting andcontrolling abnormalities promptly to prevent criticality during
fuel debris retrieval.

Technology for repairing leaks in the primary containment
vessel (PCV) prior to fuel debris retrieval.

Detector installed on the arm
to measure subcriticality

The S/C shell and downcomer assembly

71 Development of technology for establishing
the water circulation systemin the primary containment vessel (PCV)(development / full-scale test)

Descends

Structural test
(dropped vertically)

Descends
Pump Pit

The dry-well intake line

Ⅴ Retrieving Fuel Debris

75 Development of trial retrieval
and technology for increasing
the retrieval scale of fuel debris

Structural test
(dropped at an angle)

Ⅵ Treatment and Disposal of Solid Waste

109 Research and developmentof treatment and disposalof solid radioactive waste

Dustproof area
(controlled with negative pressure)

For transporting HIC etc

Research and development related tothe treatment and disposal of solid waste,covering everything from understanding
the properties of waste to handling its disposal.

Injector
Dehydrator
Slurry extraction and transport equipment

Pressurized ﬁltration method

Collection equipment (prototype)107 【 IRID Research and Development 】
IRIDʼs History in Photos

79 Development of technology
for retrieval of fuel debris
and reactor internals

115 R&D Cooperation with Oversea Organizations

Research and development aimed at manufacturing
the equipment required for fuel debris retrieval.

Removing obstacles fromthe bottom of the reactor

11

101 Development of technology for containing,transfer and storageof fuel debris
Development of canisters for containing,transporting, and storing fuel debris safely
over the long term.

Systems that pump water into the dry well (D/W) andsuppression chamber (S/C).

Development of methods for experimentally retrieving
fuel debris and for continuing to gradually increase
the scale of the retrieval.

Detector installed next to the processing
tool to continuously monitor the processing
work and to watch for approaching criticality
before and after debris processing

117 Human Resource Cultivation in R&D
12

IRID R&D Progress in 10 years
Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022Ⅰ.

Spent FuelManagement

Integrity of fuelassemblies
removed fromspent fuel poolⅡ.

Decontamination Remote-operated
anddecontamination
Dose Reduction

Detection offuel debris(cosmic-ray muon)
Non-destructive
detection of radioactive
materials in the suppression
chamber (S/C)
Identiﬁcation of
conditions inside
the reactor

Fuel debris
characterization andanalysis technology

Ⅲ.

Investigating
andAnalyzing
inside PCV/RPV

Investigation inside
PCV(detailed investigation)

Investigation inside
RPV

Integrity of PCV and
RPVCorrosion control
for PCV and RPV

Establishment of IRID
13

14

IRID R&D Progress in 10 years
Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Development of
repair technology

Ⅳ.

The Water
Circulation
System andRepairing the PCV

for PCV anda full-scale test

PCV water
circulation system

Retrieval methods,
systems, andfundamental andsampling technologies
for fuel debris

Ⅴ.

Retrieving
Fuel Debris

Criticality control※ Combined with project of Development of Technology
for Further Increasing Retrieval Scale of Fuel Debris
and Internal Structures (FY2018 supplementary)technology

Small neutron
detectionmeasurement
Integration
Management

Containing,transportation andstorage of fuel debris

Ⅵ.

Treatment andDisposal of
Solid Waste

Ⅶ.

OthersSolid waste
manegement

Contaminatedwater

Establishment of IRID
15

These R&D projects have been conducted as part of the subsidies for Countermeasures for Decommissioning andContaminated Water granted by the Ministry of Economy, Trade and Industry (METI).

16

Research and Development of
Nuclear Decommissioning
The decommissioning of Fukushima Daiichi Nuclear Power Station is often compared with theAsteroid Explore Hayabusa project. The two share many similar features, including the usage of
remote-controlled robots in unknown environments. They also require the stable operation ofelement components in severe environments and systems that can respond to various events. A
single prototype was subjected to mock-up testing then be applied in the site at an appropriate
timing. In addition, all hopes were necessarily pinned on that first unit.
This creates the need for various technologies—not only those for retrieving and transporting fueldebris, decontamination and investigation of the inside of the reactor but also storage methods
developed by analyzing the properties of retrieved material and the construction and connectionof earthquake-resistance structures. IRID has striven to ensure safety from many angles by treating
the entire decommissioning process as a single system and by sharing information during thedevelopment phase. We achieved this through the activities of working groups for each technicalfield; the groups operated in parallel with each project but in ways that extended beyond the
project's framework.
Since 2015, IRID has also been conducting the design review meeting through the participation of
external experts. A total of 345 review meetings have been conducted as of late March 2022.
These review meetings identify technical issues such as insufficient technical study, the potential for
new risks, and on-site challenges. Collaborating with Tokyo Electric Power Company Holding, Inc.,
which understands the on-site conditions, site surveys were also carried out to confirm details
about the state of the worksite.The participants were engineers involved in a wide range oftechnological development, so the meetings proved useful in terms of on-the-job training andcultivating talent for the field of nuclear decommissioning.

members, especially between engineers engaged in research and development. Each organization
has its own terminology and way of thinking, and the nuclear industry is no different. Through ourmany meetings where we explained technical documentation to one another, we've attained ashared understanding that extends to the conceptual level.Furthermore, we believe that
decommissioning technology is "social technology", and have worked hard to publicize technology
so that people in surrounding communities can understand the direction it is heading.
I greatly hope that many people read this pamphlet and learn about our research anddevelopment.

Dr. Tamio Arai
Vice PresidentInternational Research Institute for Nuclear Decommissioning (IRID)

17

Spent FuelManagement
Impact assessment was carried out to obtain
verified data related to the integrity (i.e. anticorrosion, etc.) of the fuel assemblies removed

from the spent fuel pool when they are stored in

the common pool, etc. for a long period of time.

I am proud of the fact that IRID has successfully eliminated barriers between its organizational		Ⅰ

Decommissioning Research and Development

Considering the possibility of physical damage

from rubble fragments and salt adhesion fromt h e s e a w a t e r, c r i t e r i a f o r j u d g i n g t h ere p ro c e s s i n g p o s s i b ilit y of t h e re m ove ddamaged fuel were created.

Ⅰ

Decommissioning
Research and Development

Spent Fuel Management

Evaluation of long-term integrity of fuel assemblies removed from spent fuel pool and study of methods to process damaged fuel1

NutRegarding fuel safety, one of the most important things is the safety of
the fuel’s upper and lower portions, which are placed under a great

deal of stress when the fuel is handled. Therefore, a strength test as
well as an immersion test were conducted on materials in a simulated■ Evaluated the long-term integrity of storing fuel that was exposed to seawater, water mixed with rubble, etc. in the common
pool for long periods of timeenvironment that contains rubble etc. (Figure 1).

The tests did not reveal any corrosion or deterioration in physical■ Assuming that dry storage will also be used, assessed the long-term integrity of damaged fuel as well as the effects of moisture
content within pieces of rubble that are wedged into crevices

strength that could impact safety. A visual inspection of fuel retrieved

Upper-Tip Plugfrom Unit 4’s pool was performed and the thickness of the oxide film

■ As a basic test, looked at the migration of seawater components to fuel materials and evaluated the effects of the radioactive
environment on corrosion from seawater and other substanceson the cladding was measured. Although the presence of white▲Cross-sectional view of test piece that simulates the upper end

sediment was confirmed, anything that would have an impact on safety■ Assessed the technical challenges in reprocessing fuel that may have sustained damage and conducted evaluations for use indetermining whether fuel can be reprocessedof a fuel assembly (left) and results of visual inspection (right)

was not observed. This white sediment is composed primarily of Mg,(90 ℃ , chlorine-ion concentration of 2500 ppm, 1,000 hours of immersion

followed by 60 ℃ , chlorine-ion concentration of 100 ppm, and 7,409 hours ofand Al and Si levels half of Mg, with Cl levels falling below detectable

immersion) (Figure 1)

limits (Figure 3). Considering the results of the electrochemical test as

Research and development progress

well, it was concluded that there is virtually no effect on corrosion.

Unit 4 Locknut

White Sediment▲

Results of visual inspectionof spent fuel from Unit 4

Mg

Al

Cl

Ca

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

IRID assessed the integrity of the long-term storageof removed fuel and conducted studies on the processing of
damaged fuelEvaluating the long-term safety of fuel stored
in the common pool(upper tie plate of fuel mixed
with rubbles) (Figure 2)

Si▲Chemical analysis of white sediment on Unit 4 locknut (Figure 3)

2

Evaluating the long-term integrity of fuel in dry storage
Assuming that dry storage will be used for some fuel assemblies

p re cip it ati o n te s t reve a l e d n o im p a c t o n th e b e hav i o r

for fuel assemblies that were being affected in multiple ways

damage from rubble and seawater adhesion, are present (Figures

removed from the spent fuel pool, the safety of using dry storage(such as by falling rubble or seawater) was evaluated byconducting a hydride precipitation test and a creep test to

ascer tain the impac t of Fukushima Daiichi NPS’s uniquePurposeThe plan is to store fuel assemblies that are removed fromCorrosion and strength tests designed to simulate actual

pool within the power station until a future decision is madewhether the common pool can be used to safely store fuelthe spent fuel pools of Unit 1 through 4 in the common
regarding their processing or disposal. However, these fuelassemblies have been exposed to a unique environment
since they mixed with incoming seawater and falling rubble
during the accident.

Therefore, the long-term integrity of fuel assemblies must be

verified after they have been transported to the common
pool. Whether potentially damaged fuel has any effect on

future processing or disposal procedures must also be
determined.

4 and 5). In addition, only a small impact on creep behavior wasobserved (Figure 6).

circumstances on the properties of the material. The hydride
DamagedSectionstorage conditions were conducted in order to evaluate

Enlarged

DamagedSectionFuel Cladding
Tube

Enlarged

Fuel Cladding
Tube
Final Test (1F Long-Term Integrity H27-28), Irradiated 9×9, Comparison

Final Test (1F Long-Term Integrity H27-28), Irradiated 9×9, Damage + Seawater

Final Test (1F Long-Term Integrity H27-28), Irradiated High-Burnup 8×8,that may have been damaged or deformed by contact with

Circumferential Strain (%)

Background(reorientation) of hydrides even when multiple factors, such as

rubble or changes in the water quality of the spent fuel pooldue to the influx of seawater. Because some level of dry

storage is also being considered to ensure that the common

pool has sufficient capacity, simulations to determine thesafety impact of using dry storage for fuel were also

Comparison

Final Test (1F Long-Term Integrity H27-28), Irradiated High-Burnup 8×8,

Damage + Seawater

JNESH15, Irradiated High-Burnup 8×8, No.64

conducted.

And to confirm the technical feasibility of reprocessing fuel—which is one of the measures for disposing of fuel that may
be damaged—the technical challenges faced in reprocessing

were worked out and conducted a basic impact assessment

Hydrides▲ Results of hydride precipitation test
(300 ℃ , cooling rate of 0.3 ℃ /h,circumferential stress of 70 MPa, damaged)(Figure 4)

50μm

Hydrides

50μm▲ Results of hydride precipitation test

(irradiated test piece, 300 ℃ , cooling rate of 0.04

℃ /h, circumferential stress of 70 MPa, damaged,
seawater adhesion, rubble adhesion) (Figure 5)

Testing Time (h)▲ Results of creep speed test

(irradiated test piece, 360 ℃ , circumferential stress of

170 MPa, damaged, seawater adhesion, testing time of 5,000 h)(Figure 6)

for determining whether reprocessing is possible.

1920

Decommissioning

ⅠManagement of Spent Fuel Removal

Research and Development3

Basic testing on the long-term safety of removed fuel
IRID has tested the behavior of seawater components as theyoxide film that is formed on the surface of fuel-rod cladding. In

showed that while seawater components do migrate into

piece, the repassivation potential for crevice corrosion wasmove into the structural crevices of fuel materials. The results

an electrochemical test conducted using a crevice-corrosion test

structural crevices, improvements to water quality decrease thehigher than the electrostatic potential in a water environment

concentration of chlorine ions (Figures 7 and 8) and there is nothat contains a chloride-ion concentration of 100 ppm or less,accumulation within the structural crevices. It was also confirmed

confirming that crevice corrosion does not occur (Figure 9).

that seawater components do not significantly penetrate theUpper-Tip Plug

Nut

Chloride-Ion
Concentration in
Test Solution

Distribution of Cl-36

Chloride-Ion
Concentration inCrevices

Test Piece Using Radioactivity Measurements

Immersion Time (h)▲ Distribution of Cl-36 on the screw at the plug of▲ Results of evaluating the concentration of

the fuel’s upper section (80 ℃ , after 50 hours of
immersion in twice-diluted artificial seawater) (Figure 7)

Room Temperature

chloride ions in structural crevices of theupper plug material (Figure 8)

Electrostatic Potential

facility, taking into account foreign substances such as seawater. By evaluating

items like the corrosive effect on reprocessing equipment (the primary
endpoint; Figure 10), the behavior during processing (Figure 11), and theeffects on the final waste product, it was confirmed that foreign substance has▲ Connection between crevice-corrosion
repassivation potential and chlorideion concentration (Figure 9)(

Literature data[1],[2]Distribution ratio of foreign
substance)

Determination limit
data is described
below.

(a) Distribution of foreign substance during thepresence of simulated FP
FP = fission productDistribution Ratio (D)1000(surface of immersed test piece, 960 hours of immersion) (Figure 10)

No pitting corrosion due to chloride ions etc. was observed; if anything, the rate ofcorrosion was slower, and it was determined that there was no equipment corrosion
caused by foreign substance.

21

investigating inside PCV for fuel debris retrieval,contamination. Comprehensive measures for theinvestigations in the future.

▲ Immersion test in high-level waste liquid tank

To i m p rove t h e wo r k e nv i ro n m e n t w h e n

was developed suited for the level of on-site

fuel would be the basis for determining whether to launch more detailed

Simulated waste liquid from
Rokkasho Nuclear Fuel
Reprocessing Facility(chloride ions: 20 g/L)

Decontamination
and Dose Reduction
remote-operated decontamination equipment

no impact. It was also decided that confirmation of the state of the removed

Simulated waste liquid from
Rokkasho Nuclear Fuel
Reprocessing Facility(chloride ions: 0 g/L)

Ⅱ

Decommissioning
Research and Development
廃炉に関する研究開発

Studying methods for processing damaged fuel
IRID studied the impact of damaged fuel when it is handled at a reprocessing

Dose Rate

Chloride-Ion Concentration (ppm)

Distribution Ratio

4

Crevice Corrosion Repassivation Potential

Chloride-Ion Concentration (ppm)

Test Solution C
Test Solution A
Test Solution B
Test Solution D

100

U

Pu

10

Without
Foreign
Substance
Added

Simulated FP+
Foreign
Substance

Simulated FP+
Foreign Substance+
Silica Colloid

D = organic phase (water-phase total for five times backextraction) / water phase (total for two times extraction)(b) Distribution ratio of U / Pu during the presence of
foreign substance▲ Assessing the behavior of foreign substanceduring processing (Figure 11)

It was found that the distribution ratio of foreign substancewas on the order of 10-2 to 10-3, making it difficult to
extract with a solvent, and that the presence of foreign
substance did not cause a significant change in the U / Pu
distribution ratio.

reduction of worker exposure were studied in

combination with remote-decontamination
technology (equipment) and radiation shielding.

Ⅱ

Decommissioning
Research and Development

Decontamination and Dose Reduction

Development of technology for remotely operation in the reactor buildings

1

IRID developed equipment for decontaminating the floor as well as

Upper Floors(2F/3F)To accommodate the various contaminated forms and objects,

Upper Section

and blasting, high-pressure water jets, and dry-ice blasting (Figures 1

Lower Section

the walls and other surfaces up to 1.8 meters above the floor.

of 1F

equipment that uses three different methods was designed: suction■ Developed equipment for decontaminating the first floor of the reactor building (upper and lower areas)■ Developed equipment for decontaminating the upper (second and third) floors of the reactor building

through 3).

of 1F

Research and development progress▲ Suction and blast

decontamination equipment (Figure 1)

2▲ High-pressure water jet decontaminationequipment (Figure 2)

mock-ups were considered. In the demonstration test, whether

water jet decontamination equipment, and the dry-ice blast

evaluated in terms of decontamination performance, safety, andremanufactured so that they could be used to decontaminate

Performance was confirmed by conducting demonstration tests

d r y - i c e b l a s t d e c o n t a m i n a t i o n e q u i p m e n t i n te r m s ofapproximate the environment were selected that the actualmachines will be operated in, and factors such as variation
within the facilities and operational priority when building the

To reduce the dose inside the reactor building so that fuelfloor and upper floors of the reactor building were planned;on the first floor and upper floors of the reactor building

reduced and improve the work environment by developing
remote decontamination technology.

23

the high-pressure water jet decontamination equipment anddecontaminating the upper section. The workability of thesuction and blasting equipment on the walls were verified after
the removal of obstacles, which is the primary target for thatwork. As a result, the goal of applying what had been learned
to the actual machines was achieved.

PurposeTo work towards retrieving fuel debris from Unit 1–3 atcould proceed smoothly, the radiation dose had to be

operability and ease of use when operated remotely.

Among other things, the results confirmed the effective range ofRe ga rd i n g t h e m o c k- u p l o c ati o n, t a rge t fa c ili ti e s t h at

impediment to those operations. To ensure that the work

the devices have the required features were confirmed andthe upper section.

at the mock-up testing facility manufactured for this purpose.

however, the highly radioactive environment was an(Figure 3)）

The three equipment created in 2013 for decontaminating thedecontamination equipment) were redesigned andFukushima Daiichi NPS, the preparation work for the first▲ Dry-ice blast decontamination equipment

Equipment for decontaminating the upper section of the first floor(height of 5–8 m)
lower section (the suction and blasting device, the high-pressure

BackgroundPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Prior to retrieving fuel debris, decontaminate the reactor
buildings with remote-operated equipment

Equipment for decontaminating
the lower section of the first floor

debris can be retrieved, some decontamination equipment
were designed and manufactured, then verified that it wasfunctioning properly by verification tests. Based on the resultsof those tests, the equipment was aimed to be improved anddevelop a remote-operating system. Through this process,▲ Suction and blast decontaminationequipment for high places (Figure 4)▲ High-pressure water jet decontaminationequipment for high places (Figure 5)▲ Dry-ice blast decontaminationequipment for high places (Figure 6)

the remote decontamination technology was established.

24Decommissioning

ⅡDecontamination and Dose Reduction

Research and Development

3

Equipment for decontaminating the upper (second and third) floors
The decontamination equipment for the upper floors uses a

executed on the same set of machines as the decontaminationto access the upper floors (2F/3F) of the reactor building andhigh-pressure water, and dry ice) so that the work could be

elevating work platform that extends from the equipment’s hatch

equipment build for the lower sections (suction and blasting,decontaminate the floors and the walls up to a height of 2

completed using a single, customizable set of machines.

meters. The machines were developed so that they could be

Wo r k b e g a n o n b u i l d i n g e a c h p i e c e o fequipment based on the method for accessing

the upper floors and the design of each device.
The shared set of machines includes a work cart,a transport cart, a support cart, a relay cart, andthe various decontamination units (Figure 7).
Relay Cart

The decontamination equipment was completed

Support Cart

in FY 2015 and thenTransport Cart

Work Cart

Decontamination equipment for the upper floors▲

Equipped with the suction and blast
decontamination equipment

4a verification test was

conducted using a mock-up unit.

(Figure 7)

Demonstration test (example)▼ Decontamination methods (Table 1)

IRID developed separate decontamination equipment

Decontamination
Methodfor the lower sections of the first floor, the higher

sections of the first floor, and the upper floors. How

Summary
ContaminationForm

ef fec tive the machines are on each form ofcontamination were confirmed and a demonstration test
was conducted to verify the overall effectiveness andapplicability of the equipment (Table 1; Figures 8–12).

SuctionBlasting

High-Pressure
Water Jet

Dry-Ice Blasting

Suctions up air
Sprays compressed air Sprays high-pressure Sprays compressed air
andliquid to collecton a steel grid toon a steel grid tocontamination
contamination
chip away and collect
cut and collectin the air
along with water

Separating

○

ー

○

ー

Adhering

ー

○

○

○

Penetratingー○（○）High Pressure][Under VeryーStructure for Higher Section (Piping)

Structure for Higher Section (Support)▲Checking positioning (Figure 9)

debris retrieval. Understanding the properties ofthe fuel debris is essential for developing

technologies to retrieve, contain, transport, andstore fuel debris as well as for studying the
Operating on uneven floor: 50mm

▲Verifying operation (Figure 10)

Excavation Depth (mm)■ Moderate (20 mm/s)▲ Slow (10 mm/s)

Emulated Speed (m/s)(Figure 11)25

Investigation technology has been developed

and the status of the structures in PCVs for fuel◆ Fast (80 mm/s)▲Test piece after blasting (20 mm/s)

Investigating andAnalyzing inside
PCV/RPVfor ascertaining the distribution of fuel debris

Decontamination Head

▲ Mock-up test facility (Figure 8)

Ⅲ

Decommissioning Research and Development▲ Checking target-decontamination performance(Figure 12)

processing and disposal methods, therefore,

analysis and estimation technologies have beendeveloped.

Ⅲ

Investigating and Analyzing inside PCV/RPV

Development of technology for detection of fuel debris in the reactor

Investigating the inside of the reactor with muonsCosmic-ray muons are constantly bombarding the Earth’s

surface, with about one muon passing through an area the■ Within a highly radioactive environment, a measuring system based on the muon-transmission method was establishedmatter, but differences in how they do so and the angles

at which they are scattered by that matter can be used to■ A detection system that utilizes the muon-scattering method (which can identify objects with a high degree of accuracy) was
produced and established an identification algorithm

Muon

Muon

size of the human hand every second. Muons pass throughMuon
Detector

Muon
Detector 2visualize the inside of an object. The transmission method(Figure 1) can obtain internal information using a single■ Muon measurements using the transmission method was conducted at Unit 1 of Fukushima Daiichi NPS and demonstrated
internal imaging technology that uses tomography

Muon
Detector 1

detector for a short time. The scattering method (Figure 2)

uses two detectors to discern objects with a high degree■ The distribution of fuel debris at Unit 3 was measured using the experimentally verified transmission method.

of accuracy. Measurements in Fukushima were only

▲ Tra n s mis sio n m e th o d(Figure 1)▲ Scattering method(Figure 2)

performed with the transmission method due to on-site

Research and development progress

limitations and other factors.

2Demonstration test of transmission method on Unit 1
The measuring equipment (Photo 1) was enclosed in an ironto measure for a period of approximately three months. By

radioactive environments. The unit consists of three layers tovalues predicted by simulations (Figure 4), it was concluded thatshielding frame so that muons could be measured even in highly
eliminate errors when counting muons during measurement. Thisequipment was placed outside the building of Unit 1 and allowed
Muon X-Y Detector

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Using muon measurement to estimate
the distribution of fuel debris in the reactor at an early time1

comparing the measured transmission distribution (Figure 3) to thethere is almost no fuel at the location of the reactor's core.

Observation data

10-cm Thick Iron Shielding Unit

RPV
圧力容器

Predicted value if neither fuelnor water are present
Predicted value if only water ispresent
Predicted value if only fuel is present

PCV

Predicted value if both fuel andwater are present

Transmission Rate

Decommissioning
Research and Development

炉心

ReactorCore

Reactor Core
RPV

PCV

▲ Muon-transmission measuring device (Photo 1)

PurposeThe goal of this project is to ascertain the location and theThere are two techniques for internal imaging using muons:

amount (distribution) of fuel debris in the reactor and tocontribute to the selection of methods for efficiently
retrieving that fuel debris so that the initiatives for

The transmission method measures the amount of muons

planned.

those particles to estimate the size of objects. IRID

decommissioning Fukushima Daiichi NPS can proceed as

After the accident, radioactivity was high inside the reactor
and throughout the premise of Fukushima Daiichi NPS, which
made investigating the inside of the reactor difficult.

Although cameras and other equipment have been placed

inside the containment vessel and are slowly gaining glimpsesof its interior, the inside of the reactor pressure vessel (RPV)

have not yet been seen. At the time, muon tomography was
focused as a method that could be used to ingestigate theinside of the RPV sooner rather than later, so there werecollaborations research institutions* both foreign anddomestic to peer within the RPV.

* High Energy Accelerator Research Organization (KEK); Los Alamos National
Laboratory, U.S.

27

the transmission method and the scattering method.

3

Based on our demonstration test at Unit 1, a smaller

developed this measuring technology within the highly

easier to handle on-site was manufactured, thenDaiichi NPS, performing actual measurements on Unit 1 todemonstrate that it can be used to assess the distribution offuel debris. Then, the technology was further developed foruse on Units 2 and 3.

The scattering method tracks muons before and after they

pass through the reactor, estimating their angles of scattering

to predict the distribution of matter within the threedimensional space of the reactor interior. This method issuperior to transmission for discerning objects, so a detector

was designed and produced, then an algorithm to identify

those predicted by simulations (Figure 4)

Measuring and evaluating the distribution of fuel debris at Units 2 and 3

that pass through the reactor, tracking the path taken byradioactive environment that is characteristic of Fukushima

rate (Figure 3)(Reactor 1; measured February–May 2015)▲Comparison of actual transmission rates and*One pixel is equivalent to approx.
25 cm of reactor cross-section

transmission-method equipment (Photo 2) that wasused it to perform measurements on Units 2 and 3
and obtain the following results.

Height OP (m)Background(Dimensions: approx. 2.5m×2.0m×2.1m [height])

Horizontal Direction (m)▲ Distribution of muon-transmission{Unit 2: measured March–July 2016 (Figure 5)}●

Verified the presence of high- density material●

S ome amount of high - densit y material is also

believed to be fuel debris at the bottom of the RPV

present in the lower part and perimeter of thereactor core

{Unit 3: measured May–July 2017}●

There are no large clumps of high-density material●

There may still be some fuel debris at the bottom ofin the original core area
the RPVNorth▲ Small muon-transmission

measuring equipment (Photo 2)(Dimensions: approx. 1 m × 1 m × 1.3 m[height])

Horizontal Distance South(m)

Lower Part of RPV▲ Distribution of matter at the bottomof the RPV (Unit 2) (Figure 5)(measurements taken on July 22, 2016)

specific materials was developed.

28

Ⅲ

Decommissioning
Research and Development

Investigating and Analyzing inside PCV/RPV

Development of technology for non-destructive detection of radioactive materials accumulated in the suppression chamber

1

Migration scenarios for radioactive material
IRID has studied various scenarios for the migration of radioactiveo f t h e s a n d - c u s h i o n d ra i n p i p e ’s e x h a u s t p o r t c a n b eof that study show there to be a low chance that fuel debrisof accumulated radioactive material — to verify that there is nomaterial into the S/C and torus room (Table 1). Since the results
exceeding the maximum permissible dose has flowed into these■ Studied scenarios for the migration of radioactive material to the S/C and torus room

areas, measurements at the bottom of the S/C and in the vicinity■ Evaluated the effects of radioactive material on water seals etc.

performed—areas where there should be relatively low amounts
material exceeding the permissible dose.

▼ Scenarios with a relatively high possibility of inflowing radioactive material■ Evaluated the nuclide composition of radioactive material and the distribution of radiation; assessed background radiation;
selected the optimal methods for detecting radioactive material; and studied methods for estimating the accumulation ofmaterial inside the S/C

Form

Research and development progress

S/C

Torus Room

Melted radioactive
material

Radioactive material in the dry well reaches the S/C vent pipe
intake port and ﬂows into the S/C.

Radioactive material in the dry well corrodes the PCV shell andsand cushion then ﬂows in through the drain pipe.

Powdery radioactive
material

The inﬂow of coolant etc. causes material to migrate inthrough the S/C vent pipe.

The inﬂow of coolant etc. causes material to migrate inthrough the sand-cushion drain pipe.

Aerosol

The ﬂow of gas causes the material to migrate in through pathways
such as the SR intake pipe (primarily from the release of steam) andthe S/C vent pipe.

The ﬂow of gas causes the material to migrate in through the S/C
and the damaged vacuum break line etc.

2-1

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Estimating and developing methods of measuring
the accumulation of radioactive material within the suppression
chamber (S/C) etc.

Evaluating the nuclide composition of radioactive material and the distribution of radiation
The nuclides originating from fuel (the measurement targets (Cm-scale was used, assessing the neutron flux and γ-ray flux aroundnuclides) were evaluated according to their ORIGEN codes. The

radiation source, the neutron flux and γ ray flux at the bottom ofbackground environment.

10 -1410 -1310 -12

10 -11

10 -10Target Zone for Analysis

Flux［1/cm 2 /source］
10 -9

Flux［1/cm 2 /source］
10 -4
10 -5
10 -6
10 -7
10 -8
10 -9
10 -10
10 -11
10 -12
10 -1310 -1410 -15
10 -16a three-dimensional system that simulates those areas at a 1:16

(neutrons) and Eu-154 (γ rays) can be measured even within the

10 -8

the computational model for the S/C and torus room (Figure 1),

the S/C was then evaluated. The results confirmed that Cm-244

10 -7

was set based on the results of analyzing MAAP codes. Regarding

10 -6

mixture ratio of fuel-originating nuclides and structural materials

the bottom of the S/C (Figure 2). Setting the fuel debris as the

10 -5244, Eu-154), background nuclides, and shielding-material

S/C filled withwater

Radioactive Material in S/C(equivalent to U10kg)

Enlarged View of S/C Bottom▲Calculation system (Figure 1)

y［cm］

Calculation results for Eu-154

y［cm］

Calculation results for Cm-244▲ Distribution of fuel-originating radiation in the vicinity of Unit 1’s S/C(Figure 2)

To acquire the information needed to repair and stop leaks

be understood; however, no method for performing such an

estimating the amount of radioactive material accumulated instate of radioactive material accumulated within them mustevaluation has been developed. Non-destructive
measurement would be ideal for this evaluation, but it isin the S/C and other areas, methods for measuring andthose areas will be developed.

unclear whether the accumulated radioactive material that

There appears to be a low chance that radioactive material

destructive measurements. In addition, developing a method

torus room, and it has been determined that it is technicallyneeds to be detected can be evaluated through non-that can be used to identify the accumulated material

throughout the entire S/C etc. has become quite thechallenge.

exceeding the permissible dose has flowed into the S/C or

2

Assessing background radiation

Flux［1/cm 2 /source］

IRID has assessed the background γ radiation (Cs-134 and Cs-137) of
the accumulated water (Figure 3).

