Contents
Session A: Research and Development related to Decommissioning..................................1
A01

Organization Profile of IRID
Isao Imamura1
1: International Research Institute for Nuclear Decommissioning (IRID),

A02

Overview of IRID R&D Projects
Isao Imamura
International Research Institute for Nuclear Decommissioning (IRID),

A03	Research and Development of the Project of Decommissioning, Contaminated Water andTreated Water Management and Connection to Preliminary Engineering
Kazuhito Yoshida1, Mariko Regalado1
1: Mitsubishi Research Institute, Inc. (Management Office for the Project ofDecommissioning, Contaminated Water and Treated Water Management)

Session B: Fuel Debris Retrieval................................................................................................4
B01	Development of criticality impact analysis method for weakly coupled systems including
fuel debris particles
Hiroki Takezawa1, Jun Nishiyama2 and Toru Obara2
1: Nagaoka University of Technology, 2: Tokyo Institute of Technology
B02	Remote Control Technology for Monitoring Inside PCV Pedestal during Retrieval of Fuel
Debris
N. Matsuhira1, R. Komatsu1, S. Nakashima1, A. Yamashita1, R. Fukui1, H. Takahashi1,
T. Shimazoe1, H. Woo2, Y. Tamura3, T. Takahashi4, Y. Yokokohji5, K. Nakamura6,
K. Naruse6, T. Hanari7, K. Kawabata7, S. Suzuki1 and H. Asama1
1: The Univ. of Tokyo, 2: Kogakuin Univ., 3: Tohoku Univ., 4: Fukushima Univ.,
5: Kobe Univ. 6: Univ. of Aizu, and 7: JAEA
B03	Navigation and Control of a Novel Shock-resistant Mechanical Manipulator for Fuel
Debris Retrieval
S. Nakashima1, R. Komatsu1, A. Moro2, A. Faragasso1, H. Woo3, N. Matsuhira1,
K. Kawabata4 and H. Asama1
1: The Univ. of Tokyo, 2: RITECS Inc., 3: Kogakuin Univ., and 4: JAEA
B04

Neutron detection system without radiation protection for criticality approachmonitoring based on diamond sensors and radiationresistive integrated-circuits
M. M. Tanaka1, T. Endo2, J. Kaneko3, Y. Tanimura4, H. Umezawa5, K. Watanabe6,
Y. Fujita1, E. Hamada1, T. Kishishita1, M. Miyahara1, H. Sendai1, M. Sakaguchi1,
M. Shoji1,T. Shimaoka5 and K. Tauchi1
1: High Energy Accelerator Research Organization., 2: Nagoya University,
3: Hokkaido University, 4: Japan Atomic Energy Agency,i5: National Institute of Advanced Industrial Science and Technology,
6: Kyushu University
B05

The world's highest radiation-resistant lubricants supporting decommissioning of thenuclear reactor
Yoshikazu Hayashi, MORESCO Corporation
Advanced Specialist of The Radiation-Resistant Lubricants

B06

Labor-saving design and construction for Fuel Removal at Fukushima Daiichi Nuclear
Power Station Unit 2Motohiro Taniyama1, Jyunro Nakagoshi1, Miho Miyazaki1, Ippei Matsuo1, Kihei Ogawa1
1: Kajima Corporation

Session C: Evaluation and Estimation Technology of Aging Fuel Debris Properties........ 10
C01

Evaluation of fuel debris behavior and α- and β- ray radiation effects on corrosion
Nobuaki Sato1, Kuniki Hata2, Tomonori Sato2 and Chiaki Kato2
1: Tohoku Univ., 2: Japan Atomic Energy Agency

C02

Preparation, microstructure characterization and mechanical property evaluation of
metallic solidified mixtures
Huilong Yang1, Kenta Murakami1, 2, Jiaxing Ren1, Ruheine Naidu Chandren2,
Sho Kano1, and Hiroaki Abe1
1: Tokyo Univ., 2: Nagaoka Univ. of Technology

C03

Effects of beta-ray irradiation and gas-phase radiolysis on corrosion
Tomonori Sato1, Kuniki Hata1, Chiaki Kato1, Atsushi Kimura2, Mitsumasa Taguchi2,
Yutaka Watanabe3
1: Japan Atomic Energy Agency, 2: QST, 3: Tohoku University

C04

Development of Estimation Technology of Aging Properties of Fuel Debris of Fukushima
Daiichi Nuclear Power Stations
Shohei Kawano1, Akihiro Suzuki2, Yusuke Miura2 Yoshiyuki Kawaharada1,
Shinya Miyamoto1,
1: Toshiba Energy Systems & Solutions Corp.,
2: Nippon Nuclear Fuel Development Co., LTD.

Session D: On-site Investigation and Analysis Technlogy................................................... 14
D01

Development of Real-Time Dose Fiber-Monitor with Red-Scintillator
Shunsuke Kurosawa1, 2, Shohei Kodama3, Akihiro Yamaji1, Chihaya Fujiwara1, Daisuke
Matsukura1
1: Tohoku Univ., 2: Osaka Univ., 3: Saitama Univ.

D02

Development of Rapid and Sensitive Radionuclide Analysis Method by Simultaneous
Analysis of β, γ, and X-rays
Hirofumi Shinohara1, Masumi Oshima1, Yuichi Sano1, Katsuyuki Suzuki1,

ii

Haifeng Shen1, Jun Goto2, Tadahiro Kin3, Takehito Hayakawa4, Masahiro Taniguchi5,
Tomoko Haraga6, Masato Asai6
1: Japan Chemical Analysis Center, 2: Institute for Research Promotion, Niigata Univ.,
3: Department of Engineering Science, Faculty of Engineering Sciences, Kyushu Univ.,
4: National Institutes for Quantum Science and Technology, 5: TAISEI CORPORATION,
6: Nuclear Science Research Institute, Japan Atomic Energy Agency
D03

Progress in the development of the continuous monitoring of tritium water by midinfrared laser spectroscopy (Part I)
Ryo Yasuhara1, Hiyori Uehara1, Masahiro Tanaka1 and Naofumi Akata2
1: National Institute for Fusion Science, 2: Hirosaki Univ.

D04

Development of the Sample-return Technique for Fuel Debris Using the Unmanned
Underwater Vehicle
So Kamada1, Kazuya Nishimura1, Tetsuichi Kishishita2, Takahiro Oyama2,
Keisuke Okumura3, and Kennichi Terashima3, and Masayuki Hagiwara4
1: National Institute of Maritime, Port, and Aviation Technology,
2: High Energy Accelerator Research Organization, 3: Japan Atomic Energy Agency,
4: National Institutes for Quantum Science and Technology

D05

Technologies for measurement and management of radioactive materials and wastes
Junichi Hikosaka1, Chris Carter2 and David John2
1: Cavendish Nuclear Japan KK, 2: Cavendish Nuclear Ltd.

D06

Local Identification of Melted Multicompnent Materials and Mapping of Distributed theOriginal-material using Laser-Induced Breakdown Spectroscopic Multi-Spectra via
Multivariate Analysis
Tomohiko Kawakami1, Sakiko Nagayama1,Koudai Okazaki1, Yuta Abe2,
Etsuyo Makuuchi3, Masahiko Ohtaka2 and Yoshitaka Takagai4
1: Kaken., 2: JAEA, 3: Art KAGAKU, 4: Fukushima Univ.

D07

Contamination of SGTS filter trains of Units 1 to 4 caused by the backflow of vent gas
Kenji Owada1, Masato Mizokami1, Takeshi Honda1, Shinya Mizokami1,
Hironori Suzuki1, Hiroshi Ide1, Cibula Michal1, Yutaka Hirokawa2, Yuta Sato2,
Ryota Hashiwaki2
1: TEPCO HD, 2: ATOX Co., Ltd.

Session W: Radioactive Waste Management........................................................................ 21
W01

Study on Rational Treatment/Disposal of Contaminated Concrete Waste Considering
Leaching Alteration (1) Overview
Tamotsu Kozaki1, Takafumi Sugiyama1, Katsufumi Hashimoto1, Naoko Watanabe1,
Shinichiro Uematsu1, Daisuke Kawasaki2, Satoshi Yanagihara2, Daisuke Minato3, Toru
Nagaoka3, Yuka Morinaga3, Naofumi Kozaki4, Yoshikazu Koma4
1: Hokkaido Univ., 2: Fukui Univ., 3: CRI EPI, 4: JAEA

iii

W02

Study on rational treatment/disposal of contaminated concrete waste considering
leaching alteration: (2) Effect of leaching alteration on 137Cs migration in cement paste
Shinichiro Uematsu1, Kyoya Watanabe1, Keisuke Matsumoto1, Naoko Watanabe1,
Tamotsu Kozaki1, Yuka Morinaga2, Daisuke Minato2 and Toru Nagaoka2
1: Hokkaido Univ., 2: CRIEPI

W03

Quantitative Evaluation of Long-Term State Changes of Contaminated Reinforced
Concrete Considering the Actual Environments for Rational Disposal
Ippei Maruyama1, Kazuo Yamada2, Kazutoshi Shibuya3, Yoshifumi Hosokawa4, Yo
Hibino5, Yasumasa Tojo6 and Yoshikazu Koma7
1: Univ. of Tokyo, 2: National Institute for Environmental Studies,
3: Taiheiyo Consultant Co. Ltd., 4: Taiheiyo Cement Corp., 5: Nagoya Univ.,
6: Hokkaido Univ., 7: Japan Atomic Energy Agency

W04

Challenge of Novel Hybrid-waste-solidification of Mobile Nuclei Generated in Fukushima
Nuclear Power Station and Establishment of Rational Disposal Concept and its Safety
Assessment
Masahiko Nakase1, Miki Harigai1, Shinta Watanabe1, Ryosuke Maki2, Hidetoshi
Kikunaga3, Tohru Kobayashi4, Tomofumi Sakuragi5, Ryo Hamada5 and Hidekazu Asano5
1: Tokyo Institute of Technology, 2: Okayama University of Science,
3: Tohoku University., 4: Japan Atomic Energy Agency,
5: Radioactive Waste Management Funding and Research Center

W05

SIAL® : Geopolymer solidification technology approved by Slovak / Czech Nuclear
Authority
Milena Prazska and Marcela Blazsekova , Jacobs (Slovakia)
Hisashi Mikami , Isamu Kudo and Nobuyuki Sekine , Fuji Electric Co.,Ltd.