That assessment confirmed that there is a high amount of radiationz［ cm］To repair and stop leaks in the S/C and other structures, the

-2

10 -8

Purpose10 -7

Backgroundcoming from water accumulated within the torus room (the red region isabout 10 -7/cm2/source). Also, γ ray flux was observed to be roughly
uniform in both the S/C and the torus room.

feasible for non-destructive detection to confirm thepresence or absence of such material under the assumed
conditions. Moving forward, measuring systems and accessequipment will be continually manufactured based on thedevelopment of methods for making repairs and plugging

y［cm］
y［cm］

▲ Distribution of fuel-originating radiation in
the vicinity of Unit 1’s S/C (Figure 3)

leaks.

29

30

Upgrading for identifying comprehensive conditions inside the reactor

Assessing what happened during
the accident
Based on a comprehensive review by a group of experts, theelement models of MAAP and SAMPSON were improved, thenstudied and assessed accident scenarios that can rationally explain

plant data from the time of the accident. The most credible
scenarios regarding fuel debris fallen within the RPV, damage to■ Collated data from a variety of sources, including actual on-site data, simulations of how the accident unfolded, and theresults of other projects, then analyzed in detail and evaluated the internal conditions of the RPV and PCV

the RPV boundary, and migration of debris to the pedestal were

distribution characteristics of fission products (FP) for use in comprehensive analysis and assessment● What is MAAP?

Decline in
RCIC function(hypothesis)

SRV opens

Core Pressure (Measured)Fire trucks begin spraying

Core Pressure (Simulated)Core meltdown begins (analysis result)
Fuel debris falls to floor of PCV(analysis result)
SRV closes (hypothesis)

compiled at the time of project implementation.

■ Estimated and evaluated the behavior and predicted accumulation/distribution of fuel debris as well as the behavior and

The Modular Accident Analysis Program is computer code that simulates■ Utilized knowledge both from within and outside of Japan through international joint research with OECD/NEA

severe accidents. Owned by the Electric Power Research Institute (EPRI), it■ Comprehensive analysis and evaluation has continued at Tokyo Electric Power Company (TEPCO) Holdings, Inc. after theis widely used both in Japan and overseas for checking the safety ofcompletion of this project in FY 2017

nuclear power plants.

● What is SAMPSON?

Research and development progress“Severe Accident analysis code with Mechanistic, Parallelized Simulations

Oriented towards Nuclear fields” or SAMPSON is computer code that uses

PCV Pressure (Measured)
PCV Pressure (Simulated)

Leak from the gas phaseof RPV (hypothesis)

multidimensional models established from mathematical and theoreticalframeworks to illustrate physical phenomenon in detail. It is well-suited for
simulating things like the characteristics and distribution of fuel debris.

2

Studying how to reduce the uncertainty
of simulation techniques

Leak from the gas phase of
D/W 9hypothesis)

Date and Time

▲Using MAAP codes to reproduce the events of the Unit 2
accident progression

3

Assessing the chemical propertiesof fission products (FP)IRID conducted large-scale mock-up testing to obtain

Along with investigating the inside of the reactors, samples

simulations as being unique to the accident at Fukushima

and knowledge that can be used to assess things such as

experimental data on areas that were identified throughDaiichi NPS (the transparency of steam within the reactor

as the accident unfolded; debris infiltrating inside thecontrol-rod drive mechanism in the lower section of the
RPV; the mixing of metallic debris and oxide debris at thebottom of the RPV; etc.). The factor models of MAAP and

SAMPSON were enhanced and had a group of experts
conduct a comprehensive review.

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Estimating the conditions inside the RPV and PCV,
which is essential for studying technology for fuel debris
retrieval

1Core Pressure (MPak(abs))

Investigating and Analyzing inside PCV/RPVPCV Pressure (MPak(abs))

Ⅲ

Decommissioning
Research and Development

collected from sediment near the fuel debris were analyzed
how the accident progression process and the status ofaccumulated fuel debris were compiled. The photo below

is an example of the many particles containing uranium that
we re d ete c te d within a s ample colle c te d from theprotective sheet covering the operating floor of Unit 2.

While that analysis and detailed assessment are currently

ongoing, early results have shown that these uranium
particles were theoretically created in one of two ways: bytransitioning to a gas then condensing, or by transitioning
to a liquid and congealing. In the latter case, compositional

BackgroundEstimating the conditions inside the RPV and PCV as

In an effort to ensure that the decommissioning of Fukushimameasures and studying methods for the retrieval of fuelestimate the conditions within the RPV and PCV through aaccurately as possible is essential to developing safety

debris that will commence in the near future. However, theextremely radioactive environments in Units 1–3 at
Fukushima Daiichi NPS make direct investigations andobservation difficult. Remote-operated equipment has beenused to investigate some areas, but it is still exceedingly

difficult to gain a sufficient understanding of the conditions
inside the reactors.

information about the fuel debris itself may have been

Purposepreserved. In addition, studies are underway to determine

whether this knowledge could be applied to estimating thepath by which the uranium particles migrated from the RPV
to the operating floor during the accident.

Daiichi NPS proceeds as planned, this project aimed tocomprehensive analysis and evaluation of various sources of

information, including on-site data (internal observations andknowledge obtained from decommissioning work conductedaround the reactor building), plant data from the time of theaccident (water levels, temperature, pressure, radiation dose,etc.), simulations of how the accident unfolded (including
simulations conducted by this project through improvements

to MAAP and SAMPSON), past knowledge gained fromevents like the TMI-2 incident and mock-up testing on severe

accidents, and insights gleaned from mock-up testing
conducted primarily in Japan following the accident at

Fukushima Daiichi NPS, including the mock-up testing done

during this project. This project was conducted under a joint▲ Plasma-heating test on
simulated fuel assembly(during heating and thetest piece after)▲Test apparatus▲The protective sheet, an analysis target obtained from thefifth floor of the reactor building of Unit 2

proposal with the Institute of Applied Energy.

3132

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

Based on a comprehensive evaluation by a group of experts, aorganized into several categories depending on its characteristics,

accumulated debris in each reactor unit as well as its distribution

stated. This fundamental knowledge was then used to makedetailed diagram was created depicting the estimated locations ofpatterns and information about damage to the RPV and PCV. When

conducting this evaluation, the debris and damaged material were

and the basis for those determinations when listing the objects were
decisions about methods of fuel debris retrieval.

Estimated Distribution of Fuel Debris and Other Conditions in the RPV and PCV of Unit 2
Rev. 15 (March 29, 2019), These are the results of continued analysis and evaluation by Tokyo Electric Power Company (TEPCO) Holdings,
Inc. even after the project ended in FY2017.
・ After estimating the amount of energy based on the increase in PCV
pressure due to hydrogen generation, it is assumed that most of the fuelis damaged or has collapsed. (measurement and analysis).

・Since upper tie plates have fallen around the pedestal's
perimeter, there may be per forations in the RPV there .
(assumption based on measurement)

・If fuel rods are present, some of them are around the perimeter (generalassumption)

・Parts of the control-rod guide mechanism and tube in theperimeter may have been melted or destroyed by fuel debris
that accumulated in the bottom of the RPV. (assumption basedon measurement)

・There is typical oxide debris formed from molten fuel that has hardened(general assumption).

・During an inspection of the PCV's interior, the control-rod drive
mechanism was observed. And since the grating is partially missing, there
are presumably (small) holes in and around the center portion of the
RPV. (assumption based on measurement)
・Although strength varies depending on the location, water droplets
falling all over the pedestal floor has been observed. This means that
there may be small perforations in the vicinity of the housing for thecontrol-rod drive mechanism at the bottom of the RPV. (assumption
based on measurement)

・Some of the debris that fell through the holes is presumably stuck tot hin g s like th e co ntro l - ro d d rive m e ch a ni s m h o u si n g. ( ge n e ral
assumption)

・The upper tie plates etc. fell to the floor of the pedestal. The
sediment in that area, which is believed to have fallen through thesame hole in the RPV is assumed to be fuel debris. (measurement)

・Sediment of various sizes, such as pebbles and rocks, wasobserved on the pedestal floor. Since pooled water was founda ro u n d t h e ro c k- s i ze d s e d i m e n t , s u c h s e d i m e n t m ay b eimpermeable. Pooled water was also observed here and therearound the pebble-sized sediment, indicating that rock-sizeds e dim e nt may lay und e rneath p e bble - like s e dim e nt .
(measurement)・Spray from the reactor's sprinkler system lowered the temperature,and increasing the flow rate caused the water level outside theshroud to rise. Therefore, there is presumably no large-scale
damage to the shroud. (measurement)

・The temperature of fuel on the perimeter has not increased
significantly, so there may be pellets and broken fuel rods
remaining on the perimeter. (general assumption, testing, andanalysis)・If heat transfer from the high-temperature fuel debris is low, thecontrol-rod guide tube will not melt. (general assumption)

・Temperature decreased when the reactor was sprayed with water, so it ispredicted that there is fuel present at the perimeter of the reactor inareas struck by the water sprayed from the low-volume sprinkler system.
(Even if molten fuel falls then solidifies on the fuel supports or controlrod guide tube, it behaves similarly as a heat source, so the exact
positioning of debris cannot be determined.) (measurement)

・Measurements with muons indicate that fuel may be present at theperimeter of the reactor. (measurement)

・Pressure inside the reactor's recycling system rose when theamount of water flowing into the reactor increased, suggestingthat there may be some level of water outside the shroud.
(measurement)

・The holes are presumably as least large enough for the upper
tie plates to fall through. (measurement)

・If particulate debris and pellets are present, they may haveaccumulated in the sediment layer. (general assumption)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

4

Comprehensive Analysis and Evaluation of Reactor Pressure Vessel (RPV) and PrimaryContainment Vessel (PCV) Interior

・The shell of the PCV does not appear to be damaged (no leakage
from the sand-cushion drainpipe), so there was presumably little
molten-core–concrete interaction (MCCI). (measurement)

・Muon measurements revealed the silhouette of a high-density
substance believed to be fuel debris in the bottom of the RPV.
Fuel that fell onto the lower plenum may still remain at thebottom of the RPV. (measurement)

・Since the pedestal walls, cable ties, supports for the control-rod
drive mechanism, and other components are intact and unmelted,
MCCI may be limited. (measurement and general assumption)

・Fuel debris may have solidified without undergoing MCCI.
(assumption and analysis based on measurement)
・Fuel debris that did undergo MCCI may have mixed with theconcrete. (general assumption)

・As a result of damage to the control-rod guide tube and thecontrol-rod guide housing, fuel debris and molten metal may
have penetrated to the housing's interior.

・Enhanced images reveal a large amount of water droplets
falling on the left side of the platform, as viewed from the CRD
transfer rail. In addition, the positions of the PIP cable andother objects in that area cannot be verified, and grating is
missing from the platform. Therefore, water that flooded thereactor may be dripping through perforations in the RPV.
(measurement)

・Radiation dose and temperature were measured from the platform to thebottom of the pedestal. While radiation increased somewhat toward thebottom, there was generally no significant variation in temperature (a slightdecrease in temperature was observed). And since no obvious damage to thepedestal substructure was found, the fuel debris on the floor of the pedestal
is believed to be generating a relatively low amount of radiation and decay
heat, and it may contain large quantities of metal. (assumption based on
measurement)
・Sediment that contains fuel debris is presumably scattered across the entire
pedestal floor (measurement).

・If water has pooled on the bottom of the PCV, then particulate debris haspresumably formed. (general assumption)

Remainingfuel rods androd fragments

Control-rod
drive mechanism

Oxide debris(porous)Control-rod drive
mechanism(containing debris)

Particulate debris

ShroudFuel debris(containing large
quantities of metal)

PelletMixed-concrete
debris

RPV perforation

Control-rod
guide tube

Upper tie plate

Damagedcontrol-rod
guide tube

Sediment(unknown material)・If particulate debris is present, it may have accumulated in the sediment layer.
(general assumption)

Provided by TEPCO Holdings, Inc.

33

34

Ⅲ

Decommissioning
Research and Development

Investigating and Analyzing inside PCV/RPV

Development of analysis and detection technology for characterization of fuel debris

Currently, there is little data on the characteristics of fuel debris—data which would serve useful in selecting methods for
retrieving fuel debris from Fukushima Daiichi NPS (Hereinafter “1F”) as well as transporting and storing it. Therefore, IRID isdeveloping analysis and estimation technologies based on the following three points to estimate these characteristics.
The most important results thus far are discussed below.
■ Development of technologies for estimating the properties of fuel debris existing in the PCV/RPV from test data by preparing
domestic and international knowledge and simulated fuel debris samples■ Development of analytical techniques for fuel debris using simulated fuel debris samples and acquisition of attributes
necessary for fuel debris management■ Acquisition of analytical data of sediment and adhered materials collected from 1F, and creation of a database for estimationof fuel debris properties based on the evaluation of the data.

1

estimate its characteristics

The fuel debris created during the accident at Three Mile Island’s

At the National Nuclear Center in Kazakhstan, UO2 was used tostructures in the reactors. However, the fuel debris from the 1F

that has melted and resolidified (Figure 2). In the test at theUnit 2 (“TMI-2”) consisted primarily of nuclear fuel and internal

accident is unique, due to its interactions with concrete.

Information about the product of this molten core-concreteinteraction (MCCI) is therefore needed.

Figure 1 depicts a test conducted using condensed-light heating.

A method for creating simulated MCCI product was created byplacing simulated corium (a disk made from Zr, which is an

test the large-scale production of a metallic/ceramic substance
NNC, about 10 kg of UO2, Zr, and B4C was heated and melted

then made to fall onto a stainless-steel bowl, thus producing thesimulated fuel debris. Through this, detailed information was

obtained about generating simulated fuel debris that has agranular or clumpy surface; the characteristics of a metallic/ceramic mixture; and related boundary regions.

element in fuel cladding etc., and stainless steel (SUS), which isused in other component material) on top of concrete fragments

then melting it with condensed-light heating. This allowed use ofdebris with different characteristics as test parameters whenobtaining data with IRID's simulated MCCI fuel debris.

(1)Rock-like (annealed)
Fuel Debris

Research and development progress

Solid rock-like
form

50mm

Section of melted
stainless steel

100mm(2)Powdery / Granular(water-cooled)

Fixed Ring
Fuel Debris

Powdery andgranular form

Stainless Steel

Before Heating

After Heating▲ Creating simulated MCCI product via condensed-light heating(Figure 1)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Understanding the characteristics of fuel debris, which areessential for studying technology for fuel debris retrieval

Utilize domestic and international knowledge to create simulated fuel debris and

50mm

100mm

▲ Test production of melted and hardened metallic/ceramic mixture at the NNC in Kazakhstan (Figure 2)

Large-scale MCCI test in FY2016

Background
There is almost no knowledge about the fuel debris formed

To develop analytic and other technology, IRID will create

during the melt-down accident of Fukushima Daiichi, so it isbased on similar examples found in domestic and international

globally unprecedented task of the decommissioning of amethods, the results of that verification through the analyticalimages of some areas have been obtained, information is stilland equipment used to conduct actual investigations of thefrom the melted nuclear fuel with reactor structural materials

simulated fuel debris that simulates the form of debris in 1F,

necessary to uncover such knowledge and to achieve theresearch. Then, upon verifying the suitability of the analysis

damaged reactor caused by an accident. While on-site visual

process will be disclosed. Material that adheres to the tools

limited. Fuel debris is a heterogeneous material that contains

inside of the 1F PCV will be analyzed, then the suitability ofcomplex mixture which also includes some hard-to-melt

analysis results, hardness and other physical characteristics fora great variety of radioactive elements; it is assumed to be acomponents. The needed information must be extracted anduse in selecting tools for processing fuel debris will be

estimated; water content, which is needed to assess hydrogenradioactive material and the accuracy of the analysis methods

elements due to heat applied for drying purposes. In addition,

employed. This will prove useful for enacting radiation safety

policies and developing the tools and equipment needed forfuel debris retrieval as well as the equipment used totransport and store it after it is retrieved.

Concrete boundary

Condition before dismantling(after removal of the upper
cylinder)Glass;Orange-colored layer

Metal particles

the developed analytic technology will be verified. From thoseorganized and acquire test data related to the many tasks
that decommissioning entails, such as the handling of highly

35

Purposegeneration during storage; and the volatilization of radioactive

there will be an enhancement of IRID’s ability to predict the

behavior of the fine particles that will presumably begenerated when processing fuel debris. Those results will be

provided in a database that is easy for people involved in

Metal layer on the bottom

Cross-sectional view of the product▲ Large-scale MCCI test: appearance of the results (Figure 3)

After removing most meltedmaterial

Meanwhile, at the French Alternative Energies and Atomic Energy

deconstructing the resulting simulated fuel debris and taking

production test that took into account the melted components atthe depth of corrosion in concrete and the characteristics ofCommis sio n (CE A ), IRID co nd uc te d a large -s c ale M CCI

1F and the type of concrete used in Japan (Figure 3). By

samples from its various sections, information about things like
MCCI product was obtained.

decommissioning work to use.

36

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

previous basic tests were reaffirmed, with no major changes

uniform composition with porous sections and areas with metallic

and the vapor pressure calculated for the fission products (FP)each structural region (Figure 4). The analysis revealed a non-detected. Figure 5 depicts the relationship between temperature

properties. As for the layers formed from oxides, it became
apparent that substances such as U-Zr oxides form a precipitatefound that the debris collected from the crust section ispressure curves than UO3—in addition to the highly volatile Cs.

conducted by the Japan Atomic Energy Agency (JAEA) (Figure

debris samples collected from three different locations were also

volatilized.

composed primarily of iron. Conditions that were estimated from

－Studying protocols for analyzing fuel debris as well as those used at TMI-2－
As part of IRID's efforts to develop technology for analyzing fuelThis suggests that fission products within the fuel debris may have

It was confirmed that the metallic layer at the bottom is an alloythat affect the retrieval and management of fuel debris

and actinide oxides that were presumably generated. This implies

the presence of Te, Sn, and Cd — which have higher vapor-in the parent phase, when silicon content is high.

2

debris, a cutting test on the TMI-2 fuel debris stored wascomprised primarily of oxides that contain U and Zr. TMI-2 fuel7). In addition to making metallographic observations andanalyzed (Figure 8). The sample from the lower head contained

performing a Vickers hardness test, a severed piece of test fuelU, Zr, Fe, and Cr oxides in the parent phase, while the grain-

was observed using a scanning electron microscope (SEM) and it

boundary sample contained deposited oxides of the iron family/

was analyzed with an electron probe microanalyzer (EPMA). The

iron and similar elements; the same degree of hardness wassevered fuel is displayed at the top of Figure 7, with the results ofobserved in both the upper and lower sections.

the EPMA analysis shown at the bottom of the figure. It was
Upper CrustSide
Boundary
Lower Boundary(metallic)

Upper Oxide
Layer
Lower Oxide
Layer

Element Map ofUpper Crust Cross-Section

Cs

Loose Debris (upper core)

TeO2
SnO

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

IRID also analyzed the product to obtain an element map for- 1 Establishing methods for analyzing fuel debris and understanding the characteristics

Upper Crust

UO3

Molten Pool
(two types)

CdO
Loose Debris (lower head) Hard Debris (lower head)
Element Map ofLower Boundary Cross-Section

Element Map ofLower Oxide Layer Cross-Section

Upper Crust Sample▲ Distribution of elements in large-scale MCCI products (Figure 4)

Grain boundary
precipitates▲Calculated results for oxide vapor pressure (Figure 5)

Parentphase

Gas cavity

Steady-state melted②① area inside sedimentArc-melting a mixture ofmaterials

Fe-Cr-Ni-Zr-U-(Si-Mo-B-C) Molten Material
U--Zr-O Molten Material

Reactor Components

melting interface

Localized condensed-light heating(temperature gradient) ofcore material atop concrete①

Concrete
Components
Completely
molten mixture~2,800 K

Molten Concrete Material(in the form of molten glass)

Pyrolysis Gas(primarily H20)

Heterogeneous
molten material
1,500 K-1,800 K
Thermaldegradation
1,500 K

Illustration of melted core sediment on top of concrete
Acquire data on
the generation phase,composition,and hardness of homogeneous
melted▲ Producing
two sections
types of simulated MCCI product (Figure 6)

200µm

Zr

20µm

SEM image of hard debris (lower head)

EPMA analysis: U-Zr oxide▲Cutting test on TMI-2 debris (Figure 7)▲ Appearance of TMI-2 fuel debris and SEM image of cut section (Figure 8)

Nondestructive analysis
Visual observation, weight measurement, radiation measurement(IP measurement andα/γ rays spectrometry),X-rays-CT , etc.

Sample collection
Equipment analysis
SEM* (EDX), TEM* (EDX), FE-SE/WDX,EPMA, hardness measurement, etc.

Acquire data on layered structure
under temperature gradientChemical analysis
Pretreatment, melting, chemical separation

Elemental analysis / Isotope analysis
ICP-MS*, etc.

Radionuclide analysis
α/ γ rays spectroscopy etc.

*SEM: Scanning Electron Microscope*TEM: Transmission Electron Microscope*Inductively Coupled Plasma Mass Spectrometry (ICP-MS)▲ Analysis procedures (Figure 9)

To obtain data on the little-known attributes of the MCCIand fission-product elements within the MCCI product bypurposes must be created (Figure 6). The left side of Figure 6

done to acquire data on the heterogeneous, layered structure

The Okuma Analysis and Research Center is planne d toscheme it was assumed the analysis of a very small amount of test

that contains potentially meltable FP elements etc. was compiled

development of the technology for that analysis, a case study on

updated in accordance with the actual retrieval process and the

product, simulated MCCI fuel debris that is suitable for testing
depicts the test in which information about composition,

hardness, etc. was acquired by using arc melting to produce a
uniform molten substance. The right side of the figure shows thetest that was conducted to obtain data on the behavior of Gd

37

U

Parent-phase hardness of
12‒15 GPa
Same as upper crustTransport sample for analysis②

Concrete
Thermally Degraded
Concrete (oxide)

Hardened Surface

Layered structure of

Vickers hardness
impression

applying condensed-light heating in a localized manner. This wasunder a temperature gradient. Attribute data on oxide fuel debris

spearhead analysis of the fuel debris from 1F. As part of IRID'sand added it to the Debris Wiki, which will be discussed later.

the analysis process was conducted (Figure 9). Since the plan is
to gradually increase the scope of fuel debris retrieval, for thisfuel; however, it is believed that the flow of analysis will be
policies for storing and managing fuel debris.

38

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Developmentthat affect the retrieval and management of fuel debris－Obtaining data for analyzing the chemical composition of fuel debris
and assessing the generation of hydrogen during storage－

2- 3 Establishing methods for analyzing fuel debris and understanding the characteristics
that affect the retrieval and management of fuel debris－Obtaining data on the generation of particles and evaluating their behavior－
The dispersion of radioactive particles during the retrieval of fuelbelieved to be the result of the different vapor pressures for eachperforming chemical analysis on a dissolved solution is therelationship between the processes used to retrieve fuel debris

generated during the heating test could be affected by oxygen inemission spectroscopy. If the fuel debris contains large amounts

piece, simulated fuel debris was used, which contained U or Hfother elements, so the impact on factors such as the minimumthat was conducted previously in France. A mechanical cutting

depicts the interference pattern for the intensity of light emitted

was performed and the generation and behavior of particles was

IRID studied methods of melting fuel debris as a preparatory step

another method was summarized. One of the issues withdebris is a challenge. IRID has conducted testing to explore thewhen internal core structures and concrete interact under high-interference that occurs with inductively coupled plasma atomic

and the radioactive particles that are generated. For the testof uranium, there will be interference when trying to analyze

that was included in the material obtained during the MCCI test

Of the processes for retrieving fuel debris, the method that

detection limit of the target element was evaluated. Figure 11

test and a heating test tailored to the assumed retrieval process

to produce particles that either travel through the water and are

by U and Pu and how it affects the analysis of Si. Emission peaks

evaluated.

wavelength for Si—which means interference for the Si analysis.

Figure 13 indicates the diameter distribution of generated

for analyzing its elemental composition. Fuel debris is createdtemperature conditions, so it is presumably difficult to dissolve inwater. Therefore, the alkali dissolution method was employed,
which is known for its ability to dissolve insoluble material. Using

an electric furnace, the simulated fuel debris reacted with sodium

peroxide at 850 ℃ , after which the material was able to becompletely dissolved using concentrated nitric acid (Figure 10).

The challenges with applying analytic technology to insoluble

fuel debris that has been dissolved via alkali dissolution or

for U and Pu are present near 288.16 nm — the analysis

component within the test piece. It also suggests that particles
the atmosphere.

involves applying heat under water (the laser method) is believed
released at the surface of the gas or are transported through thewater while in the liquid phase. To investigate things such as thephase change of particles from gas to liquid and how particles
migrate while in the liquid phase, simulations using CFD codes

particles. Compared to the mechanical cutting shown on the left

were conducted in addition to basic testing that simulates particle

graph, the average diameter of particles generated during the

behavior (Figure 14) as well as a precipitation test (Figure 15).

heating test was smaller, as is depicted on the right graph. These

generated particles are composed mainly of U and Zr, which is

Applying the data from the aforementioned heating test to thedifferent than the composition of the simulated fuel debris in thesimulations, the gas-to-liquid migration rate was estimated for theparticles during heating is evaporation and condensation, this is

generated by the heating test.

▲ Interference of U and Pu on the elemental analysis of Si (Figure 11)

CT_#4EV_25_1

simulated 1F fuel debris particles that were predicted to be

Mass distribution percentage(%)▲ Alkali dissolution testing on simulated fuel debris (Figure 10)

Mass distribution percentage(%)

heating test. Since the primary mechanism that generates

CT_#4EV_25_2
CT_#8EV_25_1
CT_#8EV_15_1

CT_#8EV_Metal_25_2

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

2

- 2 Establishing methods for analyzing fuel debris and understanding the characteristics

HT_#2_Hf-EV_T1_1
HT_#1_U-MCCI_T1_1

Aerodynamic diameter(μm)

Aerodynamic diameter(μm)▲Test results on the generation and behavior of particles using simulated fuel debris that contains uranium from the testing in France (Figure 13)

ZrO2 Precipitation Test (average particle diameter = 0.712 μm)① After 15 hours of heat treatment at 300 ℃ in a low-oxygen
environment (no cracks have formed)② After 11 hours of heat treatment at 600 ℃ in a low-oxygen
environment (cracks are visible)

One of the goals of understanding fuel debris characteristics is toof 300 ℃ , also taking into account information about theto the storage and management of fuel debris after it has been

In addition to the drying characteristics of concrete, which is one

obtain attribute data that can be used to evaluate safety related
retrieved. One of the challenges of highly radioactive fuel debris
is that hydrogen can be generated through the radiolysis of thematerial’s water content.

IRID thus obtained data on the drying properties of fuel debris
when water is removed for the purpose of preventing hydrogengeneration. For this, IRID focused on the estimated temperature

39

Particle Precipitation Rate (%)▲ Appearance of simulated fuel debris after heat treatment (Figure 12)

behavior of volatile radionuclides that was obtained in this study.

of the components of MCCI product, the drying behavior ofvarious particle sizes was also evaluated, under the assumptionthat particulate fuel debris is hard to dry. As seen in Figure 12’s

images, there was no change in the surface condition of fuelPure water
Acidic environment（pH4）
Alkaline environment（pH10）
Sodium sulfate solution（0.1mol/L）debris dried at 300 ℃ , so this could be a good guideline as far
as temperature is concerned.

Time (h)▲Testing particle migration at the gas-liquidboundary (Figure 14)▲Changes in the ZrO2 particle precipitation rate over time
for different dissolved substances (Figure 15)40

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

A variety of equipment have been used to investigate the insideof the primary containment vessel at 1F. To augment IRID's

that will presumably generate those particles.

equipment as well as the sediment that they collected.

originate from fuel was detected in the sample collected from

Oarai Research and Development Institute. At other analysis

agencies, a compound analysis with a transmission electron
microscope (TEM) was performed. Figure 17 depicts how theexistence of uranium-containing particles in the sample wereverified, and the evaluation of characteristics of the fuel debris

investigations of the pedestal. The photographs in the center ofElectric Power Company Holdings Inc. (TEPCO). All of the data

Figure 19 show the outward appearance of those samples. IRID

Collection locations for the Unit 2 samples as well as an analysis

Figure 16 depicts the SEM analysis taking place at the JAEA

conditions within the reactors) and the results of internalmaterial adhered on the surface of robots returning from internal

analytic data and experience, efforts were coordinated withmultiple agencies that sampled the material accreted on those

IRID used smear tabs and cotton swabs to collect and analyze

example are shown in Figure 18. Uranium that is believed to

conducted an analysis of elements such as U and Fe thenevaluated that data’s connection to not only analysis methods

investigations and accident simulations conducted by Tokyo

was compiled as fuel debris characteristics for each area of Units
1–3.

but also the results of prior projects (such as studies to ascertain

the operating floor inside the reactor (2u-1) and the sedimentsmear (2u -5). Als o dete c te d were Zr, which pre sumably

originated from fuel cladding and similar material; Zn, which isbelieved to be from paint; and Pb, which is thought to be fromshielding materials. Based on the presence or absence of eachelement at the various locations, IRID estimated the state ofconditions in the reactor.

Particles containing
uranium

Particles containing
Ag-Te-Pb

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

3

Estimating fuel debris characteristics through the analysis of
sediment then providing them in a database

Unit 3 Pedestal▲ Result of analyzing adhered material (Figure 19)

The fuel debris data collected by this project covers a lot of

To make the data more accessible, a Wiki-style databasematerials, analysis results of fuel debris from TMI-2 etc.,

links to other information such as sample-collection location,areas, including the various physical properties of fuel and otheroperational information about 1F, the results of internal reactor

investigations, and the results of sample analysis. Therefore, a▲Conducting a SEM* analysis on the sample of accreted material
collected inside the reactor (Figure 16)
At the JAEA Oarai Research and Development Institute*SEM: scanning electron microscope▲TEM* analysis of fine-powder sample collected from within
the SGTS intake pipes of Units 1 and 2 (confirmation of

database has been created in conjunction with the JAEA andTEPCO so that the data can be utilized in an organized manner.

platform has been created, that systematically organizes it withaccident information for each reactor, and the results of internal
reactor surveys. IRID's plan is to continue inputting the dataacquired by IRID thus far and integrating future analysis data andother information into the database.

particles containing uranium) (Figure 17)*TEM: transmission electron microscope

https://fdada-plus.info/wiki/index.php

SEI

5μm

U

5μm

Cr

5μm

Fe

5μm

5μm

Zr

5μm

Sampleサンプル外観（ろ紙）(filter paper)

Unit 2

イメージングプレート像
Imaging Plate

Zn

SEM-WDX像
SEM-WDX
Image▲Collection locations for Unit 2 samples and analysis example (Figure 18)▲ Example of the database (Figure 20)

41

42

Ⅲ

Decommissioning
Research and Development

Investigating and Analyzing inside PCV/RPV

Research anddevelopment
progress
Developmentof technology
for detailed
investigation inside the primary containment vessel (PCV)【Unit 1】

【Unit 3】
Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Investigating inside the PCV to obtain detailed information
for fuel debris retrieval■ Detailed information about the PCV interior is needed for future fuel debris retrieval■ IRID is developing equipment and conducting detailed investigations in the PCV■ Equipment for the trial retrieval of fuel debris is being developed■ Equipment for creating access routes to the PCV interior is also under development【Units 1‒3】

【Unit 3】

Research and development progress【Unit 1】

1

X-100B Pene: X-100B Penetration

2

X-2 Pene: X-2 Penetration【Unit 2】

【Units 1‒3】
BackgroundTo establish methods for retrieving fuel debris, it is necessary
1

X-100B Pene: X-100B Penetration

2

X-2 Pene: X-2 Penetration【Unit 2】

3

X-53 Pene: X-53 Penetration

4

X-6 Pene: X-6 Penetration

Investigation of the inside PCVs have been conducted on

In the case of Unit 2, the existence of pebbles and clay

debris within the PCV and the status of the structures.
Units 1, 2, and 3.

sediment throughout the entire bottom of the pedestal’s

While these surveys provided valuable information, they

understanding of the state of fuel debris and sediment as

and the size of existing access points. In addition to updating

attached structures, a larger-diameter hole (approximately

development, there are the additional challenges of creating

below) then a suitable combination of equipment andinvestigation equipment.

conduct a detailed investigation of the pedestal's interior

well as the shape and dimensions of the pedestal and its

a n d f i n a l i z i n g p l a n s f o r f u r t h e r i n ve s t i g a t i o n a n d

550 mm) will be created in the X-6 penetration (“X-6”

a c c e s s r o u t e s , a n d ve r i f y i g d e t a i l e d d e s i g n o f t h etechnology will be used to access the PCV interior andexperimental retrieval equipment.

Unit 3 is believed to have high water levels within its PCV

the pedestal via openings in the basement floor. To prepare

to these factors, a small submersible robot (ROV) has been

general situation on the basement floor. Based on thoseresults, a plan was developed to conduct investigation andobtain information about the fuel debris inside and outside

the pedestal and the status of accumulated sediments. In

43and exterior. The aim is to collect some fuel debris using

It is believed that virtually all the melted fuel in Unit 1 fellpedestal on the first floor was investigated as well as theX-6 Pene: X-6 Penetrationinterior has been confirmed. To gain a more accurate

were limited by factors such as the radioactive environment

for investigating this fuel debris, the environment outside the4

explored. At the same time, equipment to investigate theconditions inside the pedestal is underway.

into the pedestal, with some of that fuel spreading outside

X-53 Pene: X-53 Penetration

existence of fuel debris and its characteristics will beto understand the distribution and characteristics of fuelPurpose3

sediments in and around the pedestal as well as thethat investigation, the spread and depth of accumulated

along with a variety of obstructions inside its pedestal. Duedeveloped that can access confined areas within thepedestal and navigate underwater while avoiding obstacles.

After a mock-up test, it was used to investigate the inside of
the pedestal. Fallen objects and damage to structures wereobserved throughout nearly the entire area, so a 3-D map isbeing developed for use in studying fuel debris retrieval
policies.

44

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

1

3

Investigation for the basement floor and the first floor of outside the pedestal

Detailed investigation for the basement floor of inside/outside of the pedestal(creating an access route)
To investigate the inside of the PCV in detail, the investigation

inner door of the air lock, the grating, and other surfaces with an

Unit 1. Technology was developed for creating opening,

All operations were monitored remotely and maintained the PCV

equipment was entered through the X-2 penetration (air lock) of

IRID developed the PMORPH1, which enters the PCV via aestablishing boundaries, and remote monitoring in order to create

pipe with an internal diameter of 100 mm then transforms

so that it can travel reliably across the top of the grating toTemperature Sensor

investigate the area (Figure 1-1). By investigating the firstfloor grating in the PCV of Unit 1, it was learned thatthat access route (Figure 3-1). The opening technology included

abrasive water jet (AWJ) (Figures 3-3 and 3-4).
boundaries.

equipment for making a hole in the exterior door of the air lockRadiation Sensor

with a diamond cutter (Figure 3-2) and for making a hole in thefactors such as the degree of equipment damage and thedose-rate distribution. Based on the above equipment, theWinch

Crawler

PMORPH2 was developed, which is capable of lowering
an array of sensors (a camera and dose rate meter) throughRadiation Sensor (internal)

New Boundary

Investigation Camera

the cracks in the grating to a depth of approximately threemeters (Figure 1-2). Through the investigation of the PCV

basement floor in Unit 1, the level of accumulated water

LED▲ PMORPH1 for investigating

the first-floor pedestal exterior

was verified and the presence of sediment up to one meter

Measurement
Camera▲ PMORPH2 for investigating

the subfloor outside the pedestal(Figure 1-1)

(Figure 1-2)

in size was confirmed.

P/A Room
Isolation Area

2

Detailed investigation of inside/outside of the first-floor pedestal
To investigate the inside of the PCV pedestal in Unit 1, IRID isdeveloping equipment that will enter through an existing opening
and move to the top of the grating, then access the pedestal’sinterior through the CRD opening to investigate the area by
capturing images and obtaining radiation dose information(Figures 2-1, 2-2, and 2-3). The investigation was conducted instages: first the primary investigation, which explored theconditions from in front of the CRD opening to outside thepedestal, then the secondary investigation, which investigated
the above-water sections inside the pedestal.

Area preparation and obstacle removal↓
Connect to new boundary↓
Creating opening in outer door of airlock↓
P/A Room
Creating opening in inner door of airlock,
AWJ Equipmentﬁrst-ﬂoor grating, etc.
P/A Room↓
Stopper Cap
Insert guide pipe↓
Conduct a detailed survey↓
On-site demonstration complete

For the primary investigation, a roughly 70-cm extendable rod

was attached to the crawler equipment then operated it while
performing measurements; in the secondary investigation, theIsolation Area
Forward Motorrod was extended to approximately 5 meters then gathered data

Drill

P/A Room
Isolation Area

Guide Pipe

Insertion Tool

P/A Room

Supports

Cable Drum

ToolboxCreating opening in outer
door with 350A

Image from monitoring
camerawith the camera and radiation sensors on its tip. Support
equipment such as a cable-feeding equipment that grips, moves,
and feeds out cable for the investigation equipment andmonitors conditions behind it, is being developed.

Shaft
Rotational Motor

Core Bit

Isolation Area

▲Creating opening in outer door of airlock with core opening equipment (Figure 3-2)

Extendable Rod（5m）Camera (withpan/tilt
mechanisms)

AWJ

Inner
Door

During Cutting▲ Survey equipment (with rod collapsed) (Figure 2-1)

▲ Investigation equipment (with rod extended) (Figure 2-2)

Creating openingin
inner door with 350A

In the airlock

After
Cutting

Opening
開口部▲Creating opening in the inner door of the airlock (Figure 3-3)

Pedestal CRD
Opening

Composite Cable

Camera Image
Inside Pedestal
Cable-Feeding equipment▲Validation test for continuous operation (Figure 2-3）

AWJ Head

Investigation instrument

Crawler (flipper-style)45

Isolation Area

▲ Steps to create the access route (Figure 3-1)

Extendable Rod

Extendable
Rod

P/A Room

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Unit 1

Spraying thewater jet

Investigationequipment▲ Releasing cable from the cable-feeding equipment (Figure 2-4)

AWJ Head

Protruding lead wool
▲Creating opening in the grating (Figure 3-4)

46

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

Detailed investigation for the basement floor of inside/outside of the pedestal
Since water has accumulated in the lower section of the PCVsuch as the ability to create 3-D sediment maps, measure therobot is needed to investigate those areas. Therefore, IRID

detectors or other methods, or collect samples of sediment. The

and the basement floor of the reactor building, a submersible
developed the submersible IRIDOLPHIN (Figure 4-1) toinvestigate the basement flor inside and outside the pedestal.

Five types of submersible boat (Figure 4-2) and a single type ofsmall submersible boat (Figure 4-3) were developed. One of thefirst five types can attach magnetic guide rings to the basement

thickness of sediment, measure fuel debris using neutron

1

small boat will submerge to enter the pedestal and survey its

interior. In FY 2021, guide rings were attached along the routethe boats will take (Figure 4-4); moving forward, IRID will

IRIDOLPHIN-A

continue to conduct actual investigations with the other boats aswell.

with guide rings

IRIDOLPHIN-A2

Detailed visual observation

IRIDOLPHIN-B

3-D mapping of sediment

IRIDOLPHIN-C

Measures the thicknessof accumulated material

IRIDOLPHIN-D

Detects accumulated fueldebrisIRIDOLPHIN-E

Samples accumulatedmaterial

robot is long and straight so that it can pass through a guideWith this robot and a camera attached to a telescoping guide

unfolds like a scorpion to present its rear camera, which it uses in

capturing images of it and the CRD rail.

conjunction with the front-facing camera to achieve strong spatial(over 1,000 Gy [cumulative]) and improved visibility in hazy

To protect the ROV(ﬁber-optic γ dosimeter andimproved small-scale B10 detector)

Accesses the pedestalʼ s interior and conducts a visual inspectionof the CRD housing, fuel debris, accumulated material, etc.

・ Scanning ultrasonic range ﬁnder
・ Water temperature gauge

Uses a scanning ultrasonic range ﬁnder to check the height distributionof accumulated material・ Powerful ultrasonic sensor
・ Water temperature gauge

Uses a powerful ultrasonic sensor to measure the thickness of accumulatedmaterial and the state of the objects underneath for the purposeof estimating the height and distribution of fuel debris

・ CdTe semiconductor detector
・ Improved small-scale B10 detector

Drops a fuel debris detection sensor onto the surface ofaccumulated material and veriﬁes the composition of fuel debris
by analyzing nuclides and measuring neutron ﬂux

・ Suction sampler

Drops a sampling equipment onto the surface of accumulated materialand collects samples from that surface

・Dimensions of IRIDOLPHIN-A through E: φ25 cm × approx. 110 cm long; Dimensions of IRIDOLPHIN-A2: φ20 cm × approx. 45 cm long

▲ List of IRIDOLPHIN submersible boats (Figure 4-1)

Traversing a narrow
space

Conducting
investigaiton

Enter the pedestal via the CRD rail
Additional Lighting

Rear Monitoring
CameraMissing Grating

Rear camera
and lighting

Fallen Object

LED
Lighting

Guide Ring Installer▲ Internal pedestal images after processing (Figure 1-2)

▲ Remote-operated crawler investigation robot (Figure 1-1)

2

Investigating and studying sediment in the pedestal’s lower section(FY 2017–2018)
IRID's investigations revealed that the original plan of operating

via an on-site demonstration test that captured images inside theabove) would be difficult and that there is a hole where sectionsin that environment (Figure 2-2). In February 2019, the TEPCO

investigation equipment with a telescoping guide pipe (Figure2-1) was manufactured, which was then verified in January 2018
Tip-Folding
Controller

Cable Drum

Parallel Thruster
Front Lighting▲The submersible robot [IRIDOLPHIN-A] (Figure 4-2)

Bouyancy
Adjustment

Flat Bar
Warped Grating

Front camera
and lighting

PCV

Guide Pipe
(Φ110 ｍｍ)

Isolation
Valve

Front Camera

Side Cameras

Slotted Opening

Crawlerof the CRD platform’s grating has fallen. Thus, an access andAbove-water Monitoring
CameraTIP Guide Tube Support
CRD Transfer Device

Shapechanging

a small, self-driven crawler inside the pedestal (as in section 1

Propulsion Thruster

X-6
penetration at
Unit 2extreme environments, it was made highly resistant to radiation

Guide rings were installed on the jet deﬂector to prevent the cable fromtangling in structural members

Top Lighting

pipe, it was successful in investigating inside the pedestal andrecognition of its environment. To allow the robot to operate inOperational Details

To protect the ROV(ﬁber-optic γ dosimeter)

conditions by adding some highly concentrated illumination.

pipe with an internal diameter of Φ 100 mm. Once inside, it

cables. The other four types are equipped with various features

Instrumentation

The goal of the remote crawler robot (Figure 1-1) is to accesstight spaces and provide visibility. When entering the PCV, thefloor structure to prevent it from getting tangled up in the boat

Investigation Equipment

Developing crawler-type robots for conducting remote-operated investigations(FY 2014–2016)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

4

Unit 2

Shielding

Top Monitoring Camera(telephoto)

Member Research Project converted the equipment to an
investigation unit (Figure 2-3) with an additional finger that canmechanically work with sediment. The unit was then used toverify changes in the status of accumulated material. That

investigation confirmed movement in the pebble -shaped

Pedestal

Passage toContainment
Vessel (X-6
Replacement penetration)pedestal and produced data on the temperature and dose rate

sediment at the bottom of the pedestal (Figure 2-4).
Platform Control-Rod Drive
Mechanism

Cable

Cable-Feed
Mechanism

Birds-Eye
Camera▲The small submersible robot [IRIDOLPHIN-A2] (Figure 4-3)

Diagram of End Camera Mount

Pan/Tilt Camera and External Lighting▲ Investigating the inside of Unit 2's PCV using a guide pipe (Figure 2-1)
Skirt to prevent
contamination andsnags

Pedestal
Opening

Radiation andTemperature
Sensors

Jet Deflector

CameraSediment

Boat
Cable

▲ Images of submersible robot [IRIDOLPHIN-A] in action with installed guide rings (Figure 4-4)

47

Modified

CameraLighting

Investigation unit
in January 2018

Lighting

Lighting
FingerInvestigation unit that can be
opened and closed whencontacting sediment

Examining sediment▲

ClumpySediment▲ Investigating the bottom of the pedestal (Figure 2-2)

Can rotate toadjust orientation

ClumpySediment
Guide
Ring

Radiation andTemperature Sensors

Modifying the▲

Jet Deflector

Skirt to prevent
contaminationand snags

investigation unit
(Figure 2-3)

through direct contact(Figure 2-4)

48

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

5

Developing arm-type access and investigation equipment

Creating an access route: hatch-opening equipment and the isolation rooms

With the arm-type access and investigation equipment (Figure

The arm-type access and investigation equipment was unit tested

The purpose of this machine is to create a route for accessing

IRID has already completed the design, manufacturing, andrectangular steel container) then extended during internal

sections within the X-6 penetration (right side of Figure 3-1).

remotely opening the X-6 penetration hatch (Figure 5-2) via theequipment, and task training is currently being carried out so3-1), an articulated arm is housed within an enclosure (ainvestigating or the test retrieval of fuel debris to establish aconnection with the PCV via the equipment for creating anaccess route (the X-6 penetration connection structure etc.)(Figure 3-2).

in Kobe, where it was verified that it can pass through the narrow
Mock-up testing is currently underway at JAEA’s Naraha Center

for Remote Control Technology Development, and the plan is toconduct on-site demonstrations after some operational training
on the equipment.

mock-up testing of the isolation rooms and the hatch-opening

isolation rooms (the general term for the stage isolation room,that on-site demonstrations can commence.

the hatch isolation room, and the robot deployment room)(Figure 5-1) while maintaining the PCV boundary.

Robot-carry-in room

Test condition(X-6 penetration exit position)

Carriage

Wand
Minimumclearance

Link

Telescopic arm

the PCV interior through the X-6 penetration of Unit 2 by

Arm-headcameraCRD railX-6 penetration simulated structure▲ Arm unit test (X-6 penetration test)(Figure 3-1)

Robot-carry-in room
Arm-type access andinvestigation equipment(arm and enclosure)X-6 penetration connection structure

Isolation room(a full view from the back)【Isolation room in the stage】【Hatch isolation room】

Formation of boundary to cover inside theconcrete stage from the X-6 penetration sleeveto the hatch isolation room.

Boundary formation when opening the hatch.
Formation of boundary and shielding with air-tightdoor after opening the hatch.

○ Functions:

Isolation room

Site layout of arm access andinvestigation equipment (Figure 3-2)

Pedestal

PCV○Specifications:

Approx.1 ton/W1.2×L1.7×H1.8m▲(hatch isolation room)▲

Extension pipe

X-6
penetration

4

The X- 6 p e netratio n co nne c tio n struc ture has the PC V

to th e fla n ge of th e X- 6 p e n etratio n . Th e co n n e c tio n

designed so that investigation arms can pass through thet h ro u g h u n i t t e s t i n g , m o c k- u p t e s t i n g , a n d t e s t i n g i n

structure’s interior, and it is one of the equipment to create ana cce s s ro ute into th e P C V. S e lf- d rive n thro u gh re m ote
operation, the structure uses a gripping mechanism to connect
Grippingmechanism

Isolation valve

struc ture’s various func tions have already been verified

conjunction with the isolation room (Figure 4-2), and on-site
demonstrations are currently being prepared.

○Specifications:

Approx. 5.5 ton / W1.7×L5.3×H2.5m

Toward原子炉
reactor方向

X6 penetration【Hatch release】【Robot carry-in room】○Specifications:

Boundary formation when opening the hatch andcarrying equipment in and out

Approx. 2.3 ton / W1×L2×H1.6m

Nonmetallic bellows

Earthquake-resistance support○ Functions:

Opening of the X-6 penetration hatch

solation room

Lippedseal○ Functions:Concrete stage

Creating an access route: the X-6 penetration connection structureisolation function (an isolation valve) (Figure 4-1) and is

Hatch
release system
ハッチ開放装置（隔離部屋内設置状況）(installation in the isolation room)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

3

○ Functions:○Specifications:

Approx. 8 ton / W1.7×L5.3×H2.5m

Simulated X-6 penetration▲Overview of the isolation room and the hatch release (Figure 5-1)

▲ X-6ペネ接続構造の概要（図4-1 ）Arm cleaning and drying equipment

Drive / lift mechanism

Gripping and connection▲Overview of the X-6 penetration connection structure (Figure 4-1)
X-6 penetration connection structure

Shaft adjustment of X-6 penetration
Narrow area

Gripping and connection
Narrow area
Simulated X-6
penetration

49

Slope

		
Direction to the reactor

M16 volt
Hole saw

Hook tool

X-6 penetration hatch▲Test setup of hatch release operation (Figure 5-2)▼

▲Combination test with isolation room (Figure 4-2)

Hatch release(arm head)

Mock-up test

Self-driving remote operation50

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

Unit 3

1

Investigating inside PCV with a submersible robot (ROV)

Developing measurement technology

IRID developed a submersible robot (the “Little Sunfish” inaccumulated in several places inside the pedestal.

M e a s ure m e nt te chn o l o g y (s e n s o r s) a re curre ntly b e in gequipment or other investigation equipment, and each of them

(Figure 6-1) and investigate the inside of the pedestal. An actual

These results are being further developed to create a 3-D map

4-1 for an overview of each sensor.

a combination test is planned to for the sensors installed on thedemonstration test for the equipment. The demonstration test

retrieving fuel debris.

Figure 6-2) to access Unit 3’s PCV through the X-53 penetration

i nv e s t i ga t i o n w a s c o n d u c te d i n s i d e t h e p e d e s t a l a s arevealed damage to multiple structures, the accretion of whatof the pedestal’s interior that can be used to study policies forappears to be material that melted then resolidified in thedeveloped for detailed investigation inside PCV. Refer to Figure

All sensors are designed to be installable on the arm-type accessSensorpedestal’s interior, and the presence of lumpy sediment

Measurement
items

has already been tested and had its functionality verified. Next,
arm. On-site verification test is also being prepared.

Overview

Use of obtained data

Principles of measurement
Slits of light are beamed onto the target object, then the

ROV Entry Point (PCV penetration)

Structuresinside PCV

Camera formeasurement(radiation resistancecamera)Scanner head
(MEMS mirror and lens installed)

reﬂections are detected by light-receiving elements (the・To prevent the survey arm fromuses the principle of triangulation; variations in distancethe PCV, acquire data aboutof light, and that unevenness information is converted toafter the accident and reﬂect

lasor scanner method). The method of measurements

colliding with structures insidein the depth direction appear as unevenness in the slits

the internal shape of the PCV

distance.

that in the arm's operational

Peak directionof the point P

Distance to the point P in a depth direction

Water Surface

sediment(debris)Light-slit emitting window

Appearance of the sensor

Pedestal
Access Route

Slit lightsourceCameraLayout of the measuring
equipment and a target object

Distance to the point Q in a depth direction

Shape ofPeak directionof the point Q

system・C o m b i n e w i t h γ- ray d a t aobtained from the γ sensor to

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

1

Units 1-3

estimate the distribution of γ-ray sources

Peak directions whenirradiating slit lights

γ rays are measured by installing a high-radiation

detector (silicon diode) inside a tungsten collimatorthen using the collimator's slits to control the incident
γ rays.

Neutrally Buoyant Cableγ

Propulsion Thruster

sensorγ ray dose

Rollingmechanism

inside PCV(including
debris)Ascending Thruster

γ rays originating from each

Bearing

Main body
スリッSlitト

Collimator
Slit

Position sensor
Drive belt

Maintenance γ sensor

Detector

・Understanding the amount of▲Underwater ROV access route (Figure 6-1)

collimator
Detector
Signalcable

direction in the PCV・Combine with the shape dataobtained with the laser

scanner to estimate the
distribution of γ-ray sources

Preampliﬁer* ROV: Remotely Operated Vehicle

Applies a charge injection device (CID) as the camera
element.

Structures
VTcameraFront-Facing Camera

Lighting

The CID converts the light 's wavelength to an

Cameraelectrical signal.

inside PCVImages of
sediment

Light
Light

・Use to monitor when operatingLight

the arm (to avoid collisions)Air-blow
nozzleWiper
Radiation-resistant color camera
Cover glass(fuel debris)▲ Exterior of ROV (Figure 6-2)

・Use as investigation images

Light

Surface of camera (enlarged)
・A thin SiC semiconductor is doped with the neutron

originating fromthe ﬁssile material
withinNeutronsensorthe sediment(debris) or

ﬁssion products(pulsemeasurement)

Neutron absorption material
Sensor cavity
Center of gravity
140

Neutrons

473

Lead shielding High density PE
Aluminum casing

convertor 10B. α rays (10B(n,α) reaction) generated

when a reaction occurs between the convertor and aneutron (the 10B(n,α) reaction) excites charged
particles, which are then read as electrical pulses.

・The thin SiC semiconductor makes γ ray sensitivity
relatively low, allowing γ rays to be removed.

・To assess the distribution of ﬁssion products (nuclear

fuel) within the fuel debris, use a neutron adsorbent
that can measure the neutrons originating from fuelLead shielding

・Create a distribution of thenumber of neutrons in any
direction・Contribute to estimates of the
distribution of radioactive

material in fuel debris basedon the results of futuresampling of the fuel debris

debris with a relatively high sensitivity to maintaindirectionality.

▲ Lists of measurement technology (sensors) (Figure 4-1)

(damaged structures)(lumpy sediment)▲ Investigating the pedestal interior with the ROV (Figure 6-3)

51

52

Ⅲ

Decommissioning
Research and Development

Investigating and Analyzing inside PCV/RPV

Development of technology for investigation inside the reactor pressure vessel (RPV)■ Developing a top-access investigation method: Design a method for creating a hole in the operating floor at the top of thereactor building to ascertain conditions inside the reactor pressure vessel (RPV)■ Developing a top-access investigation method (improving processing technology): Create processing techniques that can
reduce the secondary waste generated during processing

1

Developing a top-access investigation method
IRID considered how to access the RPV interior from above and

IRID conducted element testing on the core technologies for thethen developed by opening a hole in the operating floor. At theboundary preservation equipment, ③ the processing equipment,

what such investigations would need. A method of access was

same time, investigation safety requirements and assessments ofradiation exposure was used to determine the features neededon the investigation equipment, which was then reflected in thedesign specifications.

top-access investigation method: ① the work cell, ② theand ④ the investigation system. The feasibility of those elements
were also verified and obtained design information that can beused to determine equipment specifications.

■ Developing a side-access investigation method: Design a method for gaining access by opening a hole on the side to avoid
interfering with work being done on the operating floor etc.

■ Developing a bottom-access investigation method: Access via the hypothetical holes in the bottom of the RPV to promptly
ascertain the RPV’s internal conditions

Research and development progress・N e g a t i v e p r e s s u r e t o p r e v e n t t h espread of contamination
Study maintenance and management
systems
・S t u d y e q u i p m e n t a n d m e t h o d s f o r
carrying it in and out

・Study the tool box
・Develop the gate valve
・Study the sealing
structure (resin packing)etc.

Air-lockcell

Operating floor

Work
cell

Well cover
PCV head
RPV insulation
RPV headPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Investigating inside the RPV to prepare
for the retrieval of fuel debris

Tool box
Gate valve on
tool box side
Tool box
transport cart

▲Tool box

Connector

Gate valve on
boundary guidepipe side
Boundary guidepipe▲Tool box transport cart

Steam dryer
Steam separator
Shroud head

▲Gate valve on
the boundary guidepipe side

Upper grid plate・Development of hole-opening
equipment
・Development of t small-diameterequipment for opening holes etc.
▲ Boundary guide pipe
(PCV head cconnecting part)

BackgroundTo dismantle Fukushima Daiichi NPS, the fuel debris and theWhen developing a new investigation method, it is necessary

retrieved. However, the location and condition of the debris

RPV. In this project, it was decided after much deliberation

reactor internal structures in the RPVs must be safely

and ithe reactor internals are currently completely unknown.

To prepare for retrieving fuel debris and structural materials,
it is necessary to gain access to the RPV interiors so that

where those materials are and what shape they are in can belearned. But accessing the RPV is difficult due to thecomplex structures in and around it—not to mention theextremely radioactive environment. For this, a newinvestigation method that can gain access is needed whileproviding shielding from afar and preventing the radioactive
dust generated during processing from being thrown up into
the air.

53

Purposeto first consider the access route, or how to gain entry to theto access the RPV from the top, side, and bottom. For the▲PCV headmachining
unit▲Openinghole for RPV
insulating
material▲ Flangerelease withspare nozzle
in RPV head

top and side access, the goal was to reach the core fromoutside the primary containment vessel (PCV) by opening

・Developing investigation system
・Developing investigationequipment
・Developing access equipment
・Visibility simulation
・Study horizontal expansion
mechanism etc.

Lift equipment▲Access equipment

Horizontal expansion
mechanism

holes; for the bottom, the RPV interior would be accessed

through the access route created in a different project forinvestigating the inside of the PCV.

The goal of this project is develop technology that will
enable these routes to conduct investigations while providing
remote shielding and preventing the radioactive dust

generated during processing from being thrown up into the▲Access equipment
for processing
the reactor
internals▲AWJ tool head
▲Superposition simulation ofobtained images

For prior conﬁrmation▲Investigation equipment

Full-scale investigation

AWJ: Abrasive Water Jet

air.

54

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

Developing a top-access investigation method (improving processing techniques)3

Developing a side-access investigation method

Of the core technology for the top-access investigation method,

can reduce secondary waste (material that was abraded etc.)

When using the top-access investigation method, there could be

PCV, and the RPV and developed equipment that inserts thetechniques. In FY 2020, sights on the body of the water

approximately 8 tons of material was abraded in FY 2019.

being conducted on the operating floor and in other areas.

maintaining the PCV boundary.

IRID has made some improvements related to processing
separator were set (see Figure 1) inside the reactor internal

structure, which is one of the targets for processing. The cutting
nozzle of the abrasive water jet (AWJ) was made smaller thenworked to determine the optimal parameters (spray angle,(refer to Figure 2 and Table 1). Preliminary estimates show that

However, this number was only 0.33 tons in FY 2020, so being
able to drastically reduce the amount of abraded material isanticipated.

delays in the investigation due to interference with other work

Therefore a goal of conducting this investigation at an earlier

date than the top-access investigation method was set. Regarding

IRID conducted element testing on the core technologies for theacce s s ro ute s, as s e s s e d wo rk risk s, and co nsid e re d theu p to th e s h ro u d h e a d, ② th e tre ate d wate r co ll e c tio n

the method for accessing the core from the side, IRID studied

applicability of the method from the viewpoint of each unit. As acutting location, etc.) as processing techniques were studied that

result, it was chosen to create an access route at Unit 2 bymaking a hole in the top of the core via the roof of the air-Work Conditions through FY 2019
Conventional Nozzle

Moisture
Separator
Body

Steam Dryer

Steam
Separator

Conventional Nozzle (yellow)

Work Conditions with the Small Nozzle

Moisture
Separator
Body

Moisture Separator (only three cylinders displayed)▲ Moisture Separator (Figure 1)

tube retention mechanism within the biological shield wall. The
feasibility of those elements were verified and design informationthat can be used to determine equipment specifications was
obtained.

WJD nozzle

Work cell

AWJC nozzle

R/B wall

Suction nozzle

ＲＰＶStandoﬀ is shortened

PCV▲Optimization of spray angle etc. (Figure 2)

Tool box AC0

Tool box ACx

Light guide tube
Biological shielding wall

▼Comparison of AWJ cutting procedures (Table 1)

Item

were created in surfaces such as the biological shield wall, theequipment, ③ the tools for the guide tube, and ④ the guide-

Small Nozzle (yellow)

Small Nozzle

Shroud Headconditioning room on the east side of the reactor building. Holes

side-access investigation method: ① the processing equipment

Moisture
Separator
Body

Standoﬀ

Shroud Head

investigation equipment into the reactor. All this was done while

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

2

Cutting Procedure through FY 2019
・The AWJ sprays diagonally downward from the steam
separator and makes repeated cone-shaped cuts down
the entire length of the separator.

Cutting Procedure with Small Nozzle
・The AWJ is sprayed from directly above the centerof the steam separator's three cylinders as the numberof steps are increased to cut the entire separator.