W06

Waste Management Symposia: The Annual Phoenix Radioactive Waste Conference
Exchanging Knowledge from Around the World
Kazuhiro Suzuki1, Gary Benda2 and Akira Ono3
1: WM Symposia Board of Director, 2: WM Symposia Deputy Managing Director/
Program Advisory Committee (PAC) Chair and 3: WM Symposia PAC Member

W07

Integrated Waste Management and its Application to Nuclear Decommissioning andDismantling Projects
Michelle Dickinson, Antonio Guida and Bill Miller
Jacobs UK Ltd., Decommissioning and Regeneration Solutions

Session E: Development of Radiation-Hardened Circuits................................................... 28
E01

Prototype of Differential Amplifier Circuits Based on Radiation hardened H-diamond
MOSFET (RADDFET)
Hiroki Fukushima1, Manobu M. Tanaka2, Hitoshi Umezawa3, Hiroyuki Kawashima4,
Yusei Deguchi1, Tadashi Masumura1, Naohisa Hoshikawa1 and Junichi H. Kaneko1iv

1: Hokkaido Univ., 2: KEK, 3: AIST
E02

Optically reconfigurable gate array VLSI without any common signal
Sae Goto, Minoru Watanabe and Nobuya Watanabe
Okayama University

E03

Total-ionizing-dose tolerance of an optically reconfigurable gate array VLSI
Kaho Yamada, Takeshi Okazaki, Minoru Watanabe and Nobuya Watanabe
Okayama University

Session G: Decontamination and Environmental Remediation......................................... 31
G01

Development of technology to reduce environmental problems via innovative water
purification agents
Naoki Asao1, Taketoshi Minato2, Natsuhiko Yoshinaga3, Kazuto Akagi3, Joseph Hriljac4,
and Neil Hyatt5
1: Shinshu Univ., 2: Institute for Molecular Science, 3: Tohoku Univ.,
4: Diamond Light Source, 5: Univ. of Sheffield

G02

Fukushima Forests: Decontamination and Restoration
D.Hildebrand1, S.Suzuki2, H.Asama2, C.Casto3, V.Dudarchik1, V.I.Kislyi1, Y.Kawashima1,
P.Molchanov1, R.Foster1, C.Chow1, V.Boksha1
1: NeuroSyntek, 2: University of Tokyo, 3: Casto Group

Session J: Research and Development of JAEA.................................................................... 33
J01

JAEA’s Efforts for Human Resource Development and Regional Commitment inFukushima
Satomi Ito1, Yoshihiro Tsuchida1, Satomi Kakutani1
1: Japan Atomic Energy Agency

J02

Summary of JAEA’s efforts for fuel debris retrieval
Shin-ichi Koyama, Naoya Kaji, Satomi Kakutani
Japan Atomic Energy Agency

J03

R&Ds on Waste Management Technologies for Fukushima Daiichi NPS with National andInternational Collaboration
Yoshikazu KOMA, Naoya KAJI1
1: Japan Atomic Energy Agency (JAEA)

J04

Experimental research related to formation of debris using large-scaled test facility in
CLADS
Yuji NAGAE
Japan Atomic Energy Agency, Collaborative Laboratories for Advanced
Decommissioning Science

J05

Advanced application method of optical fiber radiation sensor towards thedecommissioning of Fukushima Daiichi NPSv

Yuta Terasaka1
1: Collaborative Laboratories for Advanced Decommissioning Science, Japan Atomic
Energy Agency
J06

Overview of Radioactive Material Analysis and Research Facility Laboratory-1
Yuki Sugaya, Hideyuki Akutsu, Hiroshi Shibata, Koji Ichitsubo
Japan Atomic Energy Agency (JAEA)

J07

Development of Exposure Reduction Technologies by Digitalization of Environment andRadioactive Source Distribution –Current Status of Development Project–
Masahiro Suzuki, Takashi Yamaguchi, Masahiko Machida, Kuniaki Kawabata, Rintaro
Ito, Koji Okamoto
Japan Atomic Energy Agency

Session F: France..................................................................................................................... 40
F01

Overview of current CEA developments for characterization techniques
Olivier Gueton1, Christophe Roure1 and Magali Saluden2
1: CEA, DES, IRESNE, Cadarache, 13108 Saint-Paul-lez-Durance, France
2: CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

F02

Cement-based irradiation-resistant materials
David Chartier1, Stéphane Perrin1, Maxime Fournier1 and Magali Saluden2
1: CEA, DES, ISEC, DE2D, Univ. Montpellier, France
2: CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

F03

PACH3 (PACkage for H3-waste) projectDavid Lambertin1, Stéphane Perrin1, Etienne Tevissen2 and Magali Saluden3
1: CEA, DES, ISEC, Marcoule, 30207 Bagnols sur Cèze, France
2: CEA, DES, IRESNE, Cadarache, 13108 Saint-Paul-lez-Durance, France
3: CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

F04

Remediation of contaminated soils in post-nuclear accident situations: theDEMETERRES MOUSSE projectAnne Fornier1, Maxime Fournier2, Malvina Rennesson3, Marc Messalier4,
Jérôme Guillevic5, Christelle Latrille6, Sylvain Faure2, Pierre Chagvardieff7
1: CEA, DES, DDSD, DTPI, Saclay, France
2: CEA, DES, ISEC, Univ. Montpellier, Marcoule, France
3: Orano, DPS2D, Prisme, 125 avenue de Paris, 92320 Châtillon, France
4: Veolia Nuclear Solutions, Bagnols sur Cèze, France
5: IRSN, PSE-ENV/SEREN, 92262 Fontenay-aux-Roses, France
6: Université Paris-Saclay, CEA, SECR, 91191 Gif-sur-Yvette, France
7: Experconnect, avenue Franklin Roosevelt, 75008 Paris, France

F05

French facilities in support of Fukushima Daiichi fuel debris cutting and retrieval R&D
Christophe Journeau1, Arthur Denoix1, Jules Delacroix1, Nicolas Minazzo1, VivianeviBouyer1, Ioana Doyen2, Romain Garnier3, Emmanuel Porcheron4, Antonin Bouland4,
Romain Berlemont5
1: CEA, DES, IRESNE, DTN, SMTA, Severe Accident Experimental Lab., Cadarache,13108 Saint-Paul-lez-Durance, France
2: Université Paris-Saclay, CEA, Service d'Études Mécaniques et Thermiques, 91191
Gif-sur-Yvette, France
3: CEA, DES, ISAC, DE2D/SEAD, Simulation and decommissioning techniques
laboratory, Marcoule, 30207 Bagnols-sur-Cèze, France
4: Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192
Gif-sur-Yvette, France
5: ONET Technologies CN, 950 - 970 chemin des agriculteurs, 26700 Pierrelatte, France
F06

Mitigation of radioactive aerosols dispersion during laser cutting
Damien Roulet1, Rémi Delalez1, Thomas Da Silva1, Emmanuel Porcheron2, Thomas
Gelain2, Yohan Leblois2, Ioana Doyen3, Christophe Journeau4, Christophe Suteau4
1: ONET, Engineering Business Unit, 26700, Pierrelatte, France
2: Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, Gif-SurYvette, 91192, France
3: Université Paris-Saclay, CEA, Service d’Etudes Mécanique et Thermiques, 91191,
Gif-Sur-Yvette, France
4: CEA, DES, IRESNE, DTN, SMTA, LEAG, Cadarache, F-13108 Saint-Paul-lezDurance, France

F07

French Expertise and Technologies applied to Fukushima Daiichi Fuel Debris Sorting
Damien ROULET1, Yvan LALLOT1, Thomas GRUNENBERGER1, Daphné OGAWA2,
Nicolas DEWYSE2, Pierre STROCK2, Benoit TASTET2, Frédérick CARREL3,
Adrien SARI3, Roberto DE STEFANO3, Aly ELAYEB3
1: ONET Technologies, Pierrelatte, France
2: ORANO
3: Université Paris-Saclay, CEA, List, F-91120 Palaiseau, France

F08

Orano expertise to propose a comprehensive solution for Large-Scale Retrieval andInterim Storage of Fuel Debris at 1F
Daphné Ogawa1, Nicolas Dewyse1, Vincent Janin1, Arnaud Rollet1, William Rocher1
1: Orano