Conventional Nozzle
Small Nozzle

Cutting
Shape
Depth of each
step: 21 mm

Cutting the Steam Separator

55

Depth of each
step: 50 mm

Cutting the Steam Separator

Number of Steps

78 (depth of 21 mm)

33 (depth of 50 mm)

Length of Cuts

75mm

31mm

*EPDM: Ethylene propylene diene monomer rubber

56

Decommissioning

Ⅲ

ⅢInvestigating and Analyzing inside PCV/RPV

Decommissioning
Research and Development

Research and Development

Integrity evaluation of PCV/RPV and development of corrosion control technology

Developing a bottom-access investigation method
IRID studied a bottom-access investigation method to obtain

developed (wired and wireless) to gain access to the RPV (Figure

which involve the creation of holes; to apply the results ofinterior, several simple flight tests were conducted. For the wired

data earlier than the top- and side-access investigation methods,
existing technological development as much as possible; and toobtain visual data and dose-rate information. For this study, thetechnologies from completed and ongoing development projects1). To assess the feasibility of using drones to access the RPV

drone, a drone and a cable drum was chosen, while the wirelessdrone would have an extendable rod attached to it (Figure 2).

were compiled as well as data on the environmental conditions

The holes at the bottom of Units 2 and 3 are thought to beinterior and other work. The technologies that need to bem etho d s of acce s sing the R PV and to give te le s copingthat can be inferred from the results of surveys of the PCV
developed for each unit were then identified.

Since the holes in the bottom of Unit 1's RPV are believed to belarge in diameter, it was decided to utilize the crawler accessequipment currently being developed by the PCV detailed
inve stigation proje c t as metho d s for using drone s we resmall, so it was decided to establish a plan for developing
capabilities to the arm-type access equipment that will be used

Based on an earthquake-resistance assessment that considers
factors such as corrosion, evaluate the integrity of the PCV/RPV and develop technology that prevents corrosion■ Evaluated the integrity of PCV/RPV based on an assessment of earthquake resistance that considers the impact of falling fueldebris and aged deterioration caused by corrosion■ In order to determine the feasibility of the full-submersion method, IRID developed a simple evaluation method to quickly
assess earthquake resistance based on the various on-site conditions that are expected

to perform the gradually expanding retrieval of fuel debris■ Developed a corrosion inhibitor as a replacement for the nitrogen enclosure and evaluated its applicability for practical use

feasibility and evaluated the possibility of a telescopic-type■ Identified the damage to critical equipment inside the PCV/RPV when a powerful earthquake occurs as well as aftereffects(Figure 3). Several simple testing was conducted to determine
access equipment (Figure 4).

■ Considered countermeasures for preventing or controlling the above effects and confirmed their validity through an assessmentof earthquake resistance

Research and development progress

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

4

Investigating and Analyzing inside PCV/RPV

CameraTip
カメラ先端
RPV Bottom
Drone

Pedestal

Flight Path

Estimated Openings

Approx. 7ｍ

RPVApprox. 7ｍOpening
Crawler Access
Device*Upper Surface of Grating

Top of Grating
グレーチング上面* Planned uses include feeding in the power cable forthe wired drone, relaying signals to the unwired drone, etc.

▲ Accessing Unit 1 with a drone (Figure 1)▲2Accessing Units 2 and 3 with a telescoping pipe (Figure 3)

1013（Collapsed)

Drone▲ Drone flight test at Unit 1 (Figure 2)7194(Extended)

Investigationequipment

Telescopic
Pipecaused by corrosion. In addition to studying the retrieval offuel debris and the repair and sealing of the PCV from the

The severe events caused by the Great East Japan

viewpoint of earthquake resistance, countermeasures forprimary containment vessels (PCVs) at Fukushima Daiichi

applied that knowledge to preserving the structural integrity

earthquake exposed the reactor pressure vessels (RPVs) andpreventing or controlling corrosion were examined and

NPS to high temperatures, seawater, falling fuel debris, andof the PCV and RPV.

have degraded as a result. A plan is needed to maintaining

The damage to critical equipment inside the PCV/RPV andterm, while fuel debris is being retrieved from the reactor

identified while working to understand the conditions that

other hazards. There are concerns that structural materials

Cable
Winding
machine▲Telescoping equipment (Figure 4)

Backgroundthe structural integrity of the PCV and RPV over the long

their aftereffects when a major earthquake strikes were

core.

will be encountered when retrieving fuel debris, such as thewater level inside the PCV and the location of essential

Purposemachinery within the reactor building. Furthermore, IRID

IRID evaluated the structural integrity of the PCV/RPV based

those effects and confirmed their validity through an

on an assessment of earthquake resistance that considered

considered countermeasures for preventing or controllingassessment of earthquake resistance.

the impact of falling fuel debris and aged deterioration

57

58

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

3

Evaluating the feasibility of the ful-submersion method based on
the earthquake resistance of the PCV and RPV
As for plant conditions that reflect the latest repairs and other

The feasibility of retrieving fuel debris while above water andreactor building and large equipment was created, taking into

detailed earthquake evaluation was conducted for points that

submerged (Table 1). Using this model, the earthquake load(shear and moment forces etc.) was calculated on specific pointsof the PCV and RPV at each plant to assess the earthquake

had weak results.

A detailed earthquake assessment that assumed the suppression
chamber (S/C) support columns will not be reinforced was also
conducted.

electrostatic potential for crevice corrosion) was conducted in anenvironment with γ radiation and one with no radiation, then selected a・Future estimates: parameters・Future estimates: Parameters

・Damage model of reactor building

・Damage model of reactor

・In S/C: Full submersion

・D/W water levels: Approx.

・Vent pipes: Full submersion

・In S/C: Concrete OP-1050

Case 1(Partial submersion:
current water level)

Water levels

Full submersion(N/A, approx. 5,100 t and approx. 6,100 t)(2) Concrete 5%, steel 1% (for building design)Bolt(titanium)

1000x diluted sea water (19 ppm Cl-)

（ESP measurement)

BeforeElectrodeRod(titanium)

After

BeforeAfter(1) Concrete 7%, steel 4% (regulatory guide)(2) Concrete 5%, steel (for building design)

・Future estimates: Parameters (15 years and 40 years later)(15 years and 40 years later)

PCV.

・Damping constant: Parameters

Nut(titanium)

・Seismic waves: Currently Ss

5
3

24

5

3

24

▲Construction of electrode for corrosion test piece (Figure 2)

Exempliﬁed by Unit 3

Corrosion Inhibitor: Zinc/Sodium molybdate mixed with phosphate▲Testing the effectiveness of the corrosion inhibiter underγ radiation (Figure 3)

・Damage model of reactor building

・Damage model of reactor building

・D/W water level: Full submersion in the well(approx. 35 m, OP. 39920)

in the well・In S/C: Concrete OP1900

・In S/C: Concrete OP. 3570

4・Torus room: Concrete: Concrete (pin position of(pin position of the upper column support: OP.-100)

the upper column support:

・Vent pipes: consider repairs

OP. 2140)

・Additional facilities on the operation ﬂoor: Approx. 4710 t

・Vent pipes: consider repairs

・Cell installation: Yes

・Additional facilities on the operation ﬂoor: Approx. 6,100(1) Concrete 7%, steel 4% (regulatory guide)

・Cell installation: N/A・Seismic waves: Currently(2) Concrete 5%, steel 1% (for building design

・D/W water level: Full submersion

・Damping constant: Parameters

Washer(titanium)

the operation ﬂoor: Parameters (N/A, approx. 4,710 ton)

(1) Concrete 7%, steel 4% (regulatory guide)・Future estimates: Parameters

・Cell installation: Yes

・Additional facilities on・Damping constant: Parameters

・Seismic waves: Currently Ss

・Vacuum-break pipes: consider repairs

・Vent pipes: Fully submerged・Cell installation: N/A(1) Concrete: 7%, steel 4% (regulatory guide)

Gasket（PEEK）

・Torus room: OP.3200

・Additional facilities on the operation ﬂoor: Approx. 4,710 t

・Damping constant: Parameters(full submersion)・In S/C: Fully submerged

・Vent pipes: Water ﬂows in lower section

・Cell installation: N/A

Case 2・D/W water level: Approx. 6.5 m

column support: OP.-100)

inhibitor in the processed water lower than what it is when it enters the● Flag-shaped test piece test pieceElectrodeRod(titantium)building(pin position of the upper

the operation ﬂoor: Parameters

to dilute or filter collected water to bring the concentration of corrosion

・Damage model of reactor

・Torus room: Concrete

・Additional facilities ontreatment equipment, confirming that reducing those effects will require

2 and 3).

（ER, CREV, electrostatic potential test)

40 years later)

OP3100IRID has evaluated the effects of the corrosion inhibitor on water-

corrosion inhibitor that does not allow localized corrosion to form (Figures● Crevice corrosion test piece(10 years, 15 years, and

0.6 m

・Vacuum-break pipes:

no negative effects.

・Assessing the impact on water-treatment equipment

To evaluate the effects of a phosphorous-based corrosion inhibitor

・Future estimates: Parametersbuilding

・Torus room: OP3680

combined with a sterilizing agent was tested, verifying that doing so hasadhering to high-temperature surfaces

Unit 3(15 years and 40 years later)

・D/W water level: Approx. 2.9 m

conducted to verify things such as the temperature at which the substance・Evaluating the effects of a phosphorous-based corrosion inhibitor

Unit 2(10 years, 15 years, and 40 years later)

adheres. What happens when a phosphorous-based corrosion inhibitor isrepassivation potential for crevice corrosion, the rest potential, and the▼Case study on earthquake resistance including the latest updates on repairs and other conditions at the plant (Table 1)

Unit 1

To evaluate the anti-corrosion properties of the main carbon-steel
structural materials, electrochemical analyses (measurement of theresilience of those vessels.

Unit /
Case example

adhering to high-temperature surfaces, a batch test and a flush test wasanti-corrosion properties of the corrosion inhibitor

30

the current water level) and retrieving it while completely

・Using electrochemical analysis to evaluate the localized

60

account two different cases: retrieving fuel debris above water (atIRID has developed technology (a corrosion inhibitor) for controlling corrosion and evaluated its applicability for practical use.

while completely submerged (top access) was studied, then a60planned work, an Earthquake Response Analysis Model for theDeveloping countermeasures for preventing corrosion

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

1

・Damping constant: Parameters

Assessing the effects of corrosion on
the pedestal
IRID tested the exposure to the cylinder test piece, the scaled-down test

(1) Concrete 7%, steel 4% (regulatory guide)(2) Concrete 5%, steel 1% (for building design)piece (Figure 4), and the block test piece both in and out of the water・Seismic waves: Currently Ss

and under high-temperature heating. Data was acquired from concrete(2) Concrete 5%, steel 1% (for building design)

strength tests, rebar corrosion tests, and other sources and valuable

・Seismic waves: Currently Ss

knowledge for future study was gained.

2

A load was applied to the cylindrical test piece that simulated the RPV pedestal after the accident,
when it was exposed to high temperatures and was inundated with coolant. Through this, weverified things like the durability of concrete structures degraded by high temperatures and theeffects on their damage characteristics.

Simple assessment of earthquake resistance based on PCV repairs(plugging leaks) and rising water levels
After identifying and selecting the parametersthat affect the earthquake response analysis (thewater level inside the PCV etc.), analyses for the

Typical design evaluation

determining how the response to earthquake

Establish analysis conditions

dif ferent parameters were conduc ted,load changed as the parameters were adjusted.

(multipoint FEM)e v a l u a t i n g t h e e a r t h q u a ke re s i s t a n c e o fEarthquake response analysis

IRID also developed a simple method forequipment based on these combinations offactors (Figure 1). The results of this simple

Input(The cylindrical test piece was dried, heated for a long period of time at 400 ℃ and 800 ℃ , thenimmersed in water) Outer Diameter: 1,240 mm; Inner Diameter: 834 mm; Thickness: 200 mm

Simple method
Simple method* Outputs the load

without conducting
multipoint FEM

generally identical, confirming the validity of the

OutputTo improve the accuracy of predicting the amount of thinningcomponents that originated in fuel debris and core concrete aswell as their corrosive effects were studied, acquiring new
knowledge in those areas.

Evaluate each system and equipment▲ Evaluation process for typical earthquake assessments and the simple method (Figure 1)

59

Improve predictions of the thinningcaused by long-term corrosion

ove r a p e ri o d of 10, 0 0 0 h o u r s (Fi g u re 5). Th e li q u i d

Output load on each point and beam

simple evaluation method.

(miniaturized testing equipment) (Figure 4)

due to corrosion, IRID conducted a long-term corrosion test

(multipoint FEM)

method were compared to those of a typicaldynamic response analysis; the results were

5

▲Testing the durability of the RPV pedestal▲Conducting the long-term corrosion test (Figure 5)60

Decommissioning

ⅢInvestigating and Analyzing inside PCV/RPV

Research and Development

6

Establishing safety scenarios for large earthquakes
Addressing the potential risks that can occur from damage to heavy equipment during a large earthquake, the dynamic responses thatmust be prepared and the facility policies that are needed prior to the start of fuel debris retrieval were organized then created safety

scenarios (protocols for maintaining safety or responding to accidents). To create these scenarios, the following methods for assessing
earthquake resistance and effects were also developed.

① Developed methods for assessing the earthquake

・A method that combines a simple assessment which uses theresistance and effects on the suppression chamber (S/C)

amount of corrosion of RPV pedestral as a parameter and a

support columns

detailed assessment which uses a 3-D FEM elasto-plastic

To evaluate the earthquake resistance after plugging leaks inanalytical model (Figure 7)

the S/C, an analytical model that links the vent pipe and the・A method that assesses earthquake response using a model

S/C system (Figure 6) was created then a time -history

that links the reactor building and large equipment and sets as

response analysis was performed.

its parameter the decline in rigidity presumed to occur due to

For the critical evaluation points — the column supports,

corrosion of the RPV pedestal

earthquake supports, and vent headers — an FEM model was

・IRID also studied plans for verifying (through analysis, testing,also created and an elasto-plastic analysis was conducted toetc.) high-temperature corrosion and strength degradation inverify the conditions for filling the downcomer.

order to estimate the weakening of rebar inside concrete andto improve the above assessment methods. The following② Developed methods for assessing the earthquake

detailed analysis and material testing was then performed.

resistance and effects on the pedestal

・Time-history analysis of earthquake response for the S/Cmaterial data to evaluate the presumed history and distribution

・Obtained data via PCV material testing that considered the

IRID developed the assessment methods below and obtained

support columns in Unit 1of the pedestal’s temperature as well as the effects of fueltemperature history during the accident

debris on corrosion.

1/2 Sector
Solid Element Model

Vent Header Downcomer

Vent Pipe

Vent Pipe

Vent Header

Downcomer

Column Support

This research aims to provide confinement

1500mmfunctions in PCV and maintain the stability state

Conditions for pouring concrete into the S/C,
assuming downcomer leaks are plugged

Earthquake
Support

The Water Circulation
System andRepairing the PCV

745mm

Height of Filled
Downcomer

S/C

Ⅳ

Decommissioning Research and Development▲Time-history earthquake response model (Units 2 and 3) and conditions for linkage analysis of vent pipe and S/C system (Figure 6)
CRD Opening

for preventing the spread and dispersion ofradioactive material, radiation shielding andmaintaining cooling functions, therefore, therepair technology for leakage points of PCV was

CRD Opening

developed.

Additionally, various technical issues have been

CRD Opening

InsideOutsideCross-Sectional
View

studied and revised technological specifications

CRD Opening

for monitoring of the water circulation system in

Lower Access
Opening
CentralOpening

FEM model for Unit 1 at
Fukushima Daiichi NPS

the suppression chamber (S/C), the safety①Only inside is
deficient(elemental
reduction)② Scope of internaland external damage(free boundary)

Analytical model for corrosion due
to fuel debris▲ Development of a method for assessing the earthquake resistance of the RPV pedestal through fuel debris corrosion (Figure 7)

61functions required to establish an access routeto stop water leakage in S/C, reliabilit y,fe a sibilit y of c o nfin e m e nt fu n c tio n te st,earthquake resistance and long-term integrity,
etc.

Ⅳ

Decommissioning
Research and Development

The Water Circulation System and Repairing the PCVDevelopment of repair technology for PCV and full-scale test

Designing and building the devices and equipment needed to repair the full-scale test
unit for the lower PCV
IRID designed and manufactured a full-scale test unit thatworkability of the equipment for injecting concrete to reinforce

The test unit were assembled at the Naraha Center for Remote

repairs to PCV leaks), a full-scale test using water was also

simulates the bottom of Unit 2’s PCV at Fukushima Daiichi NPS.

■ Designed and manufactured equipment and system required to repair the full-scale test unit of the PCV bottom

Control Technology Development (Photo 1) . To confirm the■ Designed, manufactured, and installed equipment for water supply and drainage in the full-scale test unit of the PCV bottom
and for processing slurry

the S/C support columns (which was developed as part of the
conducted (Photo 2).

■ Developed technology for reinforcing the suppression chamber (S/C) support columns■ Developed technology for injecting fillings to stop leaks in the S/C■ Developed technology for injecting fillings in the vent pipe to stop leaks■ Developed remote-operated repair equipment for filling vent pipes■ Developed technology for filling the vacuum break line to stop leaks■ Compiled VR data for predictive simulation tests

Research and development progress▲Upper left: primary containment vessel (when building Unit 1;
simulated S/C is in the red box); Upper right: full-scale S/C(1/8 sector) CAD drawing; Lower left: vent pipe and S/C;
Lower right: assembled S/C shell and downcomer (Photo 1)

2

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Repairing leaks in the PCV in preparation
for fuel debris retrieval

1▲ Upper left: concrete-injecting equipment; Upper right: image from end
monitoring camera on concrete-injecting equipment; Lower left: mixer;
Lower right: effect of nozzle on concrete-injecting equipment (Photo 2)

Designing, building, and installing equipment for water supply anddrainage in the full-scale test unit of the PCV bottom and for processing turbid water
For the full-scale test, the following equipment were installed to simulate the range of possible dimensions, temperatures, etc. in theenvironments within Units 1–3 of Fukushima Daiichi NPS (Figure 1).

・Increasing temperature and supplying water: equipment that

・Operating floor: a work space for installing the various

・Treating turbid water: equipment for processing drainage water・Test-unit transport rail: a rail system for transporting the test

supplies accumulated water at its actual temperaturethat includes sealer (a grout containing cement)

To decrease the radiation dose during the retrieval of fueldebris, a full-submersion method to submerge the PCV wasconsidered. Since the PCV was leaking due to damage

sustained during the accident, it had to be repaired andsealed; however, the interior of the reactor building is highly

radioactive and narrow environment. Therefore methods andequipment for repairing the PCV remotely needed to be
developed.

PurposeA goal of this project was to develop technology that could

flooded with water. Since filling the S/C with water willunit (which weighs approximately 5,400 tons)

Research
Management
Building

Slurry Processor

greatly increase its weight, it was also aimed to develop

technology for reinforcing the S/C support columns to

Operating Floor

alleviate concerns about earthquake resistance.

Test Building

In addition to developing element technology, this project
also verified that the developed technology (such asprocedures and remote equipment) can be applied at thesite and performed a full-scale test for the purposes ofoperational training etc. This project was conducted at the

Naraha Center for Remote Control Technology Development
under a joint proposal with the Japan Atomic Energy Agency.

Various equipment assembled for

Heater / Water Pump▲

Backgroundvent pipe—that would presumably leak if the PCV wasmachinery needed to seal leaks

Test-Unit Transport Rail

Full-Scale Test Unit

the full-scale test (Figure 1)

repair and seal locations in the lower PCV—the S/C and the6364Decommissioning

ⅣThe Water Circulation System and Repairing the PCV

Research and Development

4

Developing technology for reinforcing the S/C support columns

Developing technology for plugging leaks in the S/C(sealing the downcomer)

For the S/C support columns, the plan is to fill them withtorus room. Flow tests were conducted with a maximum distance

To inject filling to stop the leaks in the downcomer, quencher,possessed the required properties was developed. Using theincrease their weight, so they must be filled with mortar that has

test was performed which showed that the target features can be

mm holes that are believed to exist in Unit 2, fill tests, long-conducted and its sealing capacity and workability was verified

concrete to prevent the S/C from leaking. However this willof 20 meters, build-up tests, simulations, etc. then a 1:1 scalehigh fluidity to ensure resilience to earthquakes. Various tests

expected to be achieved (Figures 2–7).

were conducted that highly fluid mortar material were poured

strainer, and other locations within the S/C as well as the Φ50distance force feeding tests of the underwater-inseparable

concrete, etc. were conducted and a composite material that

into six to eight insertion points in the first subfloor of the reactor

developed material, a seal test on a 1:1 scale model was(Photos 8–10).

building, so that it flows throughout the entire perimeter of theUnits 2 and 3

Reinforcement
Ring

Fill with reinforcing material

Downcomer← Injection
pipeStrainer

Quencher

S/C Test Unit

Interior of S/C (before filling)

S/C▲ Fill area for reinforcing S/C support columns▲ S/C support column validation test (1:1 scale; simulates Units 2 and 3)(Figure 2)(Figure 3)

It was confirmed that the reinforcement material flows through the fill pipe andwraps around the bottom of the S/C to uniformly fill the opposite side.

S/C damaged holes (before and after filling)

Interior of S/C (after filling)▲ Results of S/C water-seal test at 1:1 scale (Photo 8)
It was verified that the concrete moves over the reinforcement ring and buries the damaged holes, quencher,

Holes

Unitr 2:Eight locations

Hole closure
statusTo inject filling to stop leaks in the S/C, holes were opened in its upper surface and concrete was injected in.

Injection
Unit 1:Six locations

Φ50-mmdamaged hole

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

3

and strainer to prevent water from flowing out of the S/C.

Reinforcement Material

Reinforcement Material

Hose Feeding Device

Unit 3:
TBD

Monitoring
Equipment

Operational HQ

40m▲ Injection hole locations imagined for the reactor building’s
first floor (Figure 4)

Maximum interval between injection holes: approximately 20 meters▲ Long-distance flow test (Figure 5)
It was confirmed there to be no change in the reinforcement material’s quality
even after it was inserted from both ends and allowed to flow for 20 meters.

Utilities
Earthquake-resistance pit▲Verifying workability and remote operation (Photo 9)
Filling the S/C interior was conducted remotely from the first floor of the reactor building.

▲Simulation analysis offilling withreinforcement material
Injection hole

(Figure 6)

Filled S/C Interior

Taking into account both theresults of the full-scale filltest and a fluid analysis(complete enclosure model),Reinforced S/C

Support Column

a strength distribution map
was created. This revealed

that the compressive strengthof the reinforcement material
i n a l l a re a s s a t i s f i e s t h a t
required by assessments onearthquake resistance.

65▲ Results of filling test at 1:1 scale (Figure 7)
It was verified the state of the filler by making a cross-sectional cuton the test piece after the S/C had been filled.

① Hose feeding and collection device② Hose coiling device③ Hose feed mechanism▲ Equipment for filling the S/C (Photo 10)
To fill the S/C interior, the hose feeding and collection device and other equipment was set up on the first floor.

66

Decommissioning

ⅣThe Water Circulation System and Repairing the PCV

Research and Development

6

Developing technology for plugging leaks by filling the vent pipe

Developing devices for performing remote repairs to fill the vent pipe

Each of the elements needed to stop leaks in the vent pipe (the(self-compacting concrete that does not separate in water), ④

To achieve a waterproof seal by filling the inside of the ventth ey c an a cce s s th e wo rk target s within th e re al -wo rld

contaminated water from entering the S/C, and to create a water

operated repair equipment. Those test results showed that thethat could complete tasks such as removing obstacles or opening

equipment as well as other tools and cameras were conducted,

main path by which water leaks from the PCV), to prevent
level inside the D/W were developed separately then combined
and seal tested on a 1:1 scale model: ① the plug material (aninflatable seal), ② the secondary plug material, ③ the sealant

the repair material (heavy bentonite slurry), and ⑤ the remotesurfaces no longer leaked when water flowed over them, and weverified via a pressure test that the repaired locations wereleakproof up to 0.4 MPa (Figures 11 and 12).

pipe, IRID first developed remote-operated repair equipment
holes in the vent pipe. After designing and assembling theequipment, their workability tests were conducted to verify thatenvironment. Mock-up tests using the remote-operated repair
confirming various improvements to areas such as visibility,

reliability, and ease of use during installation and collection work.

Basement Floor

Basement
Floor

S/C

Sealed Vent
Pipe

In S/CPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 20225

Basic Specification for Vent Pipe・Original Purpose: Path for steam to flow into
the S/C・Basic Dimensions: outer diameter = 2,057 mm;
wall thickness = 9 mm

・Number installed: 8

Vent Pipe

・Material: carbon steel▲Verifying the workability of remote-operated repair equipment (Photo 3)▲Overview of the vent-pipe seal test (Figure 11)

The workability test showed that each task unit in the manned work should be completed within 20 minutes and that

① Plug material: forms a temporary barrier for insertion of the sealant. ② Secondary plug material: fills any cracks that remain after the plug material is installed.
③ Sealant: flows into the crevices of the plug material to form a waterproof barrier. ④ Repair material: repairs small cracks etc. caused by external factors andlong-term degradation. ⑤ Remote-operated repair equipment: creates holes in the S/C and vent pipe, removes obstacles, and inserts the plug material,
secondary plug material, and sealant through remote operation.
Sealing the vent pipe is done through the following process: Using the remote-operated repair equipment ( ⑤ ) from the first floor of the reactor building, theplug material ( ① ), secondary plug material ( ② ), sealant ( ③ ) and repair material ( ④ ) are utilized in turn.

a series of steps can be accomplished.
The following was achieved.

■ Drafted procedure manual for onsite application, ■ completed a test to verify workability, ■ verified thecapability for remote operation,■ verified the establishment of the procedure manual, and ■ identified challenges in onsite application

Preliminary mortar complete
1.5m3 First unit complete

Monitoring Systemleak rate (L/min)

Second unit complete

Vent Pipe at 1:1 Scale

Overhead FRM
Camera A

写真

Leak rate of 0.06 L/minafter two days

④ Repair material

leak rate

Bottom Elbow Camera
Wrist Camera

elapsed time since fill (minutes)

Vent pipe Interior
Camera② Secondary plug material

leak rate (L/min)Can be sealed even under flowing water① Plug material

Overhead FRM
Camera B

CameraLightsAfter filling with
self-injectingg concrete

After processing withheavy slurry③ Sealant
fluid pressure (MPa)Using the remote-operated repair tools and monitoring cameras, the ① plug material and ② secondary plug material was installed in the 1:1 scale test unit thenfilled it with self-compacting concrete that does not separate in water (the ③ sealant) and verified that it is capable of stopping leaks even under water flowing
at a rate of 3 m3/h. The ④ repair material was added and a water-resistance test was conducted in which it was confirmed there to be virtually no leaks at 0.4
MPa.

67

Camera to observe inside S/C

Third unit complete

Flowing Water 3m3/h▲ Results of seal test with 1:1 scale vent pipe (Figure 12)

FRM* Camera

Top Elbow Camera

Work
Operations

Can be swung on its hanging
axis to reduce blind spots*FRM: Fukushima Repair Machine (remote-operated repair equipment)▲ Monitoring cameras installed on remote-operated repair equipment (Photo 4)

The plan is to use a combination of at least two surveillance systems—the overhead cameras and the cameras
attached to the remote-operated repair equipment—to monitor work while it is being performed.
This will allow to understand the three-dimensional positional relationships from multiple angles.

68Decommissioning

ⅣThe Water Circulation System and Repairing the PCV

Research and Development

8

Developing technology for plugging leaks by filling the vacuum break line

Compiled VR data for predictive simulation tests

There are confirmed leaks in the bellows of Unit 1’s vacuum

installing a guide pipe, creating a hole in the vacuum break line,

IRID established an environment that allows operation of thecomparing the movement to that of a VR system that reflectedreactor building, a full-scale test was conducted in whichpressure (Figures 13–15).

reproduced on a virtual-reality (VR) system for training purposes.

verified through operational training (Photos 5–6 and Figures

break line. As we cannot enter just above the first floor of theand inserting watertight plugs then testing the resistance towatertight plugs were inserted via a flexible guide pipe that canremote manipulator used to stop leaks in the vent pipe to be

the collected data, then the effectiveness of the VR system wasThrough that process, the movement of the remote equipment

16–17).

was confirmed by using motion capture and other methods,

enter at an angle. Workability and feasibility was verified by

Guide Pipe

Watertight Plug

(Water level in the D/W)

Watertight Plug

Mast

Mast

Bellows

CameraPrimary boundary

Front Arm
Watertight-Plug
Insertion Tool

▲VR system with three-dimensional screen (Photo 5)

Balloon▲ Measuring movement of remote-operated equipment (photo 6)

Image from FRM elbow camera A2 (ractual equipment)
Pulled Up

Image from FRM elbow camera (VR)Compressed
the entire
part

Bellows
Watertight Plug
Chain

Watertight PlugRemote-operated movement
analysis model

Rear Arm

Guide Pipe
Vacuum Break Valve

MarkerTaking measurements▲Conceptual image of a method to stop leaks from the vacuum break line (Unit 1)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

7

① Drawing of the watertight plugs in the vacuum ② Fixing the flexible guide pipe
break line

Surface pressure
increases

Sealant(inserted in the step 1)③ Overview of a watertight plug

▲Conceptual drawing of the method to stop leaks from the vacuum break line（Figure 14）
IRID developed a guide pipe that can access the vacuum break line at an angle, since a hole cannnot be opened directly above the line and there are
obstacles around the perforations in the ground of the first floor.

The installation performance was also improved, with which the watertight plugs and accompanying insertion tool can be installed via the flexible guidepipe.

▲Compiling VR data for the predictive simulation test (comparison of actual equipment and VR cameras) (Figure 16)

Time［h］

0.02 Mpa
Results of water and pressure tests

Assumes movement that avoids obstacles such as
the spray pipe in the S/CPressure［Mpa］

Pressure［Mpa］

Pressure［Mpa］

Navigating around an obstacle placed on
top of the handrailTime［h］

0.3 Mpa
Results of water and pressure tests

Time［h］

Using the flexible guide pipe and improved watertight plugs, a workability test using the 1:1 scale test unit was conducted to verify the
In the leak pressure test, no leaks were observed at 0.45 MPa (completely submerged conditions).

69

Assumes movement that inserts the end of
a tool into a punctured hole0.45 Mpa
Results of water and pressure tests▲ Results of pressure test (Figure 15)
method’s workability and waterproofing capabilities.

Inserting a held rod into the simulated hole
placed in the top of the handrail▲ Movement test to verify VR operation (Figure 17)
A verification test comparing the performance of the actual machine and the VR system showed that there are no major differences
in movement and behavior and that the latter is sufficient for application in operational training.

70

Ⅳ

Decommissioning
Research and Development

The Water Circulation System and Repairing the PCVDevelopment of technology for establishing the water circulation system in the primary containment vessel (PCV) (development / full-scale test)

Organized technical specifications
for the water circulation system inside the PCV・When studying technology that can be used to create an(shared by the D/W and S/C) as well as design specifications

a c t u a l c o n d i t i o n s i n s i d e a n d o u t s i d e t h e P C V (t h elocations for the D/W water intake in each Unit were selected.

access route for water to flow into the D/W and S/C, theenvironmental dose rate on the first floor of the reactor■ Organized technical specifications for improving the water circulation system inside the PCV; studied work plans and created adevelopment plan
■ Developed and verified element technology for connecting to and accessing the PCV interior etc.
■ Verified through full-scale testing the technology for connecting to and accessing the PCV etc.

building (R/B), predictions of the state of the core, sediment・After looking at existing technology for creating an access(currently, during work, and when operating the water

utilize methods that were already used to enter the PCV during

accumulated within the PCV, the water level in the PCV
circulation system) were compiled for each Unit based on the
results of investigations conducted thus far.

・Establishing the examination of a water circulation system

Research and development progress

for the D/W water intake was organized, then the candidate(Figure 1) that improves the work and systems for retrieving
f u e l d e b r i s a n d o t h e r c o re s t r u c t u re s , t h e te c h n i c a l
specifications related to constructing a PCV access route

route into the PCV, it was found that it would be possible tothe on-site verification tests conducted in other projects(Figure 2). At that point, the challenges to creating the route

up to the water-intake point (pump pit etc.) in the PCV were 1)

how to lower the pump (pipe) to the basement floor via

remote operation and 2) how to remotely connect andexchange the pipe (hose).

Circulation system

Lower

Cooling system

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Establishing and verifying the water intake structure
for the dry well (D/W) and suppression chamber (S/C)
to establish the water circulation system inside the PCV

1

System for removing
insoluble nuclides
D/W
water intake
Scope of this project

Lower

during fuel debris retrieval

・IRID per formed element testing related to the remote(Figure 5), which was one of the development goals, thisas the insertion, installation, and collection of the hose at thepositioning system (Figure 6), an extension pipe to S/C auto-

bottom of the D/W (Figure 4), confirming that such a setup is

Coolant continues to be circulated into the primarycontainment vessel (PCV) at Fukushima Daiichi NPS tosteadily cool fuel debris, and the retrieval of fuel debris will

begin in the near future. That retrieval work will also require

to continue to safely manage the water within the PCV. In
particular, the work to build circulation lines for continually
pumping coolant in during fuel debris retrieval must be donein and around the PCV, so it is needed to develop
technology for connecting to and accessing the PCV tocomplete remote-operated tasks while still keeping it sealed

PurposeThere are issues regarding the water intake of the water

circulation system, such as the sealing function of liquid/gas,

its safety over the long term, and the possibility of remotefeasible. At the same time, the engineering challenges thatneed to be solved were identified, and plans for dealing withthem were studied.

・In addition to developing a plan for testing the application ofdevices during work and maintenance of the S/C water intake

applicable technology that can connect to and access inside

welder (Figure 7), a bead-removal equipment, and equipment
for creating a temporary seal.

・IRID performed element testing on each equipment or unit

testing on each prototype and verified that they have therequired functionality (Figure 8).

Gate-type Liftthe PCV for the intake of water to the water circulation
system.

Pipe
Installation
Stand
D/W
Water-Intake Pipeoff and safely managing the water inside it.

Hose▲Connection of D/W water-intake pipe (Figure 3)

71

project developed prototypes of the necessary equipment: an

installation and operation under a highly radioactive

environment. The goal of this project is to develop

line (Figure 2)

Developing and verifying element technology for connecting
to and accessing the PCV interior etc.
deployment of the pipe (hose) in the D/W (Figure 3) as well

Backgroundthe D/W water-intake

Pump Pit(Figure 1)

2

Creating▲

Water circulation system▲

S/C water intake

Pump▲ Inserting the pump into the D/W basement floor (Figure 4)

72

ⅣThe Water Circulation System and Repairing the PCV

Decommissioning

Research and Development▲ Positioning system for extension pipe
(Figure 6)

Extension Pipe

Extension Pipe
S/C

S/C

Welding Torch▲ Structure of the S/C water intake(Figure 5)▲ Extension pipe to S/C fitting auto-welder▲ Macro cross-section of weld

(Figure 7)(3 layers, 8 passes) (Figure 8)

3

Retrieving
Fuel Debris

Verifying through full-scale testing the technology for connecting toand accessing inside the PCV etc.
・Using the auto-welder prototype developed from work on the

JAE A’s Naraha Center for Remote Control Technology

remote operation (positioning the extension pipe and welding

for repairing leak points in the PCV by filling the S/C structures

S/C water intake (Figure 7), a series of full-scale tests via
it to the S/C fitting) were conducted. This testing showed that
remote operation of the S/C water intake is feasible.

・As part of the plan for preventing Unit 1’s D/W from leaking
into its torus room and S/C inner perimeter, the effectiveness

Development. This unit was manufactured to test technology
and the torus-room floor with mortar. This test confirmed thatthe torus room and the inner perimeter of the S/C could be
used as a boundary for contaminated water and powdery fueldebris (Figure 9).

of establishing a liquid boundary was tested by inserting repair

material. To do this, the full-scale test unit was installed at theSupport reinforcement + repair material⇒ Separate underground and contaminated water▲ Boundary-effectiveness testing and results (Figure 9)73

Water-level measurements for the outer perimeter(water level of inner perimeter: 6,865 mm)
Rise in outer-perimeter water level (mm)① Fill with repair
material② Insert simulated
Torus Roompowdery fuel debris
a n d c o n t a m i n a te dwater
③ Confirm whether thesimulated debris andcontaminated water
Undergroundare leaking
Water④ Confirm whether thesimulated debris andcontaminated water
have penetrated dueto the cutting of the
S/CⅤ

Decommissioning Research and Development

Before filling with repair material
After filling with repair

No increase in water level

補強材打設後継続at any point after filling
して水位の上昇無with repair material

Measurement time (minutes)

The inside of PCVs and RPVs are contaminatedwith radioactive material and their radiationlevels are high. As information about each PCV/RPV is limited and uncertain, equipment andsystem for fuel debris retrieval have beendeveloped while carefully verifying the feasibilityof the system.
Due to the differences of the reactor structure
and response to the nuclear accident, Fukushima
Daiic hi NP S ha s n eve r- b efo re -s e e ncharacteristics. Therefore, the fuel debris
retrieval systems has also been developed toensure safely containing, transporting andstoring retrieved fuel debris.

Ⅴ

Decommissioning
Research and Development

Retrieving Fuel Debris

Development of trial retrieval and technology for increasing the retrieval scale of fuel debris

Trial retrieval l of fuel debris

Utilizing some equipment from the project that investigated thePCV interior in detail—the enclosure, the equipment for creating

an access route, the arm-type access equipment, etc. — sample■ IRID is studying methods for trial retrieval of fuel debris and the gradual increase of the retrieval scalecollectors were attached to the end of the arm and the retrievalof an extremely small amount of fuel debris will be conducted.

■ Creating sampling scenarios that involve collecting fuel debris and bringing it to analysis facilities■ Capitalizing on the results of PCV interior investigations and the development of the arm for detailed investigation
inside the PCV, this project is designing, manufacturing, and testing equipment for collecting, accessing, andX-6 penetration connection structure(for detailed investigation of PCV interior)transporting of the arm.

DPTE container

Access equipment

Access equipment(for detailed investigation of PCV interior)Research and development progress
Enclosure (for detailed investigation of PCV interior)Transfer a portion of the debris powder collected in the
glove box into another container then transport it.

Transfer to sample vials then transport toexisting analysis facilities

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Trial-retrieving of fuel debris then gradually increasing
the scale of the retrieval

1-1

Type-A transport
container

DPTE container

Enclosure(for detailed survey of
PCV interior)(sealed container for
on-site transport)

Inside the reactor building

Outsidethe reactor building

Glove box

Work under IRID’s jurisdiction (in-house projects)

Work under TEPCO’s jurisdiction
Work Scope of
the Fuel Debris
Characterization Project▲Overview of the trial retrieval of fuel debris

1-2

Sample collectors for fuel debris (test retrieval)

IRID designed two types of fuel debris sample collection

bristles and a vacuum chamber. A wand was installed and a

attach to the arm-type access equipment: a brush with very fineunits.

equipment (equipment for the testl retrieval of fuel debris) to

BackgroundPurposeFor the appropreate safety management and equipment

Technology for accessing the lower PCV is currently being

some fuel debris from the bottom of the PCV and ascertain

interior. It was decided to incorporate that same concept for

design for fuel debris retrieval, it is significant to collectits composition and mechanical properties. IRID is studying

methods for collecting fuel debris and bringing it to analysis
facilities so that data that cannot be gleaned from camera
investigation or similar approaches can be obtained.

Tip of equipment for detailed investigation of
the inside of PCV

developed as part of the detailed investigation of the PCV
the development of technology that can gradually increasecombination test was conducted to verify the behavior of theTelescoping arm

Wand▲ Prototype: Ultrafine metal brush method

the scale of fuel debris retrieval. To collect fuel debris andretrieve from the PCV, however, it is needed to take steps toprevent the leakage of radioactive material throughout theentire route traversed by the debris. The retrieval of fueldebris involves bringing highly radioactive material into theenclosure, so even more than shielding the equipment itself

Prototype ofequipment fortest retrieval offuel debris

from radiation, systems that account for improving our ability
to seal the enclosure need to be established. A system forremotely transporting collected fuel debris samples from thehighly radioactive environment in the reactor building to the▲Conceptual image of combination test with the armfor detailed investigation inside PCV▲ Prototype: Vacuum tube method

analysis facility must also be established.

75

76

ⅤRetrieving Fuel Debris

Decommissioning

Research and Development

2-3

of fuel debris retrieval

Enclosure (isolation room)

Using the structure of equipment in the detailed PCV internala double door in the connecting section was installed so that the

IRID has developed a double-door mechanism that maintains an

as attaching and removing the fuel debris collection equipment

the retrieval of fuel debris can be gradually increased. In addition

PCV when the enclosure is withdrawn.

while maintaining the opening safely sealed in an emergency.

manufactured as well as remote maintenance on the accessinvestigation as a base, IRID made some modifications so that

airtight connection between the junction and the PCV opening

contaminated surfaces of the enclosure do not travel outside theto enhancing the hardness of the arm’s front half in order to

The prototype for an enclosure that allows to perform tasks such

improve its payload, the tip’s range of access was extended andEnclosureon the end of the access arm is currently being designed andequipment itself.

Double-Door
Mechanism

Equipment for creating access route(X-6 penetration connection structure)

Double-doormechanism

Dexter manipulator

Double-Door
Mechanism
Fuel debris collection equipment

X-6 penetration
PCV

Remote-operated transport cart
for fuel debris container▲ Prototype of double-door system

Connecting Section

Internal Structure

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

2

- 1 Developing technology for gradually increasing the scale

Arm-type access equipment

Neutron monitor(for monitoring criticality approach)For predictive
monitoring

For continuous
monitoring

2

performing actual work with the various

Equipment for receiving and delivering cell

Manipulator
Transportbox

The fuel debris collection equipment

IRID identified the issues faced when

Existing analysis facility

Measurement
equipment for U
and Pu1-4

equipment for collecting fuel debris —those for pebble-shaped and granulardebris (buckets and flexible grippers),

Manipulator

those for powdery debris, and those forslicing and collecting cylindricaldebris—and made design modifications

Transport
off-siteto improve their utility. tests to collectsimulated debris and cutting tests arebeing conducted, and improvements
On-sitecontainer

Samplecontainer

EnclosedbottleEnclosedbottleare being made to address the issues.

Container for transport
off-site▲ For collecting sandy▲ For collecting sandy▲ For cutting and▲ For cutting andfuel debris (bucket)

fuel debris

fuel debris

fuel debris

and pebble-shaped

2-2

The arm-type access equipment

IRID is designing and manufacturing the front half ofInvestigationequipment inside2

PCV

the arm to include the following improvements forgradually increasing the scale of fuel debris retrieval.

-5

and pebble-shaped(flexible gripper)

collecting cylindrical

Remote-operated transport system

For fuel debris collected with the access equipment, the plan is totransport it from the enclosure using a storage container. IRID has

・Reduce the number of links in the arm (shortening thedeveloped a “remote transport cart” that can load and unload fuel debris

length) and increase the payload (from approximately
10 kg to 20 kg)

Link

・Use a link with a horizontal offset mechanism tostorage containers in the highly radioactive enclosure then transport them
Arm for fuel debris retrieval

to lower-radiation areas via remote operation.

the retrieval scale

Element testing was performed related to the capability to adjustthat gradually increasing

improve the arm’s ability to pass through openings on
the pedestal platform

positioning when connecting to the enclosure and maneuverability innarrow area. Based on those results, the equipment was designed,▲Two varieties of arm-type accessequipment

77

collecting powdery

prototype units were manufactured, and plant verification tests wereconducted.

Fuel Debris Storage Container▲ Remote transport cart with fuel debris storage container

78

Ⅴ

Decommissioning
Research and Development

Retrieving Fuel Debris

Development of technology for retrieval of fuel debris and reactor internals

Research and development progress
Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Conducting research and development (R&D) for manufacturing
the equipment required for fuel debris retrieval
Below are some of the challenges faced when retrieving fuel debris. To respond to these issues so that work can be completed
safely and promptly , IRID is conducting the R&D necessary to manufacture equipment for fue debris retrieval.
■ The work must be done entirely from a remote location, since it is a highly radioactive environment that humans cannot enter■ Radioactive material must be confined to prevent its dispersion while the work is being conducted■ For fuel debris retrieval to be conducted safely, criticality needs to be prevented, limit the radiation exposure of workers, andprevent the structures from falling or being damaged■ As the work will be conducted in the dark and narrow spaces within the reactors, its efficiency through remote operation must
be improved

Research and development progress

Background
The fuel debris inside the reactor pressure vessel (RPV) andprimary containment vessel (PCV) at Fukushima Daiichi NPS
is currently being maintained in a stable, cool state. But

when the accident occurred, areas including the reactor

building, RPV, and PCV were damaged, so the plant itself
was experiencing instability. For this reason, a goal of the

project is to retrieve fuel debris safely without dispersing
radioactive material and while achieving the objectives
below.

・Ensure confinement function for radioactive dust・Establish remote-operated technologies that work inhighly radioactive environments

・Establish technologies that reduce radiation exposure
and prevent the spread of contamination

PurposeIRID will conduct the following R&D to manufacture theequipment required for fuel debris retrieval, including
technology for creating the access routes (both top- andside-access) needed to retrieve the fuel debris located in the
RPV and PCV; technology for handling fuel debris; andelement technology related to confinement functions.
・ Perform element testing etc. as needed to study
conceptual methods (both top- and side-access) for
retrieving fuel debris and assess their feasibility

・Develop technologies related to the retrieval of fueldebris (remote-operated assistance, transport methods,
d u s t- c o l l e c t i o n s y s t e m s , s o r t i n g f u e l d e b r i s a n dradioactive waste, etc.)

・Develop technology for handling fuel debris (collection
and disposal)

・Develop element technology related to confinement
functions・Develop systems for continually monitoring the inside of
the primary containment vessel

7980

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Plan

Retrieving fuel debris: Developing technology for the top-access method

1 Removal and transport of the entire structue

Plan[Basic Methodology]

In its subsidy projects through FY 2018, IRID studied methods

for cutting up the structures inside the PCV then placing them・Transport each structure as one unit to a newly built facilityof highly radioactive and contaminated areas by providing

containers. Minimize cutting operations inside the hazardous

into unit cans for transport, with the goal of reducing the scopeaway from the reactor building, then cut it up and store it inshielding around the fuel debris and preventing the dispersion ofenvironment of the core to improve throughput while loweringdust. Element testing related to cutting up the CRD housing was

the difficulty of remote work.

performed, which are obstacles on the reactor bottom.

・When transporting the structures, use large, dedicatedThis re s earch reveale d tre me nd ous dif ficultie s from thecontainers so that maintaining the objects shielded and airtightperspective of workability and operational hours (throughput),can be handled via the containers, the passageway, or someand ways to make improvements and to transport large-sizedcombination of the two.

structures as possible have been searched for since 2019. As

1 Removal and transport of the entire structures

IRID has continued to study various methods based on the policies in theprevious pages. In FY 2019 and 2020, element testing was performed toverify the manual procedures for cutting and disassembling objects on

Support Arm

the reac tor floor and refle c te d those re sult s in the throughput

RPV
Lower Li

2021, disassembly methods for the reactor core parts etc. have been

Grating

calculations. Specific issues to this process were also identified. Since FY
studied and their feasibility through element testing are being confirmed.

Cutting Too

HCU Pipe

In addition, examination was conducted on the function of preventing the

spread of contamination (airtight construction) of the dedicated transport
container (large carry-out container) used to carry out the structure

Plan 1 of that strategy, the basic policies on the removal andremoved in one unit, and it was decided to use a double-lid structure forthe lid part of the container. The feasibility of the double-lid structure istransport of the entire structure was organized.

being confirmed through elemental tests starting in FY 2020 to see if▲ Removing obstacles from the bottom of
the reactor

airtightness can be ensured by studying the operational steps of the
double-lid structure.

Support Arm

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Retrieving fuel debris: Developing technology for the top-access method

Cutting Tool
Simulated CRD Housing

Grating

Grating

Clamp

Removal oflarge structures fromthe reactor floor

Testing the removal of simulated grating

HCU Pipe

Cutting Tool

Cutting ToolAWJ Cutting Nozzle

HCU Pipe▲ Simulated the reactorbottom▲Transport with the cutting method and element testing related to cutting the CRD housing

Top Lid

Additional
Building

Work Container

Cutting test on simulated HCU* pipe
*HCU: hydraulic control unit

Retrieval
Equipment▲Testing the top lid of a dedicated (large) transport
container

Top Lid
Double Lid(currently open)
Shielded Cart

Bottom Lid

Transport
Cart

Large Transport
Container

Dedicated
Transport(Large)Container

Top and bottom lid fall
▲ Removal and transport of the entire structure

81

Container
is transported

Top and bottom lid connect
to container

Top and bottom lid disconnect
from each other▲ Moving a dedicated (large) transport container82

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Retrieving fuel debris: Developing technology for the side-access method

In an effort to improve work efficiency and reduce operational

2) A confinement area (boundary) separate from the existing

There are likely to be many changes throughout the course ofin the PCV was d evelop e d while working in narrow

aim to divide and decontaminate the structures at the same timearea will be double-sealed by a primary boundary surrounded

and the progress of operations there. However, it can beequipment, such as maneuvering around a simulated pedestal

Plan

2 Combination of dividing into the structual components and decontamination

hours (throughput), IRID compiled and studied basic policies that

Removing obstacles (flexible work arm)

reactor building will be newly created with containers. The

to reduce the load that needs to be transported. This was Plan

by a secondary boundary, and negative pressure will be used②, which was executed separately from Plan ① .

to form an additional dynamic boundar y. The primar yboundary is further subdivided into three stages with different[Basic Methodology]

1) Structural waste and fuel debris will be loaded into containers
as close to the RPV as possible then transported.

levels of contamination and radiation dose.

3) Containers will be gradually moved from the highly radioactive
areas in the primar y boundar y to the less radioactive

・Drying, sorting, transfer to containers, and other work done

secondary boundary then transported out.

in preparation for storage will take place in a separate

fuel debris retrieval due to unknowns in the work environment

e nv i ro n m e nt s s . E l e m e nt te s t i n g wa s p e r fo r m e d o n t h eassumed that the radiation dose rate will be high for work done

i n t e r i o r a n d c u t t i n g a s i m u l a t e d I C M (i n c o r e n u c l e a r

within the PCV. IRID thus developed a lithe remote-operated

instrumentation) housing. This testing showed the device to be

arm that is highly resistant to radiation and that can flexibly

capable of performing such tasks.

respond to these on-site changes. That concept was thenapplied to the equipment for removing obstacles in the inner-

Since FY 2021, IRID has been studying methods to remotely

removing obstacles in the outer-pedestal (basement floor) withof fuel debris, such as HVH (heating ventilating handling) units

pedestal, which combines the design of the equipment forremove large obstacles that will significantly hinder the retrieval

assembled beams.

facility, away from the reactor building (R/B). The goal is tore d u c e t h e w e i g h t o f t h e c e l l b y re m o v i n g s o m e

and CRD exchangers.

In FY 2019–20, a smaller equipment for removing obstacle thatfunctionality.

could be used to clear interfering objects and monitor conditions

・Obstacles will be cut down to a transportable weight using a
combination of decontamination and shielding.

CRD opening

Flexible work arm

しゃShielding Unit

Equipment forremoving obstacles

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Retrieving fuel debris: Developing technology for the top-access method

Flexible work arm

Winch

Platform beam

Hoist

Width = 3.4 m
Total Weight = 3.6 t

the outer-pedestal (basement floor)

Flexible work arm(accessing the ﬂoor)⑥ Transported by crane⑦ Transported by cart
⑧ Container decontaminatedduring transport

Movement arm▲ Equipment for removing obstacles in

Flexible work arm

Pedestal opening*Minimum size for equipment entry: approximately 700 mm

Flexible work arm
Movement arm⑤ Moved to transport containers▲ Small-sized equipment

Assembled beamfor removing obstacles▲ Equipment for removing obstacles in the Inner-pedestal
with assembled beam

Movement arm

Fragment conveyer④ Waste transferred to storage container
Load (motor)③ Lifted up

HVH

CRD

② Segmentation process

exchanger

Vent

Fragment
loader

1.00m① Transport to DSP*

Lifting machineTransport rail

Platform*DSP: dryer separator pool
1.42m

1.42m
CRD 交換機撤去Removalrobot for
ロボット
CRD
exchanger

Waste Storage Container▲ Lifting machine used in HVH dismantling

83▲ Dismantling with CRD exchanger

84

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Retrieving fuel debris: Developing technology for the side-access method

To retrieve fuel debris, a variety of equipment needs to be

In FY 2019–20, research and testing related to the welded

Retrieving fuel debris requires us to process and collect it without

・Dimensions: Width = 1,900 mm

debris outside the PCV. IRID is studying how to create adetermining the welding requirements and protocols while

will be selected during the internal investigation conducted later,		

Access tunnel

transported into the PCV and move extracted obstacles and fueltransport route by connecting the reactor building (R/B) to an

additional facility via a shielded access tunnel and how to

support the load of the access tunnel with the R/B outer walland biological shield wall so that the weight limits of the 1st floorof the R/B are not exceeded.

Access rail and robot arm

connection between the access tunnel and PCV was conducted,
identifying potential issues. Based on those results, IRID has

been conducting full-scale welding tests and exploring methods
for remotely installing the section that connects the PCV andaccess tunnel (the PCV connection sleeve) since FY 2021.

IRID's plan was to assemble the access tunnel outside the R/B to

Since the access tunnel must be able to turn (slide) near the R/B,

be carried into the R/B and set up through remote operation. In

being performed on the slide of a full-weight simulated test unit

reduce the radiation exposure of workers, after which it would
FY 2018, element testing was used to verify the design’s

feasibility by sending in a prototype with the correct shape andthe slide section is also being studies and element testing is
to confirm its feasibility.

knowing its properties or distribution. While processing methods
IRID has already developed a six-axis robot arm and a three-stage extendable access rail to guide that arm into the pedestal.
This will allow to flexibly respond to a great variety of fuel debris
processing methods.

[ Overview of the method ]・The biological shield wall near the existing X-6 penetrationpassage will be removed and the hole in the PCV wall widened[ Robot Arm ]six axes.

・Has redundant hydraulic cylinders and supply lines in case of
failures.

・Cables and hydraulic hoses are housed inside the equipment,

・The robot arm will move down the access rail and into the・Develop methods for controlling the hydraulic cylinders and[ Access Rail ]・Dimensions: Width = 700 mm

・The rail is equipped with a feature that guides the robot arm
into the pedestal.

in a compact form factor.

preventing damage from contact with surrounding obstacles.

rota r y a c tuato r s fo r e a c h a xis s o that th e tip c a n b erepositioned with a precision of ±5 millimeters.

the cell.

・The rail is equipped with a feature that transports unit canisters				

Arm Length = 7,100 mm
Height = 920 mm

・Weight = approximately 4 tons

・Load of Tip = 2 tons in the downward direction

for the purpose of fuel debris collection.

Welding head
Welding robot

PCV
Stand

Simulated Equipment Hatch

Primary
Containment
VesselStand

RPV

Access Rail

Reactor Pressure
Vessel▲Test welding the access tunnel connection▲Conceptual drawing of access tunnel

・Weight = approximately 24 tons

toward the CRD exchanger opening then extended at a・A three-stage extendable design is used to reduce the size ofAccess tunnel

Height = 2,500 mm(when contracted) / 17,000 (when extended)

・Uses hydraulic power to achieve strong processing capabilities

pedestal, where it will process and collect fuel debris.

Welding robot		to gain linear access to the pedestal interior.

downward angle.

Additional
facility

Length = 8,700 mm

・The processing tool on the end of the arm can reposition along

・An access rail equipped with the robot arm will be aimed

dimensions.

		

Pedestal

Fuel Debris

Biological shield opening
生体遮へい開口Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Retrieving fuel debris: Developing technology for the side-access method

Access tunnel (actual
shape and dimensions)
アクセストンネル（実外形）

Pedestal

Robot ArmPCV
CRD Exchanger Opening
Fuel Debris▲Overview of side-access method

Hydraulic Cylinders▲ Element testing for delivery of the access tunnel

Load support points▲Turning the access tunnel

85

Hydraulic Cylinders

Mock-up
模擬体

Sliding section

Hydraulic Rotary
Actuators

スライド部
Sliding
section▲ Full-scaled weight simulated access tunnel▲ Access rail and robot arm▲ Design of robot arm

86Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Developing common technology shared by the top- and side-access methods

Studying installation of the cell

Equipment to transport fuel debris

IRID is working to develop a side-access method in which thesupporting materials. The feasibility of this plan was then・Anticipating cases where the drying facility for fuel debris

penetration opening so that retrieval equipment can be installed

・To reduce cell size and weight, the access equipment forequipment for transporting unit canss containing undried fuelverified.

biological shield wall will be removed to widen the existing X-6

debris retrieval was changed to a fixed rail type and theand the fuel debris within the pedestal can be accessed throughheight of the equipment itself was loweredby scaling downa short distance.

the rail’s transport capabilities. At the same time, the・To solve the issue of the shielded debris-retrieval cell

thickness of shielding was optimized to decrease weight.

c ann ot b e n e ar th e re a c to r b uilding , IRID d eve lo p e ddebris to a facility located away from that building.

・Using the top-access method as the model case, the following
feasibility assessments was conducted.

■ Throughput assessment■ Explored additional methods for monitoring whether safetyfunctions remain intact■ Evaluated the radiation exposure of workers■ Considered scenarios for rescuing devices based on a
defense of depth

exceeding the weight limit of the floor when installed in the・IRID also designed a series of steps for installing the massive■ Established basic design requirements based on safety and・In conjunction with the above assessments, IRID also studied

minimizes the size and weight of the cell and reinforces theinstallation — and verified the feasibility of that installation■ Studied concepts for equipment that fits the layouts used bytechnologies such as the hydrogen treatment mechanism andreactor building, the installation was adopted that both

cells—ranging from preparation work through mounting andfl o o r b e a m s a n d wall s o f t h e re a c to r b u il d i n g w it hprocedure.

Cell for fuel debris retrieval
Spare cell for fuel debris retrieval

Dual-Arms

Maintenance Cell
Equipment Transport
Container

AccessEquipment

Transport
Container Yard
Equipment Transport
Container Yard

The fuel debris transport

Tool changer for connecting
to the hydrogen treatment
mechanism

can is connected to the hydrogen

Equipment totransport fueldebristreatment mechanism installed on

Pedestal

maintain a static double boundarywith fuel debris, it is stored in the

Storage
Canister
Cell
Cell for fuel debris
transfer container
Fuel debris transfer
container yard

4500mmHydrogentreatment
container(wheels)d u ri n g o n - s i te t ra n s p o r t , w h i c hFixed Railtakes place after loading onto the

PCV▲ Access equipment with fixed rail
Shield Doorcart. All steps in the process
Biological Shield Door

Hydrogen treatment container (outer shell)

Pallethyd ro ge n t r e a t m e n t c o n t a i n e r s

Reactor
Building

the double door structure used to transfer unit cans.
Hydrogen treatment mechanism(main/reserve unit)

Battery andcontrol panel for
movementt h e p a l l e t s . To c o n t i n u o u s l y

Mini-Cask
Celland tested the theoretical validity of essential element

fuel debris retrieval methods

e quipm e nt that hou s e s the unit

Reactor Pressure Vessel

LLW Transport Container Cell

handling demands

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Retrieving fuel debris: Developing technology for the side-access method

2500mmthrough container storage are

2500mmcompletely unmanned.

3400mm

3400mm
3400mmPCV

Installing cells and other Cell Adapter
large structures▲

Fuel debris Retrieval
Equipment▲

Fuel debris Retrieval

Fuel debris Retrieval Cell

Developing technology shared by the top- and side-access methods
Preventing contamination from spreading to the S/C

a highly feasible option. However, it was found that remotely

expand the range of debris collection and increase the risk andtime required for retrieval. IRID is currently exploring whether

barriers can be erected inside the PCV to prevent the dispersionof fuel debris.

reduced by 90% compared to manual operation2 by a veteran operator.

1 For this, the movement plan technique was utilized for the robot with multiple degrees of
freedom that was developed under the project to enhance the foundational technology for
retrieving fuel debris and the reactor internals.
2 In addition to adjusting the position and attitude of the fingers, this also includes redundantaxis movement developed as part of the project to create technology for remotely
decontaminating the building’s interior.

3-D model linked toactual machine
Moving the joystick forward
starts the device along
the trajectoryWhenhung fromgroundfloor

Approx. 500mm

型枠
Mold

After
installation,
extendRobot hand

When hung from ground floor

Pedestal side

Fills mold with mortar

Frame

Unfolding mold

▲ Illustration of folding mold

Hose
ホース

Arm

▲

Mortar

PCV shell

Approx. 3m

Mesh

Flexible structure
arm

Hose

型枠
Mold

Sides:PCV shell

Sides: Plastic▲ Installing the mold

Filling the mold with mortar

87

through the element testing, it has been verified that this strategy

Frame② This automatically calculated trajectory is then reproduced on the actual

beginner and veteran operators alike; however, the time was able to bewhich the wall is built by filling a folding mold with dry mortar;

Barrier

automatically1 (in 1–2 minutes).

is reenacted. It requires several minutes to achieve the goal positioning fordifficult. To get around this issue, IRID is considering a method inPedestal

Arm
for
堰設置用
wallアーム
installation

manipulators, a trajector y that avoids obstacles can be calculated

the forward and back directions to control the speed at which the trajectory

installing a mold then positioning a hose to fill it with mortar iscan be accomplished through remote operation.