F09

Innovative solutions for Fuel Debris Retrieval (FDR) New Lateral Opening for RPV
access and cells conceptual design for Large Scale FDR through PCV side access
Laurent David1, Daphné Ogawa1, Vincent Bessiron² and Pascal De Vito2
1: Orano, 2: Framatome

F10

DEM&MELT In-Can Process for Fukushima Daiichi Nuclear Power Station Water
Treatment Secondary Waste
Régis Didierlaurent1, Daphné Ogawa1, Thierry Prevost1, Laurent David2, MaximeviiFournier3, Caroline Michel3, Jean-François Hollebecque3, Kohei Shibata4
1: Orano Recyclage, Le Prisme, 125 avenue de Paris, 92320 Châtillon, France
2: Orano Projets, Place de Wicklow, Montigny le Bretonneux – 78182 Saint Quentin
Yvelines, France
3: CEA, ISEC, DES, DE2D, Univ. Montpellier, Marcoule, France
4: ANADEC, 4-11-3 Shiba, Minato-ku, Tokyo 108-0014 Japan
F11

Chameleon and Anemone tools - Innovative gripping technologies capable of sampling
and recovering fuel debris
D. OGAWA, K. LE FLANCHEC, V. TOULEMONDE, A. COUDRAY
ORANO

F12

Tooling Solutions for Nuclear Dismantling Projects
Heiko Herbell1, Holger Damies1, Frank Kostroun1, Christoph Stiepani1, Norio Terada2,
Vincent Bessiron2
1: Framatome GmbH, 2: Framatome Japan KK

viii

A01

Organization Profile of IRID
Isao Imamura1
International Research Institute for Nuclear Decommissioning (IRID),

1

Abstract
Ever since the International Research Institute for Nuclear Decommissioning (IRID) was established in August
2013, IRID has been committed to challenging an urgent issue—research and development (R&D) of technologies
required for the decommissioning work of the Fukushima Daiichi Nuclear Power Station (NPS) including
strengthening the foundation of decommissioning technology.
1. Roles of IRID
1-1.Scope of Work
●R&D for nuclear decommissioning●Promotion of cooperation on nuclear decommissioning with relevant international anddomestic organizations●Human resource development for R&D
1-2. Roles of the Organizations for the Decommissioning Project of the Fukushima Daiichi NPS.

1-3.IRID R&D Scope

２. IRID’s Three R&D for Nuclear Decommissioning

1

A02

Overview of IRID R&D Projects
Isao Imamura
International Research Institute for Nuclear Decommissioning (IRID),

Abstract
For the decommissioning of Fukushima Daiichi Nuclear Power Station (NPS), four organizations work together
as one team. IRID is a complex entity consisted of eighteen organizations that play a leading role in research anddevelopment (R&D) for the decommissioning of Fukushima Daiichi NPS. IRID is conducting R&D projects
including the project of “Preparation of Fuel Debris Retrieval” and “Treatment and Disposal of Radioactive Waste”.
Prior determination of fuel debris retrieval methods in fiscal year (FY) 2019, IRID has developed threetechnologies for; 1. Intensive investigation for fuel debris and the damaged conditions inside the reactor, 2.
Potential risk management and verification for the nuclear safety, 3. A reliable remote operation under highradiation environments.
1. Progress of R&D
As for the preparation of fuel debris retrieval, IRID is undertaking R&D projects based on three elements.
Firstly, detection technology that enables to directly access fuel debris in the PCV has been developed. In April
2015, a robot successfully entered the PCV at Unit 1. In FY 2016, a preparation for fuel debris investigation outside
the pedestal started. At the same time, investigation robots for inside the pedestal, and remotely operated drillingdevice for the PCV penetration have been developed for reducing worker exposure at Unit 2. Additional fuel debris
investigation inside the pedestal were performed by an underwater swimming robot for Unit 3, and newinvestigation equipment with a telescopic pipe and cameras for Unit 2. These robots remotely accessed fuel debris
and successfully obtained visual data of the primary containment vessel (PCV) internals. The Severe Accident
Analysis Code was upgraded to identify fuel debris inside the reactor, and investigations through the cosmic-ray
muon were performed. The distribution of fuel debris in the reactor was investigated from outside the reactor
building by using the muon for Unit 1. The muon investigation showed that a large amount of fuel is less likely toremain in the reactor core. The muon transmission measurement was performed for Unit 2 from March to July
2016, and for Unit 3 from May to September 2017. Essential technologies for accessing fuel debris in the RPV or
the PCV are currently developed. Therefore, ensuring the safety of fuel debris retrieval are required.
2. Future Development
The purpose of IRID is to proceed with
R&D

for

the

decommissioning

of

the

Fukushima Daiichi NPS and to acquire
knowledge and expertise from around the world.
Specifically, overseas technology for removal
and storage of damaged fuel as well as the safety
management system are required.

2

A03

Research and Development of the Project of Decommissioning,
Contaminated Water and Treated Water Management andConnection to Preliminary Engineering
Kazuhito Yoshida1, Mariko Regalado1
Mitsubishi Research Institute, Inc. (Management Office for the Project of1

Decommissioning, Contaminated Water and Treated Water Management)
Abstract
The Ministry of Economy, Trade and Industry has established the fund since FY 2013 and implemented the“Project of Decommissioning,Contaminated Water and Treated Water Management" as the subsidy program bysolicitations to support R&Ds with high technical difficultiess. In this presentation, it shall be introduced theconnections among the various subsidized projects with this program and the expected contributions to thedecommissioning of Fukushima Daiichi NPS.
1. Introduction
In order to implement the decommissioning of the Fukushima Daiichi NPS safely and steadily, it is important
to conduct R&Ds by gathering wisdom in Japan and overseas. Therefore, the Ministry of Economy, Trade andIndustry has established the fund since FY 2013 and implemented the "Project of Decommissioning, Contaminated
Water and Treated Water Management" as the subsidy program by solicitations to support R&Ds with hightechnical difficulties. Various R&Ds in the program have been managed by the Management Office for the Projectof Decommissioning, Contaminated Water and Treated Water Management. Mutual coordination among the
research projects are necessary to apply the results of R&Ds to the decommissioning of Fukushima Daiichi NPS.
2. Subsidized Projects of Decommissioning, Contaminated Water and Treated Water Management andConnection to Preliminary Engineering
The subsidized projects are classified into “Internal Investigation”, "Development of Retrieval Method",“Improvement of Work Environment”, and “Processing of Solid Waste, etc.”. The R&Ds of Fuel Debris Retrieval
have been conducted based on the information obtained by Internal Investigation. In addition, the results of R&Ds
such as Development of Fuel Debris Retrieval Method are reflected to Improvement of Work Environment. The
research projects of Processing of Solid Waste are also studied in cooperation with R&Ds for Fuel debris retrieval
and Improvement of Work Environment. In this way, the current projects are closely connected and then theengineering has been utilizing for wisdom in Japan and overseas.

3

B01

Development of criticality impact analysis method for weakly
coupled systems including fuel debris particles
Hiroki Takezawa1, Jun Nishiyama2 and Toru Obara2
1

Nagaoka University of Technology, 2Tokyo Institute of Technology

Abstract
For fuel debris removal work at the Fukushima Daiichi NPS (1F), it is necessary not only to avoid a criticality
by engineered safety measures but also to establish counter measures for the criticality based on impact evaluations.
A space-dependent kinetic analysis code specific to weakly coupled reactors including fuel debris particles has
been under development based on the integral kinetic model.
1. Introduction
It is important to evaluate the impact of a criticality that
may occur in removing fuel debris from the damaged 1F
reactors for ensuring safety of workers engaged in theremoval operation [1]. The fuel debris located at the bottomof the 1F RCVs can be a fast-thermal weakly coupled reactor
system including fuel debris particles that are possible tomove (Fig. 1). So, a space-dependent kinetic analysis code

Figure 1. Example of weakly coupled system.

specific to weakly coupled reactors, Multi-region Integral Kinetic (MIK) code, has been under development basedon the integral kinetic model (IKM) [2].
2. Methods
The IKM calculates the fission reaction rate in region i at
the present time t ( 𝑁𝑁� (𝑡𝑡) [fissions/s]) by integrating allcontributions from past fissions in source region j at past timet’ to the present fissions using:���𝑁𝑁� (𝑡𝑡) = ∑� �∫�� �𝛼𝛼�� (𝜏𝜏) + 𝛼𝛼��� (𝜏𝜏)� 𝑁𝑁� (𝑡𝑡 � )𝑑𝑑𝑑𝑑 � �,
�

(1)

�

where 𝛼𝛼�� (𝜏𝜏) and 𝛼𝛼��� (𝜏𝜏) are the probability density

functions of secondary fissions in region i at the present time twhich are induced by prompt neutrons or delayed neutrons of
the I precursor family that are generated from a source fission
th

𝐩𝐩

𝐝𝐝

Figure 2. Example of 𝑪𝑪𝒊𝒊𝒊𝒊 (𝝉𝝉) and 𝑪𝑪𝒊𝒊𝒊𝒊𝑰𝑰 (𝝉𝝉).

in region j at the past time t’ with a time difference 𝜏𝜏 𝜏 𝜏𝜏 𝜏 𝜏𝜏𝜏 [secondary fissions@i/sec/source fission@j].
3. Progress and Future Plan

Continuous energy Monte Carlo neutron transport codes, such as MVP or Serpent, will be used to calculate
�

�

�

�

cumulative distribution functions 𝐶𝐶�� (𝜏𝜏) and 𝐶𝐶��� (𝜏𝜏), the integral form of the functions 𝛼𝛼�� (𝜏𝜏) and 𝛼𝛼��� (𝜏𝜏).
��
Example of 𝐶𝐶�� (𝜏𝜏) and 𝐶𝐶��� (𝜏𝜏) functions calculated for Godiva reactor are shown in Fig. 2 and they were verified

by the comparison to Godiva delayed neutron fractions. Delay of delayed neutrons emission will be introduced in
Eq. (1) by the law of decay. After verification of the new MIK code, it will be coupled to an MPS code to reflect
fuel debris particles movement on criticality impact analysis. This work is financially supported by the Nuclear
Energy Science & Technology and Human Resource Development Project (through concentrating wisdom) fromthe Japan Atomic Energy Agency / Collaborative Laboratories for Advanced Decommissioning Science.

References[1] https://clads.jaea.go.jp/jp/rd/map/2022/issues/fdr/fdr-207.html, [2] Takezawa, Tuya, Obara, NSE195(2021)1236.

4

B02

Remote Control Technology for Monitoring Inside PCV Pedestal
during Retrieval of Fuel Debris
N. Matsuhira1, R. Komatsu1, S. Nakashima1, A. Yamashita1, R. Fukui1,

H. Takahashi1, T. Shimazoe1, H. Woo2, Y. Tamura3, T. Takahashi4, Y. Yokokohji5, K. Nakamura6,
K. Naruse6, T. Hanari7, K. Kawabata7, S. Suzuki1 and H. Asama1
1

The Univ. of Tokyo, 2Kogakuin Univ., 3Tohoku Univ., 4Fukushima Univ., 5Kobe Univ. 6Univ. of Aizu,
and 7JAEA

Abstract
This research aims to develop human resources in the field of remote technology for the decommissioningof the Fukushima Daiichi Nuclear Power Plant (1F). We conduct research on a monitoring platform for fuel debris
removal. We expect to develop research personnel through participation in projects, lectures, and facility tours.
1. Introduction
To safely and reliably remove fuel debris using a remote-controlled robotic arm, it is necessary to accurately
grasp the three-dimensional situation inside the containment vessel before each operation. We propose to constructa platform for monitoring inside the pedestal. Various sensors such as cameras, gamma-ray and neutron detectors
will move on this platform to perform measurements and visualization necessary for fuel debris removal.
2. Development Subjects
2-1. Monitoring platform (The Univ. of Tokyo, Fukushima Univ.)
Technology for building modular, split-type platforms andhighly rigid, lightweight arms that can be retracted compactly havebeen developed.
2-2. Remote control interface (The Univ. of Tokyo, Kobe Univ.)
We will develop a video presentation interface for operators andFigure 1. A monitoring platform concept

a highly realistic tele-operation system.
2-3. Radiation monitoring device (The Univ. of Tokyo)

Neutron/gamma-ray measurement devices that can operate under high dose rates and imaging techniques that
integrate positional information has been developed.
2-4. A three-dimensional reconstruction method of environmental models (JAEA CLADS, Univ. of Aizu)
A three-dimensional restoration technique of environmental models using data obtained from cameras forunderstanding the in-vessel environment easily has been developed.
3. Conclusion
We have conducted research on a monitoring platform for fuel debris removal. In addition, lectures with thecurrent status and future issues of 1F were introduced for students, and the research on master/doctor theses andconference presentations were done by younger person as a human personnel development.
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy
Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource
Development Project. This study has been conducted in the project from FY2019.

5

B03

Navigation and Control of a Novel Shock-resistant Mechanical
Manipulator for Fuel Debris Retrieval
S. Nakashima1, R. Komatsu1, A. Moro2, A. Faragasso1, H. Woo3,
N. Matsuhira1, K. Kawabata4 and H. Asama1
The Univ. of Tokyo, 2RITECS Inc., 3Kogakuin Univ., and 4JAEA

1

Abstract
This research aims at developing a novel manipulator for retrieving fuel debris on the Fukushima Daiichi
Nuclear Power Plant (1F) as shown in Figure 1. We conduct research on designing shock-resistant CVT(Continuously Variable Transmission) robot in collaboration with the University of Sussex. Our main focus is on
navigation and control.
1. Introduction
Fuel debris retrieval at the bottom of the primary containment vessel (PCV) is one of the significant tasks forthe decommissioning of the nuclear power plant and in particular for F1. The presence of radiation, water leakage
and poor light conditions make the retrieval process challenging for conventional manipulators. We tackle those
problems by the novel mechanical manipulator and its control and navigation algorithm. CVT-based actuation will
enhance the robot’s shock resistance. AI-based navigation algorithm aids the operator to navigate the robot andgrasp the fuel debris in the cluttered environment inside the PCV.
2. Development Subjects
2-1. Optimal actuation parameters (The Univ. of Tokyo, RITECS
Inc. )
We configured a simulation environment to evaluate the optimal
parameters of the long-reach manipulator robot.
2-2. Navigation and Control of CVT-VIA manipulator (The Univ.
Figure 1. Robotic manipulator in PCVof Tokyo, RITECS Inc. )We set up the machine learning environment using the dynamics simulator called “PyBullet”. A humanoperator can command the gripper’s target position by clicking on the camera output of the environment.
2-3. System evaluation and demonstration (The Univ. of Tokyo, RITECS Inc., Kogakuin Univ., JAEA CLADS )We will conduct the evaluation and demonstration of the proposed system on a real robot.
2-4. Demonstration of use-case scenarios (The Univ. of Tokyo, RITECS Inc., Kogakuin Univ., JAEA CLADS )We will validate the proposed robot system in the mock-up in the University of Tokyo or NARREC.
3. Conclusion
We have conducted research on a decommission robot manipulator featuring CVT-based actuation and alearning-based navigation system. Currently, component development is in progress. Future works include thedevelopment of the whole manipulator and the integration of the navigation system to real robots.
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy
Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource
Development Project. This study has been conducted in this project from FY2021.

6

B04

Neutron detection system without radiation protection for criticality
approach monitoring based on diamond sensors and radiationresistive integrated-circuits
M. M. Tanaka1, T. Endo2, J. Kaneko3, Y. Tanimura4, H. Umezawa5,

K. Watanabe6, Y. Fujita1, E. Hamada1, T. Kishishita1, M. Miyahara1, H. Sendai1, M. Sakaguchi1,
M. Shoji1,T. Shimaoka5 and K. Tauchi1

1High Energy Accelerator Research Organization., 2Nagoya University, 3Hokkaido University,

4Japan Atomic Energy Agency, 5National Institute of Advanced Industrial Science and Technology,
6Kyushu University

Abstract
This report summarizes the research results of the “Technology
development of diamond-base neutron sensors and radiation-resistive
integrated-circuits for shielding-free criticality approach monitoring
system” conducted until FY2021.
1. Introduction
The study aims to develop key components(diamond sensors andradiation-resistive integrated circuits) of a neutron detector without aradiation shield for a criticality approach monitoring system shown in
Figure 1. It is required high neutron detection efficiency(a few cps/nv)
under 1 kGy/h and compact-light-weight to fit constraints of thepenetration size and the payload.
Figure 1. neutron detector for
Requirements for the neutron detector are determined by numericala criticality approachevaluations of subcriticality under various signal-to-noise ratios(neutron
monitoring systemcounts/γ-ray counts) by simulation studies using measurement dataof reactor noise and simulation studies of neutron sensitivities forseveral neutron converters shown in Figure 2. One of key issues ishigh radiation environment. The response function of the detector
have to be optimized to minimize the γ-ray counts and to maximize
the neutron detection efficiency. The fabrication process of the
diamond neutron sensor is newly developed for the purpose, and thedeveloped signal-processing integrated circuits and a high speed
data-transfer integrated circuit have to work up to a few MGy.
2. Status
This feasibility study clarified that the Feynman-α method forthe subcriticality measurement can measure the prompt neutron
decay constant if the ratio of γ-ray counts to neutron counts is lessthan 1.[1] We confirmed thousands of neutron sensing devices
achived a few cps/nv based on numerical simulations, andfinished developments of a prototype neutron monitoring board
which can be mounted 512 devices with 1 MGy radiation
tolerant signal processing circuits. Diamond sensors with thesignal processing circuits also work under 0.645kGy/h as show
in Figure 3, after modification of production processes. We will
start evaluation of the neutron detector using LiF convertor
under high radiation environment in FY2022.

Figure 2. Sensitivity of a neutron sensor
using LiF. An example of the studies.

3. Acknowledgement

This work was supported by JAEA Nuclear Energy
S&T and Human Resource Development Project throughconcentrating wisdom Grant Number JPJA20P20336542.
References[1] Reports of FY2021 Nuclear Energy Science & Technology and
Human Resource Development Project, to be published.

7

Figure 3. γ-ray noise spectrum under Co γray irradiation environments.

B05
Ab s t r a c t

The world's highest radiation-resistant lubricants
supporting decommissioning of the nuclear reactor

Yoshikazu Hayashi, MORESCO Corporation
Advanced Specialist of The Radiation-Resistant Lubricants

Excellent radiation-resistant lubricants are essential for equipment in high-dose areas.
The lubricants, "MORESCO-HIRAD"s showed the world's highest radiation resistance,
not only against a γ ray but also a neutron-dominated mixed ray.
Accordingly, it has been comprehensively installed in the debris retrieval devices.