① By setting a goal (position and direction) for the fingers in relation to two

machine. While this is taking place, the operator can adjust the joystick inmortar was considered, and element testing has shown this to be

Water
level

who remotely operate the robots through narrow environments with poor visibility.

or other locations when the debris is processed, which would

Approx. 900mmplan) to increase operational efficiency and reduce the work load on operators② Reproducing
the trajectory

IRID has developed a technique for supporting remote operation (a trajectory

To build such barriers, installing molds then filling them withcontaminated water will spread to the suppression chamber (S/C)① Trajectory plan

Remote-operation support

One concern with fuel debris retrieval is that fuel debris and

Mold
型枠
NozzleFilling the folding mold with▲

Developing common technology shared by
the top- and side-access methods

mortar88

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Collecting granular fuel debris through suction

・Investigate methods for exchanging a unit can (separator)

Suction is an effective method for improving throughput when・Explore which type of pump is applicable for vacuumingcollecting fuel debris that is granular (0.1 mm to 10 mm). IRID

underwater and at low water levels

has identified and responded to the following requirements
regarding the vacuum collection equipment.

・Consider what shape of pump strainer applicable to thissystem・Study methods for detecting when a unit can (separator) is
full

Developing technology for a system that collects fuel debris dust
Designing the dust-collection and dispersion-control system

will cause dust to be dispersed into the air, worsening the work

Processing headenvironment and placing a tremendous burden on the operation andmaintenance of equipment. To address this, IRID has designed andmanufacture a system for controlling the dispersion of dust in the vicinity

developed, depending on factors such as the relative density ofother material during the processing of fuel debris, the area around thethat can collect approximately 40 kg to 300 kg per hour wasof processing work. For the purpose of preventing the spread of dust andthe target material and the water level in the environment.

processing sites could be structured to stop dispersion was examined forTool changer

Water ejector

Mist nozzle

Mist nozzle

the applicability of a system that achieves efficient dust-collection anddispersion control based on existing technologies.

Compressedair supply
line

separator

Robot arm

There are concerns that processing nuclear fuel which contains fuel debris

To meet these requirements, a vacuum collection equipment

Top lid ofseparator

Damaged fuel in the reactor
core (example of predictedfuel debris)・Conceptual design of system and assessment of applicability for

Circular saw

practical use

・Cre ate b a sic s p e cific atio n s a n d sy ste m d e sign; d e sign a n d

Intake pipe

▲ Conceptual drawings of circular-saw tool for the dustcollection and dispersion-control systemmanufacture processing headseparator
Exchangemechanism

Circular saw

Processing headPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Developing technology for handling fuel debris (system for collecting and containing fuel debris)

Metal-plated
drive motor

Developing technology for a system that collects fuel debris dust

Water pump
Metal-plated
rotating
strainer

Vacuuming under water▲ Installation on the access equipment

Designing remote-maintenance technology for the dust-collection and dispersion-control system

Ejector pump

Vacuuming at low water levels▲Vacuum collection equipment for granular fuel debris

As the equipment and systems used to retrieve fuel debris (systems

for cutting fuel debris and collecting dust; work tables; monitoring
systems; robot arms that manipulate tools; etc.) will be installed inhighly radioactive areas, they must generally be maintained
remotely. IRID has therefore organized the various approaches formaintaining such equipment, taking into account that they will be

Developing technology for handling fuel debris (system for collecting and containing fuel debris)
Transport processes applicable to debris characteristics

・Fuel rods turn brittle due to the radiation generated when a

Methods for collecting fuel debris will be selected according toreactor is running, so we should be able to tear off thatthe size and other properties of the debris.

Processing
head

Replacement ofprocessing head begins

Processing head is
moved

Unit can is lowered andhoist is collectedhandling fuel debris. The maintenance methods and their feasibility,
identified issue areas were studied, and logical policies forresponding to those areas on the actual machines were considered.

・Study maintenance methods as well as procedures forexchanging parts

Processing head isstoredHoist is attached tounit canUnit can is raised▲Concept for replacing parts with a remote-maintenance tool

material with a bucket or hydraulic cutter. If cutting or・For clumps of material that melted and resolidified, we will

breaking it with a cutter is difficult, a cutting grinder or

prioritize use of a chisel (spiky hammer), which is fast andsimilar tool will be used before the hydraulic cutter iseasy to use; brittle sections will be torn down with a bucket.

damaged.

IRID anticipate scenarios in which the target object cannot

be demolished due to its physical properties or accessibility.

Developing technology for a system that collects fue debris dust

・A collection equipment that combines multiple robot arms

Cutting grinders, laser gouging, and other methods will bewill be used for collection work. The end of the robot arm

employed in such cases.

will be equipped with a gripper, various processors, avacuum collection equipment, or other implements.
▲

Hydraulic Cutter

Test Piece

ExhaustEvaluation test to collect dust and control dispersion

Housing

Dust-measuring
deviceCircular-saw
test unit

IRID manufactured a test unit to evaluate dust dispersion then conducted testing

CameraOf the processing methods that are anticipated to have the highest throughput as▲Testing dust dispersion

in the vicinity of a processing area.

scope of retrieving fuel debris and the reactor internals, IRID selected those that
Fuel Assembry
Gripping
Mechanism

Blade of
Hydraulic Cutter

are expected to generate relatively large amount of dust. First, test elements that
could be applied to all processing methods were considered. Then, the processing

Arm C

head was designed, manufacture and installed on the dust-collection and

Arm A
Maximum 285mm
Cutting
Grinder

dispersion-control system, which would serve as our prototype for element testing.
Finally, how well it can cut the simulated fuel debris was tested, which mimics the

Arm B

predicted properties of the fuel debris. IRID also tested the rate at which dust

migrates into air and water in the space around the processing head and evaluatedthe efficiency of dust collection to verify the applicability of the system to control
the dispersion of dust.

89

Amount ofdispersion controlled (%)

determined by the project that developed technology for further expanding the

M

E

PW

M+E

PW + M

PW + E

M: Mist spray E: Water ejector PW: Processing aid (water) spray
M+E: Combination of M and E M+PW: Combination of M and PW
E+PW: Combination of E and PW▲ Relationship between spray type and dust-

collection efficiency (processed with ceria)

90Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Developing technology to sort fuel debris and radioactive waste

The coolant-circulation system depicted in the diagram below isrequirements for each of its components. Next, the literature

If we can measure the amount of nuclear fuel in the material

method, passive neutron method, passive γ method, muon

numbers of α nuclides and other radioactive nuclides originating

removed were conducted to select the candidate material fordecommissioning of Fukushima Daiichi NPS then sort it into fuel2020–2021, IRID used simulations to evaluate how much the

Equipment for removing nuclides from circulated coolant

being considered for use during the retrieval of fuel debris. Largein fuel debris are expected to dissolve in the circulated coolant
due to increases in surface area resulting from fuel debris
processing, changes in the quality of water accumulated within
the PCV, and other factors. To reduce the risk of leaks anddecrease the effects of radiation exposure on the public, thecoolant- circulation system may need to be continuously
operated while these nuclides are removed.

Development through FY 2020 focused on first designing theoverall concept of the liquid system and determining thestudy and adsorption tests for the α nuclides that must be

adsorbing the nuclides. Based on those results, the concept for asystem that removes soluble nuclides was designed and thechallenges involved in applying that concept to an actualmachine were identified.

Since FY 2021, IRID has been studying plans for conducting
suction tests in a simulated gas phase environment and one thatuses actual coolant to further assess suction performance within
the above-water sections of the PCV as well as the impact that
coolant components have on suction performance.

Reactor water
injection

Particle
removalDrainwater/
backﬂow

Adsorption tower for removalof soluble nuclide

A

P

Particle
removalP
Drainwater/
backﬂow

Supernatant waterA

Tank (including
boric-acid control)

Supernatant water/
Permeable waterP

Underground water

For existing water
treatment system

SludgeSludge collection andsedimentation separation system

Sludge collection andsedimentation separation
systemTreatment system fordehydration andstabilization

Permeable water

For canisterstorageproperties, compositional diversity, and other characteristics ofeverything from the retrieval of fuel debris to its storage, which

fuel debris affect measurements taken; this was done to identify

will help stabilize Fukushima Daiichi as soon as possible.

the challenges in developing each of these measurement

Unfortunately, the technology for doing so does not currently

techniques (Figure 1-2). Moving forward, aiming for on-siteexist anywhere in the world. Thus, IRID studied multiple sorting

application, it is necessary to develop measurement systems,

scenarios (Figure 1-1) and conducted the technical investigations

reassess sorting scenarios, conduct verification tests usingneeded for each sorting point.

simulated debris, and take other action according to plan.

Through this process, the following measurement techniques that
could be applied to sorting were identified: the active neutron

R/B and additional buildings (for each reactor)

Storage facility (shared use)

FueldebrisB

Sludge (ﬁlter)(preparation prior to storage)

Ba
Unit canCanister

PP Criticality Concentration

Cf

Cd

To canister storage(fuel debris storage)

Ca

CanisterCanisterLarge diameter
canister

To canister storage

Targetobjects

Storage facility

C: Storage facility

A: PCV InteriorSupernatant water* Particle removal: Used to collect drainwater after coarse ﬁltration
P …pump○▲ Flow chart of the liquid phase system (example)

debris and radioactive waste, it should be possible to streamline

B: PCV Exterior

To existingwater-treatment system

Treatment system fordehydration andstabilization

Complete ﬁltration system
for sludge collection

A*

scattering method, and high-energy X-ray CT method. In FY

SludgeTo canisterstorage

Treatment system fordehydration andstabilization

B
SludgeConcentrated water

Supernatant water

retrieved from the primary containment vessel during theShape and Dimensions

RO membrane for removalof soluble nuclide
Adsorption tower forsoluble nucleic acid removal

Sludge collection andsedimentation separation
system*Used only for the application of boric acidCollected boric-acid solution

Boric-acidcollection system

B

Water for fuel debris retrieval

Boric acid (silo)

Tank for collecting
boric acid

CoolerStudying sorting scenarios and evaluating measurement technology

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Developing technology for handling fuel debris (disposal of fuel debris and accumulated material)

Large diameter
canisterCa(beforedismantlement)

Large
subcriticalcontainer

Visualobservationand criticality

Criticality

Large subcriticalcontainer

Large subcriticalcontainer

PP
Cf

Ba

Developing technology for handling fuel debris (disposal of fuel debris and sediment)

Waste

Boron-adjusting equipments u b s t a n c e s i n u n d e r g r o u n d w a t e r, a n d b o r o n t h e nThis re s e arch e xplo re d th e is su e s with u sing a s o dium -pentaborate solution to evaluate criticality, a method that wasdeveloped under a subsidy project as part of establishingmethods for evaluating criticality and developing technology for

determined a rough guide for managing concentrations.

・To collect the boron, evaporative separation and a reverse-

osmosis (RO) membrane was chosen. The theory behind the

B (boric acid)
Na (Sodium)βγ (Cs, Sr, etc.): Low
Impurities (Ca etc.): Low/High

Boric-acid solution(soluble nuclide removal system)P

Receiving tank forconcentrated solution

P

CondensationtankP

硫酸 or 硝酸等

Receiving
tank

P

Ca
adsorption(collection tank)

Boric-acid waste solution(Soluble nuclide removal system)

Red line: Flow of boric acid
Black line: Other than above

Permeable water (B)pHadjustment

ROmembrane

ReceivingtankConcentration tanketc. omitted

P

Concentrated solution(including uncollected B)

Additional water +pH adjustment

P

Concentrated solutionROmembrane

P
P

Concentration needed tocontrol criticality(at least 5,150 ppm)

Boric-acid solution

Components of boric-acid control system (when using RO membrane)▲ Boric acid collection technology

91

P

Separation byevaporation

朱線：ほう酸の流れ
黒線：上記以外

Processing capacity 22 m3/day
Soluble component
B (boric acid)
Na (Sodium)βγ (Cs, Sr, etc.): HighImpurities (Ca, etc.): Low/High

Receivingtank

CfSorting points are marked with ●.

PP, not covered bysafeguards

Waste-storage
canister(contamination level)

Criticality, PP (physical protection) etc.

indicate the technologies needed at each sorting point.

Waste-storage canister(contamination level)▲ Sorting scenario plan (Figure 1-1)

Particle Motion Simulation

Boric-acid solution(collection tank)Na(mg/L as Na)(soluble nuclide removal system)
Processing capacity 22 m3/day
Soluble componentCIadsorption

PP

Waste-storage canister(management level)(neutrons, photons, electrons, and muons)Components of boric-acid control system (used for separation by evaporation)Caadsorption

Waste-storage
canister(management level)

within a slightly acidic environment was confirmed.

・Obtained deposition data on concrete component s,

Receivingtank

Criticality

Waste-collection
container(subcritical mass of
nuclear material)(waste storage)

PP, covered by safeguards

points at which boron passes through the RO membrane

controlling it. The results of this study are summarized below.

Boric-acid solution

Waste-collection
container(subcritical massof nuclear material)

Storage facility

Cd

Deposition of
sodium tetraborate

No deposition
Operatingrange
Deposition of boric acid

Distilled water(soluble nuclide removal system)

Container

Potential Detector(tallied)

Spontaneous Fission(source of neutrons)(Cm244,Pu240)
NeutronU235,Pu239
Divisions Triggered
FP（Cs137,Eu154.Co60）Photon(used when adding water)

Boric-acid waste solution

1-7(Increased Combustion)

Muon

Deposition

No Deposition

Literature Value▲ Scope of operational

adjustments (under continuous
use of boric acid)Stored Material

Approximately
Linear Fill Rate

1-4
(Decreased
Fill Rate）
1-19(MCCI)

1-1 (Base)
1-8,9 (FP Emission
Rate)

1-2
(Low Uranium
Content)

1-17(Metallic Debris)Appoximately
1-16
Linear Debris（High Uranium
Construction
Content）(Uranium Ratio)

1-10,11(Cooling Period)

1-5（Increased
Fill Rate）
1-3（MOX）

1-6(Decreased Combustion)

Uranium (kg)▲ Simulation of passive γ measurements (Figure 1-2)

92

ⅤRetrieving Fuel Debris

Decommissioning

Research and Development

When further expanding the scope of fuel debris retrieval,interior that occur due to fuel debris retrieval work, this projectchanges in the PCV over long periods of time during retrieval

safety requirements and whether operations are continuous) as

necessary technology is to continually monitor environmental

will organize and select monitoring items (taking into accountwork; this is to ensure safety and maintain throughput while【The reasons for work delay: Nothing during processing debris 】

Airflow
Primary boundary (*)

monitoring methods while considering on-site feasibility. The

radioactive environment that contains many unknowns. To

Cell (R)overall development process is indicated in Figure 1.

continuously monitor the environmental changes in the PCV

Debriscontaining
mechanism

Nitrogen【The reasons for work delay: Acceleration of
jig abrasion】

Nitrogen

Study policies forutilizing on-site data

Accumulatedwater

Arrival of some dustthat entered a liquid
phase

assessments described in ❶ and selected the following
areas to study in earnest moving forward:

・Organizing the technical specifications neededof the processes for retrieving and transporting fuel debris,

Our critical monitoring items included things like measured

identified as causing operational delays. For the risk assessment,

differential), the environment (status of fuel debris processing

example of the results for critical monitoring items.

for monitoring① Temperatures in before/after electric heater② The amount of dust in the heater entrance and nuclide composition

D/W
accumulated
water transfer
pump A/B

Cell (R)Cell (Y)Cell (G)Intracellular Transport Mechanism (1)Intracellular Transport Mechanism (2)HEPA
FilterCharcoal
Filter

Pre-Check forExtraceullar
Transport

Door mechanism/
outside of cell

NitrogenSupplyPreProcessing
DeviceContainer Sealing
Mechanism

Door Mechanism Y/GDoor Mechanism R/

Well

Extracellular Transport Mechanism

ExhaustTank

Tank
Particle Remover

PCV accumulated
water discharge
pump

Cooler
To wastetreatment system❸ Examining management method of
integrated management support technologies
So that operators can more rapidly and appropriately respond tofor plant data supplied by a continuous monitoring system within the・Explore ways to install measurement equipment

PCV. This was done from the perspective of supporting the daily

・Identify technical challenges

activities of operators and helping with analysis and the planning ofwork improvements over the medium and long term.

・Studying policies for utilizing on-site data

areas:

continuously.

・Exploring overall protocols for on-site dInner-PCV continuous
monitoring system

Fuel Debris
Retrieval Model
Debris-Retrieval
Performance
Debris-Retrieval
Performance

Safety Model

SafetyPerformance

To existing water treatment facility▲ Protocol for identifying critical monitoring items for the PCV interior (Figure 2)

93

PCV accumulated
water supply
pump

e tc .), a n d w h e t h e r s o m e t h i n g n e e d s to b e m o n i to re dWork Equipment
Model

Tank

PCV
accumulated
water buffertankMoving forward, these concepts will be applied to the following

Waste
Port

Soluble Nuclide Remover

D/W
accumulated
water large
particle removing
equipment

physical quantities (such as dust concentration and pressure

Secondary
Boundary

Cell Transport Mechanism

Exhaust【The reasons for work delay: Acceleration of heater deterioration by dust irradiation】

Supporting the daily activities of operators

Debris Storage Mechanism

Charcoalfilteron-site conditions each day, IRID studied concepts of the operation

・Study additional methods of monitoring・Examining the critical monitoring items① Differential pressure of filter② Amount of dust accumulated in filter inlet and distribution ofparticle diameter▲ An example of the extraction result of important monitoring items (Figure 3)❷ Exploring monitoring methods❶ Investigating the monitoring items for the PCV interior

Secondary boundary【Cause for delay: Acceleration of clogging filter due to dust accumulation】

To wastetreatment system

IRID analy ze d the tre nd s in the re sult s fo r the ris kthroughput into account (Figure 2). Also, Figure 3 depicts an

HEPAfilter

① Chemical properties of liquid flow into the pump (pH, chloride ion
concentration and chemical compositions)② Pump flow volume, etc.

▲ Development process for a continuous monitoring system inside the PCV (Figure 1)

we devised a weighted method that takes safety, operators, andPre-treatment
system【The reasons for work delay: Acceleration of partly clogging due to dust
contamination】

Concepts for
applying monitoring data

including additional safety features, and assessing the risks

Part of dust transferredin the air phase reaches.

↑ Purified gas↓ PCV atmosphere

Debris【The reasons for work delay: Enlarged opening hole owing to
PCV corrosion】

Secondary boundary

To exhaust port

DebrisAdjacent to D/W. The volume of mist coming to the PCV walland chemical properties (pH and chloride ion concentration)

Study overall protocols
for on-site data

IRID determined the critical monitoring items by creating models

Neutron absorbent supply
system for processing work

Dust dispersion during
processing fuel debris③ Neutron flux that is not affected bydarkness, fog and cooling water.
④ Dimensions of debris formation

Cell (G)Area that is not affected byenvironmental change in PCV
caused by the door mechanism
during processing debris.

Cooling
water【The reasons for work delay: Fuel debris criticality
approach】

Monitoring method isa critical development item

Organize speciﬁcations
needed for monitoring

Debrisprocessing andcollection

Nitrogen① Degree of wear on the processing jig② Compressive strength of fuelIdentify technicalchallenges

Select criticalmonitoring items

Dust transferredin the gas phasemoves to the cell
(R).

Nitrogen supply system

Study methods for
installing instrumentation

Risk assessment to
identify potential causes ofoperational delays

Cell transfer mechanism

Well

Study additional
monitoring policies

Cell (Y)Door mechanism R/YExisting subsidy projects
and other data

Dust concentration needs to be detected as it will be acontamination element of the cell (Y) during debris
transportation.

*Boundary of piping
connecting PCV is notindicated.

Liquid flow

well as compile and study the underlying concepts of various

remotely operating machines within a highly contaminated andfuel debris
retrieval model

Developing systems for continually monitoring inside the PCV (Proposal)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Developing systems for continually monitoring inside the primary containment vessel (PCV)

Assess risks andidentify criticalmonitoring items▶Top phenomena thatcause obstructions▶Deﬁne mechanismsthat lead to
obstructions▶Weighted fromthe perspective ofsafety, the operators,
and throughput

Changes in the
PCV's internal
environment altervarious types ofequipment data

Monitoring datafor on-siteequipment

Dataset fortechnology applied
to the support ofoperators

Planning data for
medium- andlong-term operations
Updateoperatonal plans
工事計画更新Analyzedby person in
運用支援責任者にcharge ofよる分析
application support

Supervisor
verifies situations
指揮者の状況
and認識を判断
makes decisions
Operators
conduct work
オペレーターの実行

Work and safety equipmentused to retrieve debris

Helping improve medium- and long-term
operational plans

Data on daily work
achievements

Evaluation data to
reduce on-site
uncertainty

Matchable assessment data of
the internal PCV environment
and medium- and long-term
operational plans

Analyzedby person in charge
運用支援責任者による分析
of application
supportSupporting medium- andlong-term analyses▲Concept for on-site application of a continuous monitoring system inside the PCV (Figure 4)

94

Ⅴ

Decommissioning
Research and Development

Retrieving Fuel Debris

Development of technology for criticality control of fuel debris

1

Criticality scenarios and risks based on the estimated distribution of fuel debris
Overall, there is believed to be little risk of criticality occurring

Nevertheless, there is currently a lack of information on theduring retrieval. Although there remains a significant amount of

which is why IRID is developing technology to prevent anddue to fuel debris becoming submerged or undergoing changes■ Determine the risks of criticality occurring in fuel debris; understand the effects of radiation exposure on criticality;

fuel debris with relatively large exposed sections, the risk ofdevelop technology for appropriately managing criticality

criticality should still be low considering what it could possibly be■ Monitor approach to criticality and develop methods to prevent criticality

detailed distribution, composition, and properties of fuel debris,
mitigate criticality if it does happen to occur.

composed of.

■ Develop methods to swiftly detect and halt criticality if it occurs■ Finalize methods for applying the developed criticality-control technology to the on-site environment

Research and development progress

1

2

1
3

2

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Detecting and controlling abnormalities promptly toprevent criticality during fuel debris retrieval

3
4

4*CRD: Control-rod drive mechanism▲ Locations where fuel debris is assumed to exist

2

General overview of technology for controlling criticality during fuel debris retrieval
To maintain subcriticality, IRID is studying plans for managing② monitoring technology used to ensure that criticality is not

This includes:③ technology that can quickly detect a state of criticality; andimminent while work is being performed;① restrictions placed on work protocols so that reactivity stays

BackgroundPurposeMonitoring data from the plant has revealed that the fuelTo ensure the safety of workers and the general public andwater content of fuel debris can be expected to change as

retrieval of fuel debris, IRID has worked to develop

debris is currently in a subcritical state. As the properties anddebris retrieval and other operations are conducted,
however, criticality needs to be properly managed so that it
does not occur in the future. This will allow fuel debris
retrieval to be completed safely and without issue.

low, and if necessary, technology that prevents criticality

suppressing it in the rare chance that it occurs.

① Prevent criticality

a. Restrict and control work
protocols to prevent
criticality① determine the risks of criticality occurring in fuelby criticality;② develop techniques to detect if criticality is near andin the unlikely event that it occurs.

・Monitor to observe if approaching criticality(near the retrieval worksite)

Specifically, the goals are to:debris and the effects of radiation exposure caused④ halting criticality through an emergency injection of boric acid

② Monitor for criticality approachto prevent them from being exposed to radiation during thetechnology for avoiding criticality as well as detecting andthrough the use of a neutron adsorbent;

criticality based on multi-stage protection.

b. Inject neutron absorbent(1) Soluble neutron absorbent (boric acid)(2) Insoluble neutron absorbent

Neutrondetector③ Detect criticality

・Detect signs of criticality at an
early stage (throughout PCV)

Gas radiation
monitor

Emergency injectionof boric-acid solution④ Mitigate impact

・Promptly suspend
operations whencriticality is approached.

prevent it from occurring;③ develop techniques to detect if criticality has
occurred and to halt it; and④ devise ways to apply the developed criticality-

B₄C
sinteredmetalmateriaGlassmaterial
containing
B and Gd

Gd2O3
particles(Note) Required technologies will be selected based on the risk of criticality.

management technology to specific on-siteenvironments.

95

96Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Technology for preventing criticality (restriction and control of work protocols)5

Technology that detects the recurrence of criticality via a PCV gas radiation monitor

It is pos sible to imagine scenarios in which critic alit y is

By limiting the amount of fuel debris that can be retrieved per

The PCV gas-management system is continuously measuring thebeing measured to detect the recurrence of criticality; however,consistency and water content.

of criticality can be prevented (for example, the graph up to 16

released into the environment.

detector) that allow Kr-87 and Kr-88 to be measured will speed

radioactive concentration of nuclides to monitor how much isshift, unexpected increases in reactivity from leading to a state
cubic centimeters in Figure 1).
〈Before Processing〉

Collected Core

▲Core boring; accumulation of water fine debris powder and in the cored hole

〈After Processing〉

Scope of
Retrieval(Cube)

Water

200cm

200cm

●

Measurement
Delay

110cm

Inert GasGe

Inert Gas
Generated▲ Evaluation system (added reactivity due to the perfect mixture of

Gas Chamber[Additional]

in the vicinity of fuel debris retrieval work.

9.1h

16.5s

2.8h

Kr
88
Half are formed
immediately

Short life span

(100x concentration)

Xe-135 Criticality Detection(concentration 1 Bg/cm3）

Signs detected quickly
in about 5 hours

Blue: ﬁssion yield Orange: half-life▲ Behavior of monitored nuclides▲ Relationship between additional reactivity and the amount ofTime (h)▲ Evaluating criticality detection when a criticality event is suspected (Figure 3)

debris retrieved per shift (Figure 1)

6

Technology that measures neutrons to monitor for approaching criticality
To monitor for approaching criticality, neutrons will be measured

Br
88

Xe
135
1.7％

First nucle withshort life spans

Single-side length (cm) of retrieved fuel debris cube

▲ Laser gouging; dispersed material accumulated in specific locations

4

1.4％

▲1回の取り出し量と添加反応度の関係（図1）

Raw Material

6.6h

Initial Concentration Ratio

Gouging
Head

I
135

even for fuel debris that is unevenly distributed (Figure 2).
*Kyoto University Critical Assembly

Scope of simulated
fuel debris(large reactivity)

signals (core noise measurement by Feynman-alpha method),n order to differentiate fuel debris according to its properties,
IRID studied absorption material with different characteristics: asolid (granular) type and a solidified (water glass) type.

In testing conducted at KUCA*, IRID learned that the effective

multiplication factor can be estimated by analyzing neutron

Technology that uses an insoluble neutron adsorbent to prevent criticality

Mediumreactivity

Smallreactivity

Polyethylenethat simulates water

A variety of tests related to basic material properties, adhesion
and miscibility, secondary effects due to radiation (hydrogengeneration, changes in water quality or structural materials, etc.),
Retrieval Equipment

Neutron Detector
Scope ofsimulated
neutronadsorbent

NeutronFuel Debris

NeutrondetectorAir(reactor core simulating the neutron adsorbent and an uneven debris

Estimated value for eﬀective
multiplication factor (keﬀ)

distribution)

Eﬀective multiplication factor is estimated bymeasuring core noise(with the Feynman-α method)candidate material.

・Verifying the miscibility of fuel debris and adsorbent (Figure 4)

・Confirming that the material can be injected underwater at
a stable speed (Figure 5)

fuel debris processing
tool: chisel

Neutronsource▲ Example of testing system that simulates fuel debris at KUCA

Neutron Signal

workability, and other areas were conducted to select theGlasscontaining
B and Gd

Gd2O3
particles▲ Solid Adsorbent

Glass containing
B and Gd

Water glass /Gd203 particles

Surface before test

Adsorbent▲ Solidified Adsorbent

Simulatedfuel debris

Fuel Debris
Adsorbent (Solid)

Crushedfuel debris

After adsorbent spraying
and processinga) Underwater testing with chiselb) Surface appearance(simulated lower pedestal)
before and after testing▲ Injection/Crushing test on fuel debris with insoluble adsorbent (Figure 4)
Veriﬁed parameters
・keﬀ and H/U of the system
・Type and arrangement of
neutron sources
・Positioning of detector
・Size of the system
・Non-uniform distribution
・neutron adsorbent etc.

▲ Applying the solid-type adsorbent

Acrylic Tank

Neutron absorbent
feeding device

Adheres to surface of pebble- and rock-like debris thenpenetrates into crevices
Assessment
Tray

Tray Weight / arb.unit

Melted Section

Concentration :4.0wt%

Additional Reactivity

Dispersed
Material[Existing]Concentration : 5.0wt%

Kr-88 Criticality Detection

Initial Concentration Ratio

Concentration : 4.0wt% , Height : 30cm , Void Fraction : 0.29−−Kr88 −−Xe135 −−Amount of Fission

0.08％

6.3％

Concentration : 4.0wt% , Height : 80cm , Void Fraction : 0.18

Laser

A＝135
About the same yieldfor spontaneous andinduced ﬁssion

Mass Number (A)Concentration : 5.0wt% , Height : 30cm , Void Fraction : 0.25

Water

Spontaneous, gas-generating Cm244 ﬁssion
U235 induced ﬁssion▲Yields for spontaneous and induced fission

Concentration : 5.0wt% , Height : 80cm , Void Fraction : 0.13

After Pulverization

A＝88
Yield diﬀers for
spontaneous andinduced ﬁssion

Shielding(Pb)▲ PCV gas-management systemwater and fine fuel debris powder)▲ Pulverizing debris with a chisel; mixture with water due to crevice expansion

subcriticality.
●

N2

Fuel Debris

up the detection process and enable estimation of the degree of

90cm

Water

Fuel Debris

making improvements (such as a highly sensitive germanium

At present, Xe-135 (which has a primary peak at 250 keV) is

Fission Yield

approached due to a perfect combination of fuel debris

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

3

▲ Estimating effective multiplication factor bymeasuring core noise with the Feynman-α method

Reference value for eﬀective multiplication factor (keﬀ)▲ Results of evaluating effective multiplication factor with
the Feynman-alpha method (Figure 2)

97▲ Applying the solidified-type adsorbent
High viscosity, so it covers the surface of fuel debris as
it changes shape(a) The test setup

Time/ｓ(b) Changes over time in the amount ofadsorbent injected▲Verifying the performance of underwater injection (Figure 5)98Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

Methods for criticality approach monitoring technology in the on-site environment
IRID has established protocols for monitoring if criticality is being

approached during fuel debris retrieval and formalized methods

for the on-site application of such technology. Monitoring fornecessary depending on how factors such as the work location

Different neutron detectors will be used to suit each of those

before and after the processing of fuel debris, and ③ continuouswith retrieval work as much as possible.

monitoring of fuel debris processing. Protocols are followed as❶ Ascertain conditions prior toName

Subcriticality detector

Type

Arm-type device positioned above debris

Neutron
Detector

Monitoring Protocol(Objective)

retrieval work

Based on the developmental progres s of equipment for

absorbent would solidify on the surface of fuel debris andprotocols for using an insoluble absorbent in spraying andprevent the generation of hydrogen). Therefore, its drying

negatively affect its drying behavior (the debris will be dried to

injecting devices. There were concerns that a water-glass

purposes, and they will be set in a way that avoids interferingcharacteristics using simulated fuel debris were tested.