1 . In t r o d u c t i o n

It is shown below as the changes in each lublicant's properties depending on the doses that theobtained data from the concerted studies with JAEA or the European accelerator projects.
A lubricant having less change in its properties even at a high dose has an excellent resistance.

2 . Th e i r r a d i a t i o n - e va l u a t i o n s a n d t h e f i n d i n gs f o r t h e l u b r i c a n t s
2 - 1 . W i t h J AEA ～ Th e p r o p e r t i e s c h a n ge s ( = d e t e r i o r a t i o n ) o f t h e o i l s w i t h a γ- r a y

2 - 2 . W i t h t h e Eu r o p e a n a c c e l e r a t o r p r o j e c t s , CERN & ES S ～ Th e gr e a s e s ' c o n s i s t e n c y ( = h a r d n e s s ) c h a n ge s ( = d e t e r i o r a t i o n ) w i t h a n e u t r o n - d o m i n a t e d m i xe d r a y

Stable( * The appearances of RG-42R-1 for each dose )(*1) and (*2) is calculated with each sample's respectively value at the time of non-irradiation as "1".

3 . Th e a d o p t i o n ( = i n s t a l l a t i o n ) e xa m p l e s o f " M O RES CO - H IRAD " a s a gr e a s e f o r m

Co n t a c t d e t a i l s f o r i n q u i r i e s - Te l : + 8 1 7 8 3 0 3 9 0 1 0 , e - m a i l : h a ya s h i @ m o r e s c o . c o . j p

8

Labor-saving design and construction for Fuel Removal at
B06
Labor-saving designand construction
for FuelPower
Removalat Fukushima
Fukushima
Daiichi Nuclear
Station
Unit 2 Daiichi Nuclear
Power Station Unit 2Motohiro Taniyama , Jyunro Nakagoshi1, Miho Miyazaki1, Ippei Matsuo1, Kihei Ogawa1
1

Kajima Corporation

1

Abstract
Since the reactor building of Unit 2 did not experience a hydrogen explosion and its structure frame remains, theexisting frame will be utilized to the maximum extent. However, the core has been melted down and the radiationdose is very high, so it will be required to consider minimizing the exposure of the workers. This paper presents
the labor-saving design and construction will be used for the fuel removal at Unit 2.
1. Labor-saving design
The fuel removal structure made of steel is a rectangular structure of 27.0 m in the east-west direction, 32.7 m in
the north-south direction, and 44.75 m in height. In order to control the relative displacement to the reactor building
and to suppress the deformation of the upper part of the structure, four oil dampers will be installed between thebuildings. The runway girder will be fixed in the structure with straddling the reactor building through an opening
in the south wall of the reactor building. Elastic bearings and oil dampers with springs will be installed between
the runway girder and the floor of the reactor building. This equipment will be installed by not fixing in the existingbuilding with high radiation dose in order to reduce radiation exposure of the workers.
2. Labor-saving construction
In order to reduce the amount of work at the site where the ambient dose is high, to assemble a part of the steel
frame is planned at the off-site area where the ambient dose is low. About 70% of the entire steel frames areassembled at this assembly yard. These assembled units are transported to the site by a self-propelled multi-axle
cart. The transported units are lifted, and the steel frame of the platform structure is constructed. The large units
can be incorporated with ancillary facilities, which contributes to reducing the amount of work other than steel
frame construction. The ground improvement machine has been also modified to allow remote operation, and theoperator can check the status of the improvement work from the remote-control room.

Runway girder

Fuel handling machine

Unit 2 reactor building

Fuel removal platform

West side structure

Front room

Structure

Foundation

Overview of Unit 2 fuel removal structures

9

C01

Evaluation of fuel debris behavior and α- and β- ray radiation
effects on corrosion
Nobuaki Sato1, Kuniki Hata2, Tomonori Sato2 and Chiaki Kato2
Tohoku Univ., 2Japan Atomic Energy Agency

1

Abstract
To know the behavior of fuel debris generated by severe reactor accident, chemical structure and stability of
nuclear fuel debris were evaluated. Radiation effects such as radiation dose and range by of α- and β- rays fromradionuclides in fuel debris were also analyzed.
1. Introduction
The special corrosive environment of inner part of 1F PCV should be considered if structural materials
directly contact with α/β nuclides of nuclear fuel debris at any positions in the PCV. Though the radiation distanceof these α− and β− rays are shorter than that of γ ray, the locally high energy deposition would accelerate thecorrosion of materials. The radiation distance and dose are dependent on the configuration of fuel debris. Types offuel debris are classified as oxide debris, alloy debris and so on. Oxide debris is mainly consisted of UO2 and ZrO2,
with fission products (FP) and minor actinides (MA). Alloy debris is the mixture of UO2 and melted alloy such as
zircalloy and iron steel. The MCCI debris is the product of molten core concrete interaction when molten core
dropped down into concrete part. In this paper, characterization of fuel debris and radiation effects of α- and βrays were evaluated for α/β nuclides of fuel materials and minor actinides contained in the fuel debris.
2. Results and discussion
2-1. Oxide debris
According to the pseudo binary UO2-ZrO2 system, the cubic ZryU1-yO2+x phase is stable in uranium rich region,
while UyZr1-yO2 phases (monoclinic, tetragonal, cubic) appears with increasing temperature. In each oxide phase,Pu, MA and FP elements coexist in the (U,Zr)O2 forming solid solution. As for the radiation effect of these elements,
two β emitters (90Sr, 239Np, 241Pu) and four α emitters (238Pu, 239Pu, 240Pu, 241Am) were considered. Radiation doseand distance were calculated using PHITS 3.14 program with cylindrical UO2(1g) source from its surface (z=0cm)
to water. Distances from debris surface were found to be 40µm and 2mm for α- and β- rays, respectively, while
their surface dose rates (0.1 Gy/s) were higher than that of γ−ray.
2-2. Alloy and MCCI debris
The alloy debris is thought to be a mixture of (U,Zr)O2 and Fe-Zr alloy. Most of high radioactive elements
are contained in the oxide phase. In the case of homogenous mixture of oxide and alloy, the dose rate decreases
with increasing alloy ratio, while the distances are the same as those of oxide debris. If the oxide phase is coveredby thin alloy layer, radiation shielding effect seems to appear, while a part of α/β radiation comes out through thealloy film thinner than 0.1µm. Similar behavior of radiation effect was observed for MCCI debris.
3. Conclusion
α/β radiation effects from high radioactive nuclides in fuel debris were evaluated for different types of debris.
The alloy or concrete thin layer of heterogenous debris seem to suppress radiation effect from oxide debris.
Acknowledgments
This study was supported by the JAEA Nuclear Energy S&T and Human Resource Development through concentrated wisdom[grant number JPJA20P20333127].

10



C02

Preparation, microstructure characterization and mechanicalproperty evaluation of metallic solidified mixtures
Huilong Yang1, Kenta Murakami1, 2, Jiaxing Ren1, Ruheine Naidu Chandren2,
Sho Kano1, and Hiroaki Abe1
Tokyo Univ., 2Nagaoka Univ. of Technology

1

Abstract
In this study, the SUS-Zr-B4C mixtures were prepared to simulate the metallic-dominated corium formation
in 1F. The microstructure was characterized to reveal the phase constitution in these mixtures, in addition, themechanical property of each constituent phase was evaluated. The results will be helpful for reaching a better
understanding of the fundamental property of the 1F fuel debris.
1. Introduction
Achieving a better understanding of the fundamental property of fuel debris is required for the retrieval, longterm storage and processing, and aging behaviors of 1F fuel debris. For example, the mechanical properties suchas hardness, Young modulus, and fracture toughness are important indicators for selecting the cutting method
and/or tools for fuel debris retrieval. The fuel assembly in 1F is mainly composed of fuel, Zircaloy fuel cladding,
B4C control rod and its stainless steel (SUS) cladding, for the sake of simplicity, the research objective of thisstudy is starting from the corium formation of non-fuel components, i.e., the solidified mixture from SUS-Zr-B4C
melt. The acquisition of the mechanical property of the constituent phases in the metallic fuel debris is the purposeof this study.
2. Experiments
The SUS-Zr-B4C solidified mixtures were prepared by high frequency vacuum melting method. Two typesof ingots were prepared with the ratio of SUS/Zr of 1:3 and 1:1 respectively, and the ratio of B4C was 5 wt.%. The
microstructure of the fabricated ingots was characterized using scanning electron microscopy-energy dispersice
spectroscopy, Vickers hardness and nano-indentation tests were conducted to achieve the hardness of the bulk
specimen and each constituent phases.
3. Results and discussion
Microstructural characterization results indicate SUS-75Zr-B4C specimen consists of Zr3Fe, Zr2Fe, and ZrFe2
type intermetallic compounds, and Zr, ZrC, ZrB2 phases, whereas SUS-50Zr-B4C specimen is dominated with
ZrFe2 phase. The phase constitution approximately follows the Zr-Fe binary phase diagram, although the volume
fraction of the constituent phases was deviated from the equilibrium phase diagram. The mechanical properties of
the constituent phases in these ingots, such as nano-hardness and Young modules, were successfully achieved.
Results show that the hardness and Young modules of ZrFe2 compound is the greatest among the intermetallic
compounds, which is responsible for a much greater hardness of SUS-50Zr-B4C ingot relative to SUS-75Zr-B4C
ingot. In addition, it is thus assumed that the Fe-rich metallic corium is more likely to exhibit a higher cracking
susceptibility than the Zr-rich metallic debris.
Acknowledgement
This report includes part of the results obtained from the project “Investigation of Environment Induced Property
Change and Cracking Behavior in Fuel Debris”, supported by Japan Atomic Energy Agency (JAEA) Nuclear
Energy S&T and Human Resource Development Project through concentrating wisdom.

11

C03

Effects of beta-ray irradiation and gas-phase radiolysis on corrosion
Tomonori Sato1, Kuniki Hata1, Chiaki Kato1, Atsushi Kimura2, Mitsumasa
Taguchi2, Yutaka Watanabe3
1Japan Atomic Energy Agency, 2QST, 3Tohoku University

Abstract
To understand the local corrosion environment around fuel debris, effects of β-ray irradiation on the corrosion
behaviors of carbon steel and stainless steel were investigated by electrochemical tests. Effects of gas-phaseradiolysis on the corrosive environment in the liquid-phase were also investigated by γ-radiolysis experiments.
1. Introduction
The Primary containment vessels (PCVs) in Fukushima-Daiichi nuclear power station (1F) is exposed to the
corrosive environment under irradiation. Effects of γ-rays on corrosion of materials in contaminated water had
been investigated and “general effects” of radiation on corrosion were summarized in our previous work[1].
However, the materials nearby the fuel debris are exposed to not only γ-ray but also α- and β-rays from fuel debris.
Effects of gas-phase radiolysis could also be important if large amount of gaseous species are continuously
irradiated at the headspace of PCVs. This gas phase radiolysis affects corrosive environment in some isolated water,which might locally exist in the PCVs. To understand the corrosive environment in 1F PCV in detail, these “local
effects” of radiation should be understood as well as the “general effects”. In this study, some electrochemical tests
and radiolysis experiments were carried out to estimate the “local effects” of radiation on corrosion.
2. Effects of β-ray irradiation on corrosion of steels
The electrochemical tests using a sealed β-ray source (3.0 MBq Sr-90) were performed to evaluate the effectsof β-ray irradiation on corrosion behaviors of steels. The β-ray source was set at the bottom of a test vessel, and200-times diluted artificial seawater was poured in the vessel. The test solution and the β-ray source were separated
by polyimide film. A working electrode made of stainless steel covered by peek was set in the vessel with a certain
distance from the β-ray source, and the corrosion potential of stainless steel were measured. The corrosion potential
showed a higher value than that measured without irradiation. The result indicated that the irradiation of β-ray
would enhance the localized corrosion of stainless steel. The polarization resistance of carbon steel was alsomeasured by using the same system.
3. Effects of γ-ray irradiation to gas phase on corrosive environment
Radiolysis experiments were carried out at the γ-rays irradiation facility in QST. Liquid samples with a certain
amount of headspace were irradiated, and the concentrations of nitrogen oxides (nitrite and nitrate ions) were
measured. The current results show that the concentrations of nitrogen oxides tend to increase when the gas/liquidratio of samples increased. Oxygen in the headspace also enhanced the nitrogen oxide production.
4. Conclusion
In this study, effects of β-ray irradiation and gas-phase radiolysis on the corrosive environment were
investigated. These data would be valuable information for precise estimation of corrosion behavior of thestructural materials in 1F PCVs.
Acknowledgments
This study was supported by the JAEA Nuclear Energy S&T and Human Resource Development throughconcentrated wisdom [grant number JPJA20P20333127].
Reference
Tomonori Sato, et al., “Database for Corrosion under Irradiation Conditions” JAEA-review-2021-001 (2021).

12

C04

Development of Estimation Technology of Aging Properties of Fuel
Debris of Fukushima Daiichi Nuclear Power Stations

Shohei Kawano1, Akihiro Suzuki2, Yusuke Miura2 Yoshiyuki Kawaharada1, Shinya Miyamoto1,
1

Toshiba Energy Systems & Solutions Corp., 2Nippon Nuclear Fuel Development Co., LTD.

Abstract
The understanding of long-term stability of fuel debris (FD) is one of the important issues for the decommissioning
work. Nine types of simulated FD samples underwent water immersion tests or air exposure test to evaluate theprobability of micro-particle generation due to chemical aging under the environments to which FD is exposed in
the Primary Containment Vessel (PCV) interior, or during the retrieval and storage operation. The experimental
results revealed the particle generation occurred on specific FD composition and the environmental condition.
Introduction
The objective of this study is to declare the probability of long-term aging of FD under the circumstance, and to
estimate the aging property to evaluate the influence on the defueling operation. It is reported that part of Chernobyl
fuel containing materials (FCM) becomes fragile and micro-particles generation are progressed, over 30 years has
passed since the accident. The radio-active particle generation is an important aging phenomenon from theviewpoint of the risk during Fukushima Daiichi (1F) decommissioning. Therefore, based on the Chernobyl
knowledge, we defined the probable factors which induce the aging of FD under 1F environment, and conducted
experiments to evaluate the property of particle generation using several types of simulated FD including glassy
and non glassy composition.
Experimental procedure and results
The surrounding of FD is considered to be changed with the progress of decommissioning work, that is, Nitrogengas injection circumstance may be changed to oxygen containing atmosphere. Hence, we presumed the aging
mechanism of FD under 1F environment as follows. (1) Particle generation due to the oxidation of inclusion or
precipitate contained in FD. (2) Particle generation due to dissolution of FD matrix. Nine types of FD simulated
samples which have aging factors mentioned above were fabricated for examinations. Air exposure tests wereconducted at 323 K for 500 h, and at 383 K for 100 h in atmospheric condition. Water immersion tests wereconducted at 303 K for 500 h, and at 363 K for 100 h in air saturation water.
Table 1 shows the types of FD samples and experimental results. U-containing micro-particles observed in case ofZr(O) containing (U, Zr)O2 after air
exposure, grassy FD with UO2 or (U,Zr)O2 inclusion or after water immersion.
Conclusion
It is found that particle generation of FD
occurred on specific FD composition andthe environmental condition.
This study is based on the results ofProgramDecommissioning

andWater Management”.

“Project

Type of FD sample

Test(U, Zr) O2 + (Zr, U) O2

Air
Water

Zr (O) + (U, Zr) O2
Glass + UO2 inclusions
Glass + Zr-U-O inclusions
Glass + Fe inclusions

ACKNOWLEGEMENT

Subsidy

Table 1 Types of FD samples and experimental resultsofContaminated

Air
Air
Water
Air
Air
Water
Air
Water
Water
Water

Results
No particle detected
No particle detected
U-containing particles foundNo particle detected
U-containing particles foundNo particle detected
Fe- containing particles foundFe- containing particles foundFe- containing particles foundGlass + FeO inclusions
Glass without inclusions
No particle detected
Glass containing BoronNo particle detected
Glass + (U, Zr) O2
U-containing particles found- Glass: SiO2 containing FD which simulates molten core concrete
reaction. - Air: Air exposure test. - Water: Water immersion test.

13

D01

Development of Real-Time Dose Fiber-Monitor with RedScintillator

Shunsuke Kurosawa1,2, Shohei Kodama3, Akihiro Yamaji1, Chihaya Fujiwara1, Daisuke Matsukura1
1

Tohoku Univ., 2Osaka Univ., 3Saitama Univ.

Abstract
We have novel red-emission scintillator with a high light output of over 60,000 photons/MeV compared to otherconventional scintillators, and we demonstrated real-time monitoring system for high-dose-rate with wide dynamic
dose range .
1. Introduction
Mapping of high-dose-rate distribution in Fukushima Daiichi Nuclear Power Station is the one of the firststeps for decommissioning, while conventional techniques are hard to monitor the rate due to the such high-dose
rate condition of over 1 Sv/h. We have developed a fiber-type monitor with red-emission scintillator, because large
mount of noises is generated in a photo detector excited by radiations. Scintillation photons are read with a photodetector in lower rate thorough an optical fiber. On the other hand, fiber made by glass itself are also generated
several noises excited by radiations: Cherenkov and scintillation photons in the glass. These noises have generally
emission wavelengths of blow 550 nm, and to discriminate the noise, we have developed red-emission scintillator,
Cs2HfI6 with an emission band of over 600 nm.
2. Methods
As the starting materials 99.9%-pure HfI4 and 99.999%-pure CsCl beads were mixed inside the glove box
filled with the inert Ar gas. The ampoule was evacuated and sealed-off, and then the crystal growth was performed
using the vertical Bridgman-Stockbarger furnace. The X-ray excited RL spectrum was measured with a CCD
camera, DU420-OE (ANDOR). The Cu-Kα X-ray was used for irradiation from a Mini-X (AMPTEK) with anaccelerating voltage of 30 kV and beam current of 100 mA. The intensities were corrected by a CCD sensitivity,
grating efficiency and reflection efficiency of the mirror in the spectrometer.
3. Results
Cs2HfI6 had an emission wavelength of around 700 nm excited by X-ray, and this scintillator was estimated tohave light output of ~63,000 photons/MeV. Also, The FWHM energy resolution at 662 keV was calculated to be~4.4%. Using this scintillator, we have developed the fiber-reading radiation dose monitor system with a 20-mlong optical fiber and a CCD spectrometer. This results shows we performed the real-time radiation detection test
using a ~100 TBq 60Co gamma-ray source, maximum dose ratio of 1.1 kSv/h. The scintillation signal intensities
were measured as a function of the dose ratio, then the monitorable dynamic range was determined as the order of
10-3 up to 100 kSv/h [1].
4. Conclusion
This novel scintillator enables us to monitor high-dose-rate condition with optical fiber, and we succeeded indemonstration with this monitoring system with wide dynamic dose range .
References[1] S. Kodama and S. Kurosawa et al., APEX 13, (2020) 047002