Water-containing
porous material coatedwith water glass

Store raw material★Produce
absorbent❷ Monitor before and after fuel debris❸ Monitor neutron flux during

processing (determine when processing
begins/continues)

Measure raw material

fuel debris processing(detect unforeseen changes)

Criticality detector
Arm-type device installed alongside processing tool

Monitoring and
Measurement Methods

30〜100kg

30〜50kgLess than 30kg

MethodMeasure neutron ﬂux before and after theprocessing of fuel debris (for each ﬁxed
volume of processing such as 4,000 cm3) todetermine if criticality is approaching (or if
processing can continue).

Monitor neutron ﬂux during processing.
Detect signs of criticality (a sustained
increase in neutron ﬂux) and use them todetermine if the criticality danger has passed.

★Absolute-value measurement of theneutron multiplication factor★Transport
absorbent

Installed alongside processing tool and above fuel debris

M e a s u r e t h e d e g r e e o f s u b c r i t i c a li t y(Feynman-alpha method) only once prior to
the start of fuel debris retrieval to establish
the initial state.

Several hours to one week
(depending on site conditions)

About 10 minutes

Continuously during processing

Measurement
Points

One point in the vicinity of the retrieval work

Move with processing location as appropriate

Same as left as well as places
that overlook processing locations

Thermocouple(measures temperature of
the specimen)＊If water-glass
type

The upward-retrieval
process (example)

Scale(measures changes in thespecimen's weight)

Load device with adsorbent

15-L hopper

Toolchanger(a) Overview of drying test conducted in the crucible

Transport device to worksite★Inject
absorbent

Spray/Apply absorbent onto fuel debris
Enlargement of the surface

Return device

Nozzle shut-oﬀ lid(b) Observations of water-glass appearance
after drying test

Screw drive
motor★CleandeviceHeating causes cracks to form between chunks and particles of the water-glassmaterial, which suggests that the water content of the porous material could be
suﬃciently dried out. Moving forward, IRID will investigate this process further.

Clean device▲ Example protocol for the on-site

AppliedAppliedAppliedWhen removing obstaclesfrom within the PCV−AppliedAppliedWhen retrieving fromthe core−

AppliedAppliedExample of Side-access Retrieval

Example of Top Retrieval

injection of insoluble neutron▲ Injector for granular Gd absorbent

9Proposals for applying criticality-control technology applicable tothe fuel debris retrieval work
The table below contains several ideas for applying theextract large amounts of fuel debris and cause significant

factors such as the amount of fuel debris extracted and the

IRID will strive to make a variety of criticality-control technologies

changes in its state have a relatively higher risk of criticality, so

criticality risk for each retrieval method. Retrieval methods thatavailable for selection.

Smallalongside the processing tool

Tiny (a few g)

Retrieval method

Gripping, suction, etc.

Work
restrictions(①a.)
Monitor for
approaching
criticality (②)②Monitorfor
Insoluble neutron
approaching absorbent
criticality(①b.(2))
Soluble neutron
absorbent(①b.(1))❷：Detector monitors for approaching criticality
alongside the processing tool, both before and❶：Detector installed on the arm to measure subcriticality▲ Monitoring for approaching criticality when retrieving
PCV fuel debris from the side

99

after fuel debris is processed▲ Monitoring for approaching criticality when retrieving
PCV fuel debris from the top

Large

fuel debris retrieval

investigations

Retrieval amount①Prevent
criticality

Degree of change in fuel debrisInternal

Retrieval work❸：Detector continuously monitors▲ Drying test of porous material coated with water glass

absorbent

developed criticality-control technology, taking into account❷+❸：Detector runs alongside
the processing tool and continuously
monitors for approaching criticality before
and after fuel debris is processed

Crevices between
particles

Multiple cracks visible inthe surface

Nozzle shut-oﬀ
cylinderApplication points and stages for each monitoring protocol
When retrieving
accumulated material
from inside the pedestal
and around the lower
RPV head

Thermocouple(measures temperatureinside the heating furnace)

Explodedview ofinjector

Nozzle★Absolute-value measurement of theneutron multiplication factor

Measurement
Time

CrucibleMix raw material＊

Continuous-monitoring detector

Weight★Absolute-value measurement of theneutron multiplication factor

Using insoluble neutron absorbent at the plant
processing and collecting fuel debris, IRID has formalized

and operational steps affect the risk of criticality.

approaching criticality involves three steps: ① understanding thestate of conditions prior to retrieval work, ② monitoring both

8

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

7

Scope of retrieval gradually expanded
Small (a few kg)

Full-scale retrieval

Small(several to ten kg)

Full(several hundred kg per day)Core boring etc.

Core boring, chisels,
etc.

Gripping, suction, etc.

Core boring etc.

Method has no eﬀect
on fuel debris

・Limit the amount ofprocessing per shift
・Limit the distance between
retrieval sites

・Limit the amount ofprocessing per shift

・Limit the amount ofprocessing per shift(expanded)ーーMonitor neutron ﬂux

Monitor neutron ﬂux

Monitor neutron ﬂux

ー

ー

ー

ー

Measure degree ofsubcriticality 1

ー

ー

ー

ー

Insolube neutron
absorbent 2

ー

ー

ー

ー

Sodium-pentaborate
solution 3 (6000 ppｍ)

Monitor PCV radioactive gas
and neutron ﬂux

Monitor PCV radioactive gas
and neutron ﬂux

Monitor PCV radioactive gas
and neutron ﬂux

Inject sodium-pentaborate Inject sodium-pentaborate
solutionsolution

Inject sodium-pentaborate
solution

Method has no eﬀect
on fuel debris③Detect
criticality

Detectcriticality

Monitor PCV
radioactive gas

Monitor PCV
radioactive gas④Alleviate
eﬀectsResolvecriticality

Inject sodium-pentaborate
solution

Inject sodium-pentaborate
solution

1, 2, 3 Can be selected based on the speciﬁcs of retrieval work and on-site conditions. * The numbers ① through ④ correspond to the ﬁgure on page 96-2).

100

Ⅴ

Decommissioning
Research and Development

Retrieving Fuel Debris

Development of technology for containing, transfer and storage of fuel debris■ Constructed containing, transport, and storage scenarios and identified problem areas■ Developed containing technologies (prototype full-scale canister and structural testing)■ Developed transport technology (studied ways to predict the generation of hydrogen and deal with the gas)
■ Developed drying technology and systems (studied the expansion of basic specifications, acquired data, anddevice concepts for drying equipment as well as technology for measuring hydrogen concentration)■ Studied canister filters and topics related to the handling and storage of powdery, slurry, and sludge fuel debris

Research and development progress

1

Developing containing, transport, and storage scenarios and identifying issues
IRID developed scenarios covering everything from the sorting

long-term water quality maintenance, and the fact that existing

was done to develop technology related to developing andwas tentatively established as the goal of this technology

of fuel debris to managing facilities for its long-term storage. Thispool facilities are not expected to be diverted, and a scenario

using special canisters for safely and logically containing,development. Therefore several scenarios were created with thetransporting, and storing fuel debris at Fukushima Daiichi NPS.

goal of achieving that technological development. Based on the

Each scenario was developed by establishing hypotheticalimagined scenarios, IRID then studied the flow of debris handlingconditions based on the actual situation at the plant, taking into

up through its dry storage (Figure 1).

account the handling of radioactive material and safety design(subcriticality, heat removal, quarantining, shielding, structures

From that process, challenges (technological development) thatphenomena), materials (long-term integrity), hydrogen measures,storage of fuel debris were also identified. Finally, the results of(c o n s i d e rati o n of u s a b il i t y; re s ili e n c e a ga i n s t p o s s i b l emust be addressed to achieve the containing, transport, andfire prevention, throughput (processing capabilities) for the

all these investigations to select the areas that require continued

estimated amount of fuel debris, layout (required space), ease of

study (as technological development items) after FY 2019, frommainte nance, etc .). In ad dition to ove r s eas studie s andthe viewpoint of the safety designs studied thus far (subcriticality,

investigations on this topic conducted through FY 2016, otherheat removal, confinment, shielding, structures, materials,

related developmental data when creating the scenarios werehydrogen measures, fire prevention, etc.) were used.

used.

Those problem areas are listed below. For the results of

As a result of the study, it was confirmed that long-term dry

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Developing canisters forcontaining, transport,
and storage so that fuel debris can be stored safely
over the long term

technological development that addressed the below challenges,

storage of fuel debris is a promising storage method comparedplease refer to the pages that follow.

to wet storage (pool storage) in terms of safety, the need for
Challe nge s (Te c hn ologic al D evelo p m e nt Ite m s)

[Structural integrity of canisters] — Develop storage technology (prototype large-scale canister and structural testing)[Measures for hydrogen generated by fuel debris] — evelop transport technology (study ways to predict the generation ofhydrogen and deal with the gas)
[Plan to reduce the hydrogen generated by fuel debris] — Develop drying technology and systems (study the expansion of basic
specifications, acquired data, and device concepts for drying equipment as well as technology for measuring hydrogenconcentration)[Measures for powdery, slurry, and sludge fuel debris] — Study the handling and storage of fuel debris as well as the canister
filterReactor building<Area Classification>
Red: Highly contaminated area, unmanned operation (remote-operation from outside the area)
Yellow: Moderately contaminated area (area that is controlled but may be contaminated)
Unmanned operation (generally remote-operation from outside the area, but people may enter for
maintenance etc.).
Green: Lightly contaminated area (area with no substantial contamination) Manned operation

Example of handling flow for side-access method
Additionalbuilding Lifting machine

Manipulator

Lifting machine

Fuel debris treatment cell

Fuel debris
retrievalequipmentWork
monitoring
cameraBackgroundallows to safely and logically contain, transport, and store
retrieved fuel debris is needed. Because fuel debris contains

nuclear fuel, it is particularly important that radioactive
material (to prevent the spread of contamination) is sealed
and handled so that it remains in a subcritical state.

During the decommissioning of the Unit 2 at Three Mile
Island (TMI-2) in the U.S., collected fuel debris wascontained and handled in special containers (canisters) while
using existing technology for transporting and storing spent

fuel and for managing radioactive waste to keep the material

sealed, logically achieving the safety requirements. Employing
existing technology by utilizing canisters that fit the individual

101

Lifting machine

NPS. The latter was inundated with seawater at the start ofcomplexity, those such as the requirements for canisters andHandlingjigcanisters and related technology so that fuel debris can becontained, transported, and stored safely and logically at

Fuel debris treatment cell

Transport Container transport device
Weighing device container(surface contamination inspection)

Sealing cell

Transport standPreparation cell

On-siteStorage building
Area monitor

Work
Shieldedmonitoringcover
cameraCanister

Gate
Transport
container

Scaffolding
for work

Stored-item
verificationequipment

DryingsystemTransport
container

Verification area for
stored items

Drying area

Primary lid oftransport
container

Secondarylid oftransport
container

Vent pipe

Gas treatment system
Equipment f
or measuring
surfacetemperature

Transfer containerrackmethods for transporting fuel debris are also different. For

that reason, IRID has been working to develop special

Drying system and inside of canisterare highly contaminated.

Ventilation
systemthe accident, and pieces of its melted core have reached thebottom of the reactor pressure vessel. Due to this additional

Secondary-lid
Preparation
fastening system/
cell
gas injection system/leak check system

Hydrogen-gas processing system

PurposeThere are differences between TMI-2 and Fukushima Daiichi

Drying system Canister transfer system

Access tunnel

In the reactor

To decommission Fukushima Daiichi NPS, technology thatPrimary-lid fastening
system/gas injection system/leak check system /Transfer vehicle

UC transfer systemapproach.

Sealing cellGas injection
systemAccess tunnel

Fuel debris

realities of each situation should thus be an effective

Port

Canister-lid fastening system(conduct a surface contamination inspection)
CanisterDrying system

Lifting machine

Storage facility delivery
entrance for large items

Receiving area ①

Receiving area ②

Storagepreparation area

Storage area

▲ Example flow based on a scenario that involves containing, transferring, and storing fuel debris (figure 1)

Fukushima Daiichi NPS .

102

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

3

Developing transport technology andstudying methods to predict the generation of hydrogen

Based on the safety requirements for canisters, IRID developed

In addition, comparing the results of structural testing and

To develop transport technology, IRID

acquired knowledge

Based on those results, a method for calculating the energy-

manufacturing prototype canisters (Figure 3), and conducting

relative offset of the lid and flange with regard to maintaining

methods and exchanged the opinions from experts while

formula was propos e d to arrive at a pre dic tive metho dhousing technology by designing structural drop tests (Figure 2),
structural tests (Figures 4 and 5).

The results of the structural drop test on a full-scale canister
confirmed that the canister specifications and structural designs

developed by IRID are capable of maintaining safe conditions

from Japan and overseas about hydrogen-generation prediction

analyses allowed to verify the behavior when dropped, theabsorption rate using a linear model and particle-transport

studying the items that must be considered before transport

safety, the inner diameter of the canister body, and the fact thatappropriate for fuel debris. The amount of hydrogen generated

requirements can be established. Hydrogen generation using

deformation was generally consistent. In this way, the applicability

in a canister was estimated and the transport requirements thatspent fuel (Figure 6) was tested and methods of predicting thatof the analysis method was confirmed.

will facilitate safe transport were considered.

generation to verify the suitability of a method (linear model)

were studied for evaluating and confirming the validity of the(quarantining and subcriticality).

energy-absorption rate.

Pressure indicator

Test container
Crane

Crane

Crane

CanisterCanister

9m

9m

BasketManipulator

Pieces ofspent fuelpellet

Test waterBasket7m(a) Dropping canister in(b )Dropping canister atthe vertical position

A transfer
container orstoragecontainer

top of another canisteran angle（b）Test setup▲ Hydrogen-generation test using spent fuel (figure 6)▲Conceptual images of drop events (figure 2)

A hydrogen-generation test using pieces of spent fuel pellet was conducted to verify the feasibility ofA structural verification test was planned, in which a canister dropping event was simulated during handling.

the hydrogen prediction method.

4

Main dimensions

・Outer diameter: Approx. 3 m
・Inner diameter： Approx. 0.2 m
・Height: Approx. 1 m

Test container

Test container

Fragments ofspent fuel pellet（a）Conceptual image of the test

Canister(C) Dropping canister vertically on

Lid

Developing transport technology and studying hydrogen measures
The accumulation of hydrogen is a problem whentransporting fuel debris between buildings to theReaction
VesselGas

storage facility, and IRID has been studying methods

Measure hydrogenconcentration

Reaction
VesselGaseous mix ofhydrogen and oxygenof dr ying the d e b ris to re d uce the am o unt oftechniques that involve the use of a catalyst to rebind

Catalyst(installed internally)

the hydrogen to oxygen was also studied.

Thermostatic chamber

Catalyst

The catalyst was subjected to a circulation-style

reaction velocity test (Figure 7) to obtain data on its▲Concept for testing flow reaction velocity (figure 7)

performance at different environmental temperatures,

effects, and other areas. This testing verified that thecatalyst is an effective way to deal with hydrogen.

700

Even under the hypothesized conditions ofhydrogen generation, the catalyst’s performance was

Bufferstructure

more than enough, with the hydrogen concentration
staying sufficiently below the lower threshold for

explosion (4 vol%). Based on the acquired catalyst
data, the concentration distribution within a canister

was also evaluated and the best catalyst arrangement▲ Proposed canister structure (figure 3)

A prototype canister was designed and created to verify

103▲ Structural test (dropped at an angle) (figure 5)

The structural tests were conducted by using a full-scale canister(prototype) that were designed and manufactured.

for each quantity of generated hydrogen (Figure 8)
was determined.

600

Height (mm)te mp e rature, wate r vap o r, an d stro ng inte rnal

Total length of canister: 800mm

800

500
400
300
200
100
0
0

Max. hydrogenconcentration
0.85 vol%1

2

3

Hydrogen concentration (vol%)Φ220mm10mm

dependence on hydrogen concentration, toxicity

60mm

We tested the flow reaction velocity to verify the performance of the catalyst.

t h e p re s e n c e o r a b s e n c e of wate r va p o r, t h e

Gap between the lid
and catalyst layer

Catalyst layer(Diameter: Φ220 mm,
thickness:10 mm)

360mm▲ Structural test (dropped vertically) (figure 4)

360mm

Canisterbody

ExhaustMeasurehydrogenconcentrationhydrogen generated. As a backup plan, however,its structural integrity.

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 20222

Developing containing technologies + prototype large-scale canister andstructural testingΦ210mmFuel debris(The UC thickness isdisregarded and the outer
diameter up to Φ210 mm is
considered the fuel debris
layer.)
Canister4▲ Distribution of hydrogen concentration in canister (figure 8)

Taking into account the catalyst performance obtained in the test, the distribution of hydrogenconcentration was evaluated with regard to the arrangement of catalyst within the canister.

104

Decommissioning

ⅤRetrieving Fuel Debris

Research and Development

6

basic specifications, acquired data, and concepts of equipment
A technical issue that occurs when transporting fuel debris
between buildings to a storage facility is the accumulation ofhydrogen inside the hermetically sealed containers due to water

within the debris. Removing that water content via drying is thus
an effective way to reduce hydrogen generation.

equipment to dry the fuel debris at Fukushima Daiichi NPS,

method and operational requirements have on drying behavior
was verified.

Dryer

Condenser

HeaterCold trap

can be accomplished during the drying process and when inert

such measurements to be performed. From this research, it waswithin contaminated zones.

learned that constant monitoring is possible once debris has

handle material that is difficult to dry (Table 1) was also

and 10) was conducted. Through this, the effects that drying

studied constant monitoring in which the concentration ismeasures the hydrogen concentration when a process allows

factors such as retrieving fuel debris and maintaining equipmentof porous material. Parameters that were effective for drying

behavior was performed, and a full-scale drying test (Figures 9

differences in the thermal conductivity of gas species is a suitable

continuously measured as well as periodic monitoring thattaking into account the aforementioned results as well as other

Based on the fact that there is great diversity among fuel debris

were defined, component testing to obtain data on drying

With regard to measuring the hydrogen concentration within acanister prior to transporting it to the additioal building, IRID

IRID then studied proposals regarding basic specifications forSince the case of TMI-2 in the United States revealed that fueldebris is porous in nature, IRID focused on the drying behavior

Developing drying technology and systems andstudying technology for measuring hydrogen concentration

been stored in a transport container and that obser vingand its byproducts, the data needed for technology that canmeasurement method. It was also found that periodic monitoringgas is being injected; in both instances, applicable measurement

methods include proton conduction by way of a ceramic material
that selectively allows hydrogen to pass through it as well as gas
chromatography that utilizes differences in the adsorption anddesorption rate of a vaporized sample (Table 2).

▼Classifications of measurement points (work stages) for hydrogen concentration and assumed environments (Table 2)

Measurement points for hydrogen concentration were selected, then measurement methods were studied based on each assumed atmosphere.

expanded and supplemented and plans for testing that would

ascertain the drying properties as needed were devised toestablish drying device concepts and operational requirements(processes after storage inthe transfer container)that reflect the target water content and drying speed.

(drying process)Porous Material (Zeolite)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

5

Developing drying technology and systems; studying the expansion of(inert gas injecting process)7
Vacuum pump

To verify the effects on drying behaviors affected by the drying method▲Test piece in drying test (Figure 10)

in mind when handling powder y debris, approache s for

The following table lists drying target candidates. A test plan was designed to understand drying characteristics, particularly for materials that are difficult to dry.

Pebble-shaped andparticulate
fuel debris

fuel debris
retrievalprocess

Structural materials
withadhered nuclear fuelSlurry and sludge

Water treatment
ﬁlter

Gas treatment
ﬁlter

Details

Conceptual images

SUS

SlurryConcrete

Unmelted fuel assemblies
that remained in the reactor.

〇

−

−

−

Molten fuel that formed lumps
by slow cooling.

〇

−

−

Not yetacquired

Molten reactor-core materials
that became fragmented pieces
due to rapid cooling.

〇

〇

−

−

died

Powdered andﬁne particulate debris−

Filters with powdered andﬁne particulate debris adhered
to them−

Filters with dry debris powder
adhered to them−

1 Data for one set of conditions was acquired by FY2020.Additional data will be collected in FY2021-2022.

105〇

Problem areas and technology that must be developed for thetaking into account the results of case studies as well as a set of
32 processes (the aforementioned Figure 1) which assume thatcanisters will be similarly used to house, transport, and store
lumpy fuel debris.

▼Case study on storing powderized fuel debris (Table 3)

The technical issues were identified pertaining to the dry storage of powderized fuel debris and items for technological development were determined.

Status of data acquisition (-2020)
ZeoliteFuel debris adhered to structuresthat remained unmeltedWater
treatment
systemdry storage of powdery debris (Table 3) were also identified,referred to as “powdery debris” below), including factors to keep

▼ Data acquisition for drying target candidates (Table 1)

Lumpy fuel debris
and MCCI

abroad, then analyzed the experience, expertise, and data

storing powdery, slurry, and sludge fuel debris (collectively

and verified.

Unmeltedfuel debris

maintaining safety, and ways to reflect what we know in designs.

needed to construct a system for housing, transporting, andDrying tests were conducted using zeolite, a porous material.

and operational conditions, a full-scale drying system was manufactured

Generated By

IRID researched examples of powdery nuclear material andradioactive waste being handled and stored within Japan and▲ Appearance of full-scale drying test equipment (figure 9)

Type

Studying the handling and storage of powdery, slurry,and sludge fuel debris

〇

−

−

1 set ofconditions(ZrO2) 1−
Not yetacquired−

Not yetacquired

Not yetacquired−

Not yetacquired−

Scope of data expansion in FY2021-2022

8

Studying the canister filter
IRID examined the various

environmental conditions andrequirements for the canister filter,
identifying 13 different causes of filter
degradation (Figure 11). The effects

that these causes have on the life
span of the filter were studied andevaluated and the results of thatexamination were used to formulate ageneral strategy for testing filter life
span.

Hydrogenembrittlement

Radiation
degradation

Hydrogengeneration
Water content
Temperature
Time

Hydrogengeneration
Radiation dose
Water content
Hole in ﬁlter
Filter strength
Particle diameter
Particle density

Hydrogengeneration
Radiation dose
Water content
Hole in ﬁlter
Particle diameter
Particle density

Gas generation

Thermalalteration

Thermaldeformation(container)

Radiation
Temperature
dose
Filter strength
Irradiation
duration
Filter materialTime

Temperature
Container shape
Debris shape
Debris positioning incanisterParticle diameter
Particle strength

Thermalconvection

Frequency/
Amplitude
Acceleration
Hole in ﬁlter
Particle diameter
Particle density

Dust
generation
via vibration

ImpactTemperature
Filter strength
Thermal expansion
coeﬃcient of ﬁlter
Filter material
Thermal expansion
coeﬃcient of container

Internal/Externalpressurediﬀerential
ParticlediameterParticledensity

Particulate
dispersonWallthinning

Initiation stress
Filter strength
Debris composition
Debris shape (lumpy)
Container strength
Container material

Corrosive ion
concentration
Temperature
Humidity
Water content
Hole in ﬁlter
Time

Formation ofoxide ﬁlm

Electrolytic
corrosion

Corrosive ion
concentration
Temperature
Humidity
Water content
Time

Corrosive ion
concentration
Temperature
Humidity
Water content
Hole in ﬁlter
TimeUV
degradation

Container material
Filter material
Debris composition
Water content
Temperature
HumidityTime

Temperature
Humidity
External wind
speed
Aerosol diameter
Aerosol density

Salt
Accumulation
precipitationof aerosols

UV strength
Irradiation duration
Filter materialTime

Temperature
Water content
MicrobesTime

Microbeﬁlm

DamageBlockage
Temperature
Humidity
Water content

Water
ﬁlm

▲Causes of degradation in the canister filter (Figure 11)

IRID examined the environmental conditions and requirements for the canister filter andidentified 13 different causes of filter degradation.

106

IRID Research and Development

IRIDʼs History in Photos
2014

May
Veriﬁcation testing of the elevated decontaminationequipment (a dry-ice blaster) on the ﬁrst ﬂoor of Unit 3

March

Conducted demonstration tests on the suction/
blasting decontamination equipment

Elevated decontamination equipment▶

Suction/Blasting decontamination equipment▶

April

2017

Conducted veriﬁcation testing on low-lying
decontamination equipment (dry-ice blaster)Conducted veriﬁcation testing on
low-lying decontaminationequipment (high-pressure water
sprayer)

May

March

Surveyed the PCV interior of Unit 1with PMORPH 2

Investigated Unit 1ʼs suppression
chamber with the upper-S/C
investigation equipment

PMORPH 2▶

The upper-S/C investigation equipment▶

May‒September
Measured Unit 3 with the muon transmission method
Unit 3 under observation▶

Floor-traveling robot▶

Surveyed the wall surfaces ofUnit 2ʼs torus room witha submersible robot andone that travels across the ﬂoor

June

Conducted full-scale testing ontechnology for plugging leaks inthe suppression chamber (S/C)

Submersible robot▶

September
Investigated Unit 2ʼs suppression chamber withthe lower-S/C investigation equipment

July

Surveyed the PCV interior of Unit 3
with a submersible robot (ROV)The ROV▶

The lower-S/C investigation equpment▶

2015
February‒May,

May‒September

Measured Unit 1with the muontransmission method

Unit 1 under observation▶

Investigated the PCV interior of Unit 1 with
PMORPH 1

2016

managing and storing solid waste inconsideration of the characteristics of the waste
generated from the accident. The goal was todevelop plans for treatment and disposal

approximately by the end of FY2021 and to

Investigated Unit 2ʼs PCV witha telescoping investigationequipment▲Inside Unit 2ʼs PCV▲Telescoping investigation equipment

2022

March‒July

Measured Unit 2 with the muon transmission method

Unit 2 under observation▶

February

Conducted a detailed investigation Unit 1ʼs
PCV interior with IRIDOLPHIN

April
Completed full-scale testing at JAEAʼs Naraha Center
for Remote Control Technology Development
Full-scale test unit▶

▲IRIDOLPHIN

Treatment andDisposal of
Solid Waste
c o n c e p t s o f t re a t m e n t a n d d i s p o s a l f o r

2018

PMORPH 1▶

Decommissioning Research and Development

IRID has studied safety evaluation methods andConducted full-scale testing on
reinforcement technology forthe suppression chamber (S/C)
support columns

January

April

107

Ⅵ

Scorpion robot▶
▲Dry-ice blaster

High-pressure water jet decontamination equipment▲

July

February

Investigated the PCV interior of Unit 2
with the Scorpion robot

make technical perspectives related to thesafety of the waste treatment and disposal.

Ⅵ

Decommissioning
Research and Development

Treatment and Disposal of Solid Waste

Research and development of treatment and disposal of solid radioactive waste

IRID’s research and development on the treatment and disposal of solid waste has focused on integrating the following four
major areas of study, which cover everything from understanding the properties of waste to handling its disposal.
■ Ascertain waste properties: Studying the solid waste generated during decommissioning activities requires to gather
information on the properties of that waste. Develop technology for remotely collecting analysis samples, techniques foranalyzing them, methods for evaluating that analysis data, and tools for publicizing the data

■ Study disposal technology: Solid waste will be processed so that its radioactivity can be sealed away and it can be isolated(disposed of) in a place unaccessible by people. Study improvements to safety and confinement that can be implemented
beforehand so that solid waste can be handled safely with an eye toward disposal■ Study disposal concepts and safety-assessment methods: Concepts for disposing of waste will be chosen based on theproperties of that waste. Study disposal concepts and develop (safety-assessment) methods for determining if those concepts
are suitable■ Study storage and management methods: Solid waste contains things like highly radioactive waste as well as material
generated during the processing of contaminated water (secondary waste from water treatment). Study methods for safely
storing such waste

Research and development progress

IRID’s research and development process for understanding the properties ofsolid waste as well as storage/management and treatment/disposal
Radioactive waste can be a solid, a liquid, or a gas. IRID has
worked to ascertain the properties of solid waste and to researchand develop the storage/management and treatment/disposal of

expedient analysis methods, and developed methods forcollecting highly radioactive samples for analysis.

such waste, with the goal of reducing the risk of radiation

I R I D d e s i g n e d a m e t h o d fo r s e l e c t i n g t h e a p p l i c a b l ethere is a wealth of past experience, the waste generated by thesolidifying slurry waste were studied, which is unique to 1F for

exposure. Compared to conventional radioactive waste for which
1F accident has never been dealt with before, and it has some
unknown properties. It also has different types and comprises alarge amount of material.

The areas where the results of all of IRID’s research anddevelopment will be applied are systematically depicted in the

figure below. First, traditional waste-management practices wasreferenced to apply existing disposal concepts such as estimating

the solidified objects present and assessing the instantaneous
release of radionuclides at the disposal site. There are clearlysome issues, however, such as the large variance in waste

solidification methods as a preprocessing technique. Methods forwhich there is little knowledge.

A basic safety assessment was conducted and existing data usedby methods of assessing disposal safety was organized. Based on
those results, potential disposal concepts were reexamined andinvestigating the utilized scenarios and parameters continued.
The results of the safety assessment were then passed back toeach R&D area. With waste categories, for instance, the mostproblematic waste were further narrowed and reestablished andthe new knowledge needed for safety assessments were applied,such as selecting the important chemical substances in waste.

properties and the increasing number of disposal categories. It is

M eanwhile, data was gathe re d and pre pare d for future

approaches that reflect overseas and domestic knowledge—such

understanding waste properties by analyzing collected samples.

now appare nt that it is ne e d e d to intro duce pragmaticas the formation of solidified objects and the expansion ofexisting disposal concepts — in addition to the handling ofassessments while aiming to improve the analysis precision for
The collected data was made available worldwide as a database.

storage and management.