14

D02

Development of Rapid and Sensitive Radionuclide Analysis Method
by Simultaneous Analysis of β, γ, and X-rays
Hirofumi Shinohara1, Masumi Oshima1, Yuichi Sano1, Katsuyuki Suzuki1,
Haifeng Shen1, Jun Goto2, Tadahiro Kin3, Takehito Hayakawa4,Masahiro Taniguchi5, Tomoko Haraga6, Masato Asai6

Japan Chemical Analysis Center, 2Institute for Research Promotion, Niigata Univ., 3Department of1

Engineering Science, Faculty of Engineering Sciences, Kyushu Univ., 4National Institutes for Quantum
Science and Technology, 5TAISEI CORPORATION, 6Nuclear Science Research Institute, Japan Atomic
Energy Agency
Abstract
We aim to realize the rapid analysis of radioactive nuclei in fuel debris and waste. Thus, we are developing a newspectrum analysis method (SDM) to apply to the latest measurement system such as the multiple γ-ray detection
system, and to integrate the research of β(+X)-ray, γ-ray, and multiple γ-ray spectrum.
1.

Introduction

In the future, trial removal of fuel debris in Fukushima Daiichi nuclear reactors will begin. It is essential tounderstand the characteristics of the fuel debris to remove the fuels safely and smoothly from the reactors, andthus various tests will be conducted. When analyzing radioactive nuclei in fuel debris, the conventional radioactive
analysis method requires individual measurements of radioisotopes after complicated chemical separation for eachelement, which takes a great deal of time and effort. In addition, since radiation protection during work and trainingof engineers are issues, it is necessary to develop a simple and rapid method for analyzing radioactive nuclei.
2.

Experiment

While the conventional analysis method (Total Peak Area Method) uses only the full energy peak for quantification,
we have developed the SDM for γ-ray determination that uses the entire spectrum, including peaks and continuum
components [1]. Compared to conventional methods, the SDM has high sensitivity because of using more number
counts for quantification. Furthermore, the SDM has high accuracy because of directly derives radioactivity values
without using peak analysis or nuclear data analysis.
3.

Simulation and analysis

To develop the SDM, we have created standard spectra of 40 nuclides. When measuring a spectrum from standard
radiation sources is difficult, we create it using the Geant4 simulation code. In addition, a machine learning model
was built to identify simultaneously multiple nuclides such as Co-60, Cs-134, Cs-137, Eu-152, etc., and a
convolutional neural network (CNN) was used to improve the accuracy of the SDM. We also examined the
optimization of chemical separation suitable to the measurement conditions of the SDM.
4.

Conclusion

The SDM is a reliable γ-ray determination with a simple procedure demonstrated in our study. Thus, we can expect
to shorten the chemical separation process by simultaneous quantification of several nuclides with the SDM to
reduce worker exposure and to obtain accurate results without advanced chemical separation skills.
Acknowledgment
This work was supported by JAEA Nuclear Energy S&T and Human Resource Development Project throughconcentrating wisdom Grant Number JPJA20P20333366.
References[1] M. Oshima et al., J. Nucl. Sci. Tech. Vol.59. No.4 (2022) pp.472-483.

15



D03

Progress in the development of the continuous monitoring of tritium
water by mid-infrared laser spectroscopy (Part I)
Ryo Yasuhara1, Hiyori Uehara1, Masahiro Tanaka1 and Naofumi Akata2
National Institute for Fusion Science, 2Hirosaki Univ.

1

Abstract
The present study aims to demonstrate the principle of short-time measurement of tritiated water using acavity ring-down spectroscopic measurement system with a mid-infrared laser. A specially constructed opticalcavity for mid-infrared CRDS was constructed. We have also demonstrated a mid-infrared laser that has ameasurement advantage with broadband wavelength tunability. In addition, stable isotope heavy water samples
were prepared as standard samples.
Our group has been developing a detection technique for hydrogen isotopes using mid-infrared lasers. The midinfrared region has very large absorption lines of vibrations of hydroxyl groups, which can be used for highly
sensitive detection. Recently, we have demonstrated the H2O and D2O in the liquid phase were measured using
mid-infrared lasers with wavelengths of 2.9 μm and 3.9 μm, respectively[1]. The present study aims to demonstrate
the principle of short-time measurement of tritiated water at the "60 Bq/cc level" using a cavity ring-down
spectroscopic (CRDS) measurement system with a mid-infrared laser. In order to achieve the above goal, (1)research on the CRDS system and (2) evaluation of hydrogen isotope composition under environmental conditions
and preparation of standard samples (subcontractor: Hirosaki University) were conducted this fiscal year. In (1), a
mid-infrared CRDS test was conducted. An optical bench (3m x 1.2m) was set up in the laboratory, and a designed
optical cavity was constructed on the optical bench. Next, a laser source essential for the CRDS measurement wasdeveloped. We have constructed a 4 μm wavelength Fe:ZnSe laser using a 2.9 μm wavelength Er:YAP laser, which
was originally developed by our group, as the pump source. By assembling a Littrow configuration cavity withbrazed gratings, we succeeded in developing a laser with a tunable wavelength range from 4.34 μm to 4.72 μm, a
maximum power of 22 mW, and a beam quality of M2=1.1. In (2), we prepared hydrogen isotope standard solutions
using commercially available heavy water standard solutions from several reagent companies to prepare stable
isotope heavy water samples as standard samples. We also purchased commercially available heavy water reagentsand prepared a standard sample of approximately 100 Bq/L. In addition, preparations were made for isotope ratio
measurements both indoors and outdoors. The performance of a low-background liquid scintillation counter was
evaluated, and the lower detection limit was confirmed to be approximately 0.6 Bq/L after 2,400 minutes ofmeasurement with a sample volume of 10 mL.

References[1] Hiroki KAWASE, Hiyori UEHARA and Ryo YASUHARA, “Preliminary Results of H2O and D2O Real-Time Measurement
Using Mid-IR Lasers with a Wavelength of 2.9 μm and 3.9 μm, Plasma and Fusion Res. 17, 1205057 (2022).

16

D04

Development of the Sample-return Technique for Fuel Debris Using
the Unmanned Underwater Vehicle
So Kamada1, Kazuya Nishimura1, Tetsuichi Kishishita2, Takahiro Oyama2,
Keisuke Okumura3, and Kennichi Terashima3, and Masayuki Hagiwara4

1

National Institute of Maritime, Port, and Aviation Technology, 2High Energy Accelerator Research
Organization, 3Japan Atomic Energy Agency, 4National Institutes for Quantum Science andTechnology

Abstract
To develop a fuel debris sampling system comprising a high radiation tolerance neutron detector, a camera,
an underwater Lidar, and end-effector with powerful cutting and collection capabilities, and a manipulator under
the Japan-UK joint research team. These are incorporated into an unmanned underwater vehicle (UUV). Also, wewill develop a system that supports remote control utilizing a VR system with multiple sensors.
1. Introduction
To retrieve the fuel debris, understanding its distribution and its properties of fuel debris is a critical issue.
Therefore, we will develop a technique to collect a small amount of fuel debris and detect a small number ofneutrons from the fuel debris under a high gamma-ray dose. Therefore, under the bilateral joint research system
between Japan and the UK, we will develop an unmanned vehicle-based debris sampling technology equippedwith a debris mapping tool that combines a highly radiation-resistant neutron sensor, lidar, and a camera.
2. Devices
2-1. Neutron sensor
We adopted Micro-Structured Neutron Detector (MSND) from RDT of the United States as a neutron sensor,
designed a signal processing circuit suitable for measuring weak MSND signals, and conducted characteristic tests
using a neutron source. In addition, using a water tank, the difference in the detector response to neutrons due tothe difference in water thickness was confirmed in the test.
2-2. UUV platform
We discussed the operation method of the neutron sensor to be developed, considering the interlocking withthe underwater camera and sonar. Subsequently, we designed the communication/control circuit and interface. In
addition, we examined tools for projecting image data from underwater cameras and sonar onto VR
2-3. Radiation environment simulations in PCV
Based on the internal investigation of Unit 3, we created a radiation source model assuming the fuel debrisof Unit 3 by using the PHITS code. Subsequently, a gamma ray transport calculation was performed, and it was
confirmed that the gamma-ray characteristics in the pedestal could be obtained.
2-4. Sampling tool
The Stewart platform required for testing manipulator development and UUV (Blue ROV2) were procured
and assembled. These were used to integrate the manipulator and control system. As a result of designing themanipulator and developing and evaluating several conceptual designs, The UK team created a design drawing forpreliminary actuators, component selection, manufacturing, and essential design specifications.
3. Japan-UK collaboration
We worked closely with CLADS to carry out the research. In addition, meetings were held to promoteresearch implementation plans, including meetings with UK team.

17



D05

Technologies for measurement and managementof radioactive materials and wastes
Junichi Hikosaka 1, Chris Carter 2 and David John 2
1Cavendish Nuclear Japan KK, 2 Cavendish Nuclear Ltd.