Through this process, an understanding of the waste’s properties

To understand the waste’s properties, IRID first created somewaste based on those properties were presented, and a method

general categories for waste, designed methods for estimating
inventory and plans for identifying and analyzing quick andPro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Safely managing and treatment of waste

1

was gained, safe and rational methods for storing and managing

for logically selecting the preprocessing method was designed.
The main results will be introduced on the following pages.

Comprehensive Results of Research and Development
Established tentative waste categories and evaluatedthe amount of hydrogen generated by waste basedon waste data extracted from studies on the fueldebris retrieval process

Developed methods for estimating
inventory, designing analyses, quickly andeasily performing analyses, and collecting
highly radioactive samples

Studied storage andmanagement methods●

Just as with regular waste, radioactive waste generated

At the end of FY 2021, the Nuclear Damage Compensation

as much as possible and handled so that the useful portions

published a technological forecast related to policies for

Conducted on-site validation testingto verify that the newly developed
measurement device is sensitive toalpha radiation in an environmentthat is widely contaminated and that
it is greatly aﬀected by noiseof those policies. At IRID, the goal was to work on

1 Structure of processing technology, operating
conditions, etc.

during the decommissioning (1F waste) must also be reducedof it can be reused and the nonreusable portions can bedisposed of properly. This 1F waste has different properties
than the radioactive waste that originates from traditional

nuclear power plants. For example, there are many types of
1F waste, including the rubble that was dispersed by ahydrogen explosion immediately after the accident, trees

and shrubs that have radionuclides stuck to them, soil, etc.

There is also secondary water-treatment waste such as theabsorbent used to remove radionuclides from contaminatedwater. These circumstances make it difficult to study safety
management and disposal policies for 1F solid waste.

109

PurposeApplied to 1F storage andmanagement methods

and Decommissioning Facilitation Corporation (NDF)
processing and disposing of solid waste as well as the safety

developing processing and disposal methods so that 1F

waste can be safely stored and disposed of. Progress has
been made on a variety of technological development

related to 1F waste with the goal of presenting the NDF witha technical foundation for understanding the properties ofsolid waste, processing/disposal, and storage/management.

2 Set requirements for more logical techniques,
such as waste speciﬁcations and focused
nuclides3 Combined processing technologies withdisposal concepts that could be applicable●●

Techniques forascertaining
waste properties

Techniques forassessing the safetyof disposal

Techniques for assessing
the applicability ofprocessing technologies

Waste properties

Potential disposal
concepts
Scenario
parameters

Waste speciﬁcations

Derived the absorption
reduction factor using the six
primary chemical substances
contained in waste

Assessment ofdisposal safety

Fed back to each study 2

Established preprocessing option

Background

Developed technology forassessing contamination
when sorting solid waste

Designed selected methods
Established stream option 3●

Studied and evaluated storageand management methods

Redeﬁned the method for assessing radiationdose and used that new method to consider
representative waste disposal categories

Data onprocesses etc. 1
・Designed inspection techniques for determining if
solidiﬁcation is possible as well as an approach forevaluating the feasibility of processing technologies
・Conﬁrmed the maximum ﬁll rate for slurry during the
low-temperature processing of simulated waste(carbonate: about 40 mass%; iron co-precipitation:
25 mass%)

Proposed safe and logical storage and management methods based on waste properties
Designed techniques for selecting preprocessing methods in a logical manner* 1F: Fukushima Daiichi Nuclear Power Station

110

ⅥTreatment and Disposal of Solid Waste

Decommissioning

Research and Development(understanding the properties of waste)

IRID analyzed waste and contaminated water with samples

via the Web-based FRAnDLi database (Figure 1) along with(TEPCO), transporting them to analysis facilities located primarily

data points amassed by the end of FY 2020 were put to use inprovided by Tokyo Electric Power Company Holdings Inc.

relevant data published by TEPCO. The approximately 12,400

in Ibaraki Prefecture. The analysis samples covered a wide range,

areas such as IRID’s study of the storage/management andincluding rubble, soil, vegetation, contaminated water (bothprocessing/disposal of 1F waste.

accumulated water and treated water), secondary waste createdthrough the treatment of contaminated water, and more. ThisFor example, data on nuclides that contribute to the radiation“difficult to measure” nuclides that emit alpha and beta radiation.

hydrogen generation, which is an important factor in the storage

project is notable for also attempting to analyze the so-called

dose and heat generation of waste is useful for evaluatingU s i n g e x i s t i n g m e t h o d s a s a f o u n d a t i o n , a p p ro p r i a t eof waste. Data on each nuclide that has a long half-life will also

improvements at each research facility were made to arrive at

be instrumental in conducting safety assessments from thethe analysis methods for 1F waste. The analysis results for theperspective of treatment and disposal.

gained understanding of waste properties have been released

3

Researching and developing the methods for stabilization and solidification processingthat are expected to be applicable to actual processing (studying processing technology)
This section introduces the results of IRID’s studies regarding

w a s t h e n e x a m i n e d , a c q u i r i n g t e c h n i c a l d a t a t h ro u g hwate r t re at m e nt , w h i c h i s p a r t of t h e s o li d wa s te t h atcomparison charts (Table 2 and Figure 3). Through this,technology for processing the secondary waste produced by

exp e rime ntation, and organize d that data into a s et ofcharacterizes 1F.

techniques that allow multiple technologies to be selected forthe preprocessing disposal methods used with the secondary

The primary materials currently being generated are slurry fromwaste produced by water treatment were established.

the multi-nuclide removal system (ALPS) and waste zeolite
produced by the cesium adsorbers. IRID has worked to develop

The waste generated from now on will be nonuniform andsolidification technology for these target materials. Four highly

inspection techniques that allow to check whether cement or AA

a method for conducting versatile assessments of applicablecontain a variety of materials. IRID has developed simple

mature candidate technologies employed in domestic andsolidification is possible when actually processing on-site and tooverseas waste dispos al were sele c te d, ke eping on -site

study which mixtures will ensure quality when performing

installation in mind: cement, alkali activated materials (AAM),

solidification processing. The applicable scope of waste has been

melting, and glass. A ssessment criteria for comparing the

expanded and criteria for judging the quantitative applicability

technologies was also established (Table 1).
Data updated
and uploaded■ Readable at a glance, the streamlined interface condenses the categories for waste and analysis items.

of processing technologies have been studied.

Referring to standards related to current regulations etc.,
information related to each piece of solidification technology

Possible
Partially Impossible
Impossible

Manager▼ Established focus of evaluation (Table 1)

Web server
PHP program
CSV data
User

Searching the FRAnDLi database

Data viewed

and downloaded

FRAnDLi structure and features▲Overview and features of the FRAnDLi database (Figure 1)

Evaluation Item
Technical Results
Process Performance

Processing speed,
Cs volatilization rate, etc.

Operability and Safety

Process risk, maintenance details, etc.

Cost-Eﬀectiveness

Volume reduction, construction area, etc.

2

Solidiﬁed Matter-2

Developing sampling technology by using equipment for collecting highly
radioactive analysis samples (understanding the properties of waste)

Evaluation Sub-Item
Development stage,
operational achievements, etc.

Properties ofImpermeability, resistance to radiation, etc.

Technology

understand waste properties.

understanding the properties of 1F waste. IRID therefore

Target Secondary Waste from
Water Treatment → Solidiﬁed Productdeveloped equipment that collects samples while working to

As with the cesium adsorbent, collecting samples for analysis canremotely reduce radiation exposure. Figure 2 depicts a drawing

be difficult because they are highly radioactive and inside aof the sample collector used to obtain samples of the cesium

sealed container. Regarding the waste from the cesium adsorbent

adsorbent tower. The sample collection equipment can create

tower, the radionuclides included in the adsorbent and their

Processing Temperature
Process

Performance

opening in the top of the absorbent tower, collect samples, thenco nce ntratio ns are e xp e c te d to und e rgo large change s

seal the opening.

depending on operating conditions such as the t ype of

MaintenanceFrequency

Sample Collection Equipment
Cost-

EﬀectivenessProcessing Speed

Process Risk▲

AAM Solidiﬁcation

Carbonate slurry→ cement solidiﬁcation

Carbonate slurry→ cement solidiﬁcation

Carbonate slurry→ AAM solidiﬁcation(outside of drum)(in drum)

Room temperature

Room temperature

Room temperature

0.12 t/h0.24 t/h0.12 t/h

Does not volatilize in principle
Quickly hardens when cement contains
2% or more of sodium carbonate

Quickly hardens when cement contains 2%
or more of sodium carbonate

None in particular

Maintain meter calibration anddustproof facilities

Maintain meter calibration anddustproof facilities

Primary Solidiﬁcation
Equipment

Mixing facility(stirring blades, agitators)

Mixing facility (stirring blades,
agitators, mixing container)

Mixing facility(stirring blades, agitators)

Consumables

Stirring blades (per batch);
dust ﬁlter (periodically)

Stirring blades (periodically);
dust ﬁlter (periodically)

Stirring blades (per batch);
dust ﬁlter

None

None (reused in cleaning water)

None in particular

7 (ﬁll rate 30%; 28 days)

5 or more (ﬁll rate 30%; 28 days)

0.16-0.17

0.05-0.28

Properties of

G value [1/100 eV]

MatterImpermeability

Cs: 96%, Sr: 5%, Sn: ND, Ce: ND (assess leaching rate with ANS/ANSI-16.1)

Cs: 24%, Sr: 0.5%, Sn: 5%, Ce: 0.5%
(assess leaching rate with ANS/ANSI-16.1)

Heat ResistanceWithout decreasing in strength even at 80 °C drying

40-50% decrease in strength even at R.H. 60%,
which is less drying strength than 80 °C drying

Solidiﬁed

111

Cement Solidiﬁcation(in drum)

Maintenance Frequency
and Details

Strength[Mpa, N/mm2]

The sample collection equipment for cesium adsorbent towers

Cement Solidiﬁcation

Maintain meter calibration anddustproof facilities

Secondary Waste
Generated

Absorbent Tower

slurry) (Figure 3)

Cs Volatilization Rateand Details

Sample Collection Equipment(possible scope of cement solidification for carbonate

the comparison chart for low-temperature solidification processing technologies) (Table 2)adsorbent, the operating time of the device, and the flush
duration. This is why analyzing the adsorbent is essential for▲Comparison table for compiled technologies▼Comparison chart for compiled technologies (excerpted from

This se c tion introduces te chnolog y for colle c ting highly
radioactive analysis samples, which is part of the initiative to

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

2

- 1 Releasing data on the understanding of waste properties as a database on the web112Decommissioning

ⅥTreatment and Disposal of Solid Waste

Research and Development

5This section discusses IRID's efforts to research and develop theour disposal categories such as trench, pit, burial at a medium

task of disposal.

characteristics of the various waste types that are available was

depicted in Figure 4. First, IRID referred to the properties ofThis list of requirements includes measures for dealing withconsidered. The goal was to be able to suggest specific disposal

accident waste that have already been identified then classifiedhydrogen generation, which has been identified as a major

concepts and safety-assessment methods for these eight waste

the waste into 27 comprehensive types. Next, a preliminary

concern when it comes to safety assessments. Even in the casetypes. As noted in section 1, the results of this disposal researchsafety assessment was conducted, taking into account thewater content, radioactivity, and how long radiation has beenwater that is contained within individual pieces of waste isdifficult to do on the work site, thereby deciding that drying thesolid waste is a logical strategy for preventing the production ofhydrogen. The concept for how to dry the waste was thenactualized.

Drying Method (with vent)of conventional storage containers, ensuring that they remain

were fed back into studies on understanding waste properties

hypothetical application of each of those waste types to one ofinteracting with the water. However, assessing the amount ofconsidered.

radioac tive conce ntration, phy sic al trait s, and che mic al

the time being. The procedure for studying disposal concepts is

pro duce d whe n fuel d e bris is retrieve d, the n prop os e dcompiled and candidates that fulfill those requirements were

When making the selections, how much information about the

reduce the volume of waste, tasks which will be performed for

The amount of hydrogen generated depends greatly on thestorage. A list of desired features for the storage container was

representative types of waste was selected for further study.

other areas such as recycling as well as processes to stabilize and

IRID studied methods for safely containing, transporting, andhypothetical scenarios for the process of handling it up throughTo i nve s ti gate d i s p o s a l c o n c e pt s i n m o re d e t a il, e i g ht

future, disposal must be studied while simultaneously exploring

during fuel debris retrieval (studying storage and management methods)

storing the highly radioactive waste (reactor internals etc.) that is

depth, or deep geological disposal.

Since waste that cannot be reused will be disposed of in the- 2 Studying methods for storing and managing the highly radioactive waste generated

Vent Filter

sealed is an essential safety step for preventing the dispersion ofand techniques for storage and processing.

radionuclides; the emission of hydrogen gas, however, requires
to also keep them ventilated, so a plan that achieves both isneeded. IRID looked at foreign and domestic examples of highly

27 Waste Types

Allocating Nuclear Waste

KurionAREVA
Carbonate sediment slurry
Rubble 1
Rubble 2

Preliminary dose
assessment

RPV waste(before accident L1)
PCV internal metal(before accident L2)<Trench Disposal>
Waste A, B...

Characteristics ofeach waste type

Study disposal concepts;
evaluate dose

<Pit Disposal>
Waste C, D...

ProcessA

Disposal
X

Process
B

DisposalY

Process
C

DisposalZ

pipes have been employed in cases where hydrogen gas is being
a storage method was also formalized.

- 1 Developing technology for stabilizing (removing water from) the slurry generated

by the multi-nuclide removal system (studying storage and management methods)

Of the 1F waste, the secondary water-treatment waste carries a

relatively high storage risk. In particular, the multi-nuclide

radioactive concentration.

5

solid waste (studying storage and management methods)

Since α particles have severe effects on the human body whenoverlooked when using this method, so it is difficult to know themust be stored and managed with even more caution. It ismeasurement.

Fortransporting
HIC etc.

Dustproof area
(controlled withnegative pressure)d istrib utio n of co nta minatio n a cro s s th e e ntire a re a ofcontamination for the waste generated during the demolition of

To make it easier to sort waste that is contaminated with αpreliminary steps for the retrieval of fuel debris at Fukushimacomprehensively measure the α contamination on a target

structures and removal of equipment is ascertained — which are

Daiichi NPS — and reflect that information in the operational
plans and methods for managing the work.

quantitative determination made with a smear sample. However,there is always the possibility that a hot spot etc. will be
α Camera(includes shielding unit)

reducing the storage risk of ALPS slurry, suggested some basic

TOF* camera

equipment designs, and demonstrated the feasibility of using

Detector(PMT*)

radiation, IRID is working to develop technology that cansurface in a short period of time. Efforts are now focused on
improving the developed device (alpha camera), for example byb r i n g i n g i t i n to t h e 1F s i te a n d p e r fo r m i n g s o m e te s tmeasurements. Figure 7 contains a photo of the machine, while

the principles behind how it takes measurements are depicted in
Figure 8.

Visible-Light Filter
Mirrorpressurized filtration and disc drying in slurry stabilization

γ Shielding
Unitin the design of the water-removal process that uses pressurized

Lens

processing (Figure 5). The results of this research were reflectedfiltration, which was suggested as a candidate technology, as

TEPCO works toward the actual on-site implementation ofstabilization processing for ALPS slurry.

Storage
保管容器
Container- 3 Developing technology for assessing contamination with regard to the sorting of

A lp ha co nt a minatio n is tra d itio nally ve rifi e d th ro u gh a

IRID thus studied the applicability of dehydration technology for

Injector

UV Radiation

Dehydrator
Slurry extraction and transport equipment▲The pressurized filtration process (Figure 5)

Pan/Tilt and Lift
Mechanisms▲The measurement device (Figure 7)* TO F: tim e of flight

113

Storage
Container▲ Foreign and domestic examples of storing highly
radioactive waste (Figure 6)

therefore imperative that the presence and degree of alphamaterial characterized by its fluid nature and comparatively highDust (contains
alpha particles)

Dried Waste

Waste

generated (see Figure 6). The concept for a vent filter system as

they enter it, waste that is contaminated with alpha radiation

remover generates large amount of slurry (ALPS slurry), a

HydrogenControlled

Waste B

▲ Process of studying disposal concepts (Figure 4)

5

Dust (contains
alpha particles)

Hydrogenradioactive waste being stored, verifying that vent filters andProposed Disposal
Concepts

Waste A

Drying Method (no vent)

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

4

Developing disposal concepts and safety-assessment methods based on
disposal technology that can be applied to solid waste(studying disposal concepts and safety-assessment methods)

EmittedLight

α Radiation
Uranium etc.

■ Measurement principles
・The interaction of α particles with nitrogenatoms produces ultraviolet radiation

・This UV radiation is focused with a lens tocreate an image; the distribution of the UV

radiation that reaches the photodetector is thenused to measure the distribution of α radiation
・Sinceγ radiation that reaches the detector

would interfere with the measurement of alpharadiation, the equipment is shielded to prevent

γrays from entering

※TO
F…Tim
▲ Principles of the alpha camera(Figure8) e O f Flight* PMT: p h oto multiplie r tub e114

R&D Cooperation with Oversea Organizations

Strengthening relationships and cooperation with international organizations

The International Advisory Meeting
The International Advisory Meeting is comprised of three foreign nuclear experts to advise on organizational operationand management.

International Advisors
USA

IRID is accelerating our R&D with entities overseas, and constantly introducing the latest technology
for the nuclear decommissioning.

Mr. Lake Barrett

Spain

Mr. Luis E. Echavarri

Technology and

the Nuclear Regulatory

NEA

at the Nuclear

Commission (NRC)d u rin g th e Th re e M il e

IRID engages in technological collaboration

UK・Sellaﬁeld Ltd: Handling and safe storage of damaged fuel;controlling criticality
・National Physical Laboratory: Developing technology forcalibrating radioactive-gas monitors
・Currently developing a robot arm together with
Veolia Nuclear Solutions (VNS)with overseas nuclear organizations

that have experience in nuclear decommissioning
and damaged fuel.

France・Atomic Energy and Alternative Energies Commission (CEA): MCCI testing

Prof. Melanie Brownridge

Former executive(former site director for

Island accident)

Initiatives involving R&D with overseas organizations (list of main projects)

UK

Independent consultant

director of the OECD/(has experience with the
IAEA's International Nuclear
Safety Group (INSAG) etc.)

Innovation Director
Decommissioning
Authority (NDA)in the UK

The 8th International Advisory Meeting
Date: May 25, 2021 Venue: Online

The international advisors reviewed the progress of IRID activities
since FY2020 and provided advice regarding the status of the R&D

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

Based on our management policy of "open frameworks," IRID is
striving to disseminate information such as the R&D results anddevelopment while strengthening our relationships with foreign
experts and research organizations.

design reviews that IRID is currently conducting, a framework forconveying knowledge, and the future activities.

▲ Meeting of the International Advisory Meeting (IRID office)

Hungary・Paks Nuclear Power Plant: Handling and safe storage of damaged fuelParticipating in international conferences
IRID has presented the results of our R&D via the forums hosted by international-relations organizations.

Kazakhstan

USA

National Nuclear Research Center (NNC):
Testing to ascertain the characteristics of fuel debris

・Paciﬁc Northwest National Laboratory: Handling and safe storage of damaged fuel・Hanford Facility: Handling and safe storage of damaged fuel;Austria・IAEA: Compiling technical data

studying ventilation systems for fuel-debris dust・Idaho National Laboratory: Gathering information on the experience gained from TMI-2

・Argonne National Laboratory: Exchanging information on our understanding of fuel-debris characteristics
・Los Alamos National Laboratory: Developing technology for detecting fuel debris
・University of California, Berkeley: R&D on the treatment and disposal of solid waste
・Mississippi State University: Studying ventilation systems for fuel-debris dust

French Atomic and Alternative Energies
Commission (CEA)

Through our international joint research with the CEA, the MCCI
simulated test was conducted to test for the reaction between

STSS / ISOFIC / ISSNP2021
Date: November 15, 2021
Venue: Okayama Convention Center (hybrid of online / in-person)

In Japan, a series of three international conferences were held—

the STSS, ISOFIC, and ISSNP — related to interactions between

humans and computers, cutting-edge systems for controlling

instrumentation, and nuclear-power systems that share a

relationship with society.

Idaho National Laboratory (INL), USA
A s par t of our proje c t s to study s afet y re quire me nt s and

STSS:International Symposium on Socially and Technically Symbiotic Systems
ISOFIC: International Symposium on Future I&C for Nuclear Power Plants
ISSNP: International Symposium on Symbiotic Nuclear Power Systems(under IRID's project, "Developing technologies for containing,appearance and separate layers of oxides and metals.

INL to exchange information and engage in discussion. IRID thustransporting, and storing fuel debris"), a workshop was held at theCondition before dismantlement(after removal of the upper cylinder)

Metallic particle

Attendees primarily from Japan, China, and Korea joined thecriticality is occurred when retrieving fuel debris. In order to improve

R&D Planning Department, delivered a special lecture titledconcentration of Kr-88 and the amount of γ rays measured. At the

NPL, a world-renowned authority that determines standards, actual
Kr-88 was generated then methods for calibrating detectors were
Metal layer on the bottom

115

Condition after removing most melted material

established.

Strategy Planning Department each participated in one of two panelsessions, giving presentations on the status of IRID's research anddevelopment. The meeting was attended by researchers and experts

from around the world. Mr. Lake Barrett, one of IRID's international(ICRP) as the main organizer , this online meeting focused on

IRID is working to develop detecting technology that uses Kr-88 to

the Kr-88 measurement accuracy, it is necessary to relate the3, 2021 and organized by the American Nuclear Society (ANS).

IRID's President Yamauchi and Senior Manager Okuzumi of the R&D

Led by the International Commission on Radiological Protection

National Physical Laboratory (NPL), UK
recognize approach signs at an earlier stage in the event that

The ANS Winter Meeting was an international meeting on nuclear

engineering held online from November 30, 2021 through December

Date: December 1, 2020 (lasting through December 4)
Venue: Online

with TMI-2.

Concrete boundary layer

Date: December 2, 2021
Venue: Streamed from Washington DC (online)

International Conference on Recovering After Nuclear Accidents

knowledge and lessons learned by US experts who had experience

Glass and orange-colored layer

ANS Winter Meeting〈U.K.〉International Commission on Radiological Protection (ICRP)

gleaned a large amount of valuable data from the specificCondition of MCCI test products〈United States〉American Nuclear Society (ANS)

advisors, also participated in the panel sessions.

specifications for containing, transporting, and storing fuel debris

molten fuel and concrete . The te st reveale d s ome of thecharacteristics of the MCCI test products, including its porous〈Japan〉

conference virtually. Naoaki Okuzumi, a senior manager in IRID's"Radiological-Protection Lessons from Fukushima and Beyond." Themeeting was held through the cooperation of a number of domestic
and international organizations as well as the Japan Atomic Energy

Agency (JAEA). President Yamauchi of IRID presented the current"Overview of IRID's Research and Development." During thestate of technological development with regard to retrieving fuelOkuzumi discussed the current state of investigations inside PCV

190 people from various countries. In February 2022, a summary oftechnological development for retrieving fuel debris.

org/index.asp).

lecture, approximately 120 attendees listened eagerly as Mr.
for the nuclear decommissioning as well as the status ofdebris at Fukushima Daiichi NPS. The lecture was attended by overthe lecture was made available on the ICRP website (https://icrp.

116

Human Resource Cultivation in R&D

Therefore, IRID hopes that the younger generation take more interest

Participation in various events
IRID actively participates in academic conferences, events, lectures organized by various organizations.

in the decommissioning and engage in the decommissioning industry.

IRID has been working to actively provide universities and other academic institutions
that conduct research related to nuclear decommissioning with information about

Fukushima Daiichi NPS, an urgent issue that IRID has faced, and advanced initiatives
to lead caltivating human resources for the future.

Robotics Society of Japan

Vice President of IRID, Dr. Arai gave a lecture[Robots Utilized for Reconstruction of Fukushima]

Developing robots for decommissioning Fukushima
Daiichi NPS—Challenges and required technology

129th Robotics Engineering Seminar
Date: October 21, 2020

Venue: Online seminar only

Program■ Session 1: Robotic and remote-operation technologies requiredfor decommissioning Fukushima Daiichi NPS
Prof. Hajime Asama, The University of Tokyo■ Session 2: Developing robots for decommissioning Fukushima
Daiichi NPS: Challenges and required technology
Dr. Tamio Arai, IRID

The IRID Symposium
IRID Symposium 2021

"Challenges for Fuel Debris Retrieval Ⅳ"
Date: December 8, 2021
Venue: Exhibition Hall, Iwaki Business Innovation Center (LATOV),
Fukushima Prefecture

Dr. Tamio Arai, IRID

Many different robots have been introduced at TEPCO's Fukushima Daiichi

NPS. Starting with investigation of primary containment vessel (PCV) includingin the partial-submersion / submersion method, currently detailed investigation
inside PCV has been conducted and then test retrieval of the fuel debris will
be prepared as a next step.

and larger robots with large weight capacity. I would like to talk about the■ Session 3: Developing and field testing remote-imaging
technology for visualizing the radioactive material dispersed by the
Fukushima Daiichi NPS accident
Dr. Yuki Sato, JAEA

robotic technology and the peripheral equipment that help achieve their■ Session 4: Improving the monitoring of environmental radiation
with drones: From the response to 1F to application as a nucleardisaster tool
Dr. Yukihisa Sanada, JAEA

with using robotics for decommissioning in the hope that the latest technology

many functions, as demonstrated by the decommissioning processes ofretrieving fuel debris and removing obstacles. I will then discuss the challenges

from the field of robotics can be introduced to the world of decommissioning
as well.

Atomic Energy Society of Japan

Organized by IRID, this symposium was held to report the

ANSJ Fall Meeting 2021

technological development aimed at the retrieval of fuel debris. Over 400 individuals

Date: September 9, 2021
(online meeting)latest research results and to cultivate young researchers and engineers, focusing on
participated in the FY2021 symposium, most of whom joined via internet. In theManager Okuzumi of the R&D

contribute to decommissioning, reconstruction, and technological development

Strategy Planning Department gave athrough the difficult task of retrieving fuel debris.

Fukushima Prefecture

Fukushima Robot & Aerospace Festival 2021

Despite the restrictions by▲ N. Okuzumi,

Senior Manager of IRID.

h o l d i n g t h e e ve n t i n t h epresentation on the current state of IRID's R&D as part of an activity

COVID-19 pandemic, around

Investigative Committee.

family members during there p o r t f ro m t h e Fu k u s hi m a D a ii c hi N P S D e c o m m i s s i o ni n g

Student presentations of research results

Reporting session
At the reporting session for IRID's R&D,

M a ny s t u d e n t s p re s e n t t h e i r re s e a rc h a t t h e

IRID's R&D Strategy Planning

students made such presentations.

Kenro Takamori, General Manager of

symposium every year. In FY2021, a total of seven

Department, summarized overviews ofcontainers.

Decommissioning poster session▲The reporting session

Presentations

by students

Panels presenting the latest

82nd JSAP Autumn Symposium

event's host). Many of themals o visite d IRID 's b o oth,

▲ At the IRID booth

which displayed panels introducing the current state of IRID's

R&D on the decommissioning of Fukushima Daiichi NPS and

Department reported the current status of IRID's R&D during anplayed a video showing the submersible robots currently being

decommissioning of Fukushima Daiichi NPS and the recovery of the

UK and is now being tested in Japan.

activity reported entitled"What can JSAP members do for thedeveloped as well as the Robot Arm that was developed in theprefecture?"

information about IRID's R&D

projects as well as results of student

Disseminating information to universities and research institutions

re s e a rc h w e re d i s p l a y e d i n t h e

exhibition hall. Engineers and students

involved in R&D had the opportunity

IRID is actively working to develop human resources by visiting entities such as universities and research institutions andproviding them with information.

to engage in directd i s cu s s io n s w ith
attendees.
▲ Panels at the exhibition hall

▲ Audience member listening eagerly
to student presentation

Nuclear Almanac Editorial Board

Student awards ceremonies / Site observation at the JAEA Naraha Center for Remote Control Technology Development (NARREC) and Fukushima Daiichi NPS
IRID presented awards to students selected by the2022 Nuclear Almanac

Senior Manager of the R&D
Management Department

judges for their research presentations. All attendees

contributed an academic paper titled

who presented at the symposium were invited JAEA▲

AwardsNARREC, Fukushima Daiichi NPS and the Fukushima

ceremony

Robot Test Field for observation. The purpose of thesite visit is to utilize technology for future research"IRID's technological development

related to the containing, transport,
and storage of fuel debris during thedecommissioning of Fukushima Daiichi NPS" to this book.

activities involving decommissioning.

JAEA NARREC▲▲On-site at the Fukushima Daiichi NPS

117

two-day event (as announced

S e nior M anage r O kuzumi of IRID 's R&D Strate g y Planning▲the development status of fuel debris

4,500 ardent fans and their

by Fukushima Prefecture, theJapan Society of Applied Physics
Date: September 10, 2021 (online)

R&D since the accident and described

Source: Excerpt from the Robotics Society of Japan website

Date: November 19–20, 2021
Venue: Big Pallete Fukushima (Koriyama, Fukushima)

At the ANSJ Fall Meeting 2021, Senior

opening remarks, Mr. Toyoaki Yamauchi, President of IRID, expressed the desire to

The robots to be used range from small

investigation robots that can pass through the narrow pipes to access PCV

Pro g r e s s in d e c o m mis sio nin g r e s e a rc h a n d d e ve l o p m e nt 2022

The nuclear decommissioning is a long-term project that can span 30–40 years.

Title: 2022 Nuclear Almanac

Editor: Nuclear Almanac Editorial Board
Publisher: Nikkan Kogyo Shimbun Ltd.

Tohoku University
FY2021

Light-Water Reactor Safety Seminar(Nuclear Decommissioning Edition)
Date: September 16, 2021
Venue: Tohoku University (Sendai, Miyagi)

Mr. Takamori, General Manager of the R&D

Planning Department, gave an online lecture inwhich he discussed an overview of IRID, thes t a t u s o f f u e l d e b r i s i nv e s t i ga t i o n, a n d▲ K.Takamori,

General Manager of IRID.

technological development for the retrieval of fuel debris as well as

challenges and safety design.

118