Abstract
As decommissioning and waste management challenges grow in complexity, technology and innovation isconstantly evolving to provide remotely operated technical solutions that move humans away from harm and toprovide data that can be used to optimise waste retrieval methods and waste management strategies.
1. Introduction
The poster shows examples of recent technologies that have been developed by Cavendish Nuclear for use
in waste retrieval and decommissioning projects.
2. Technologies for measurement and management of radioactive materials and wastes
PHUMS and AmCAM – a fast neutron detection
technology able to accurately and robustly assess the plutonium
holdup in single plant items such as gloveboxes as well as entire
facilities and cells. PHUMS and AmCAM measurements havebeen successfully undertaken in a number of alpha facilities in
PHUMS measurements

the UK
Particle Swarm Imaging or PSIM is an innovative
computational method developed by Cavendish Nuclear thatcreates an image of the source term using the information
obtained from multiple measurements at different positions
within a facility.

The type of measurement systems include

gamma spectrometry, dose probes, gamma imagers and neutron
Dose rate map overlayed on 3D model

measurement.
OptiSort is a used for fully automated radioactive wastesorting and segregation.

This technology includes a non-intrusive method for determining the chemical composition of thesurface of solid waste items using X-ray fluorescence (XRF) and
Raman Spectroscopy techniques to allow remote non-destructive

Schematic representation of X-ray

identification of the chemical composition of the surface of waste

fluorescencematerials.
In-Cell Decommissioning System (IDS) - The IDS uses its
spatial and radiometric scanning technologies that are integrated
with remote operation and virtual reality for safe and efficient
decommissioning operations.

IDS is used to map andcharacterize the contaminated cells using 3D LIDAR scan andImage of cell with radiological overlay

gamma spectrometry to create an accurate virtual environment

and laser cutter system.

and to conduct automated size reduction decommissioning
operations via remote VR control.

18

D06

Local Identification of Melted Multicompnent Materials and Mapping ofDistributed the Original-material using Laser-Induced Breakdown
Spectroscopic Multi-Spectra via Multivariate Analysis

Tomohiko Kawakami1, Sakiko Nagayama1,Koudai Okazaki1, Yuta Abe2, Etsuyo Makuuchi3,Masahiko Ohtaka2 and Yoshitaka Takagai4
1

Kaken., 2JAEA, 3Art KAGAKU, 4Fukushima Univ.

Abstract
To understand the hardness of fuel debris which is essential information for cutting and removing, alaser-induced breakdown spectroscopy was investigated as an estimating means of the hardness of materials.
Based on the analytical results of elements measured by LIBS, multivariate analysis was conducted estimating
the materials and the hardness of the laser-shot point of the samples.
1. Introduction
In the decommissioning of the Fukushima Daiichi Nuclear Power Plant, where boron carbide isused as a control material, the cutting and removing fuel debris, which is a mixture of materials
such as metals, oxides, and borides that are approximately twice as hard as oxides, is a challenge[1]. Laser induced breakdown spectroscopy (LIBS) is an analytical technique that cansimultaneously measure elements from light (especially boron and oxygen) to heavy elements(metals), and is expected to discriminate the analytical target based on elemental composition
ratio of the sample material. In this study, we present that the applicability of the compositional
distribution by multivariate analysis based on the analytical results measured by LIBS.
2. Method
A simulated fuel assembly that mimics the core of a boiling water reactor (BWR) containing
borides and oxides, was fabricated, and the test piece of the mimic fuel debris was then preparedby plasma heating (CMMR test piece)[2,3] (Figure 1(a)). The measurement was conducted byoriginal system which was combined devices such as LIBS, a displacement gage, and a X-Y
automation stage.

The data obtained by those devices was finally correlated and merged.

Afterthe measurement, a multivariate analysis was conducted using special software TIBCO Spotfire.
3. Resultsimulated pellet

Figure 1(b) exhibits the compositional distribution of the
CMMR test pieces constructed by multivariate analysis.
Each elemental area such as oxygen to zirconium, boron(a)

melted material(b)

carbide or control blade-derived boron or metal, and oxygen
was discriminated by the multivariate analysis in the local
points of samples (ex.

the simulated fuel pellets (ZrO 2),

cladding (Zr), and melted materials including Zr-O and Zr-B).

The observed mapping was agreed with the results ofmapping using an electron probe analyzer.
References[1] Takano, M. et. al., J. Nucl. Sci. Technol., 51 (2014),[2] Abe. Y, et al., ICAPP-17646 (2017),[3] Abe. Y, et al., J. Nucl. Eng. Radiat. Sci., 6, 021113 (2020).

19

cladding tube 2cm●simulated pellet●cladding tube
●melted material

Figure1 The results of multivariate analysis in
focusing on Zrmeasurement points
(a) photograph, (b) materials distribution

D07

Contamination of SGTS filter trains of Units 1 to 4
caused by the backflow of vent gas
Kenji Owada1, Masato Mizokami1, Takeshi Honda1, Shinya Mizokami1, Hironori Suzuki1,
Hiroshi Ide1, Cibula Michal1, Yutaka Hirokawa2, Yuta Sato2, Ryota Hashiwaki2
TEPCO HD, 2ATOX Co., Ltd.

1

Abstract
In order to clarify the cause of the contamination of the Standby Gas Treatment System (SGTS) filter trains ofUnits 1 to 4, TEPCO investigated the SGTS rooms, and obtained information on dose rates and contamination
around filter trains, vent lines, and rupture discs.
1. Introduction
The SGTS piping and filter train are connected to
the vent line (Fig.1), but are basically notcontaminated by the venting process. However, thefilter trains of Units 1 to 4 were confirmed to becontaminated in previous investigations. Since
Units 1 and 2 and Units 3 and 4 each share thestacks and Units 1 and 3 were estimated to beventedsuccessfully,

weassumethat

thiscontamination caused by the backflow of vent gas
from Units 1 and 3. Therefore, the investigation of

Fig.1 Schematic diagram of SGTS and vent line in Unit 3

SGTS rooms of Units 1 to 4 was conducted to obtain information to support that assumption.
2. Result of the investigation
In Unit 1, high dose rates was confirmed around the filter trains, with higher dose rates on the downstream sides.
In Unit 2, contamination downstream of the filter train was confirmed (Fig.2). In addition, almost no contamination
was confirmed around the rupture disc in Unit 2. In Unit 3, contamination was confirmed at the piping back flowing
from vent line to the filter train and around downstream of the filter train (Fig.3). In Unit 4, contamination was
confirmed downstream of the filter train (Fig.4). In addition, filter trains were opened in Units 3 and 4, anddosimetry and smear sampling were conducted on the filter surfaces.
3. Conclusion
This investigation clearly showed that the contamination of the filter trains of Units 1 to 4 was caused by the
backflow of vent gas from Unit 1 or 3.

Fig.3 SGTS piping and
Fig.2 Filter train of Unit 2

Filter train of Unit 3

20

Fig.4 Filter train of Unit 4

W01

Study on Rational Treatment/Disposal ofContaminated Concrete Waste Considering Leaching Alteration(1) Overview

Tamotsu Kozaki1, Takafumi Sugiyama1, Katsufumi Hashimoto1, Naoko Watanabe1, Shinichiro
Uematsu1, Daisuke Kawasaki2, Satoshi Yanagihara2, Daisuke Minato3, Toru Nagaoka3, Yuka
Morinaga3, Naofumi Kozaki4, Yoshikazu Koma4
Hokkaido Univ., 2Fukui Univ., 3CRI EPI, 4JAEA

1

Abstract
A series of investigations was carried out in order to discuss rational approaches to treat and/or dispose ofcontaminated concrete waste with a focus on underground concrete structures of Fukushima Daiich Nuclear Power
Station (FDNPS) in contact with contaminated water.
1. Introduction
Radioactive waste management is one of the integral issues in decommissioning of FDNPS. Undergroundconcrete structure, which account for a significant fraction of total concrete, is in contact with contaminated water
and when leaching alteration occurs migration behavior of radionuclides may change. The purpose of this study is
to investigate migration behaviors of radionuclides in relation to the properties of concrete altered by leaching, todevelop a model to calculate radioactivity concentration profiles and to estimate the amount of radioactive waste
in each class, and to analyze and evaluate multiple waste management scenarios.
2. Overview of the project
Hardened cement paste (HCP) specimens, which underwent simulated leaching, were used for sorption anddiffusion experiments with radioisotopes such as 125I, 137Cs, U and 14C to reveal that migration behaviors dependon radionuclides, their chemical species and degree of alteration. Leaching behavior of concrete was visualized bynon-destructive integrated CT-XRD method, and was quantitatively simulated by an improved simultaneoustransport model. Calcium Silicate Hydrate (C-S-H) and Calcium Aluminum Silicate Hydrate (C-A-S-H), which
are components of cementitious material, were synthesized, simulating different degrees of degradation, and wereanalyzed for microstructure using solid-state NMR. A method for statistically estimating radionuclide
concentrations of the concrete debris stored in solid waste storage facilities was developed using radiation dose
rate data. A simulation model for migration of radionuclides in concrete submerged in water was developed using
particle tracking method in 3D space to obtain radioactivity concentration profiles of multiple radionuclides.
Transport of Pu via the contaminated water, transfer models of 129I into contaminated water, and the source termof 14C were considered. In addition, several waste management scenarios were formulated for radioactive concrete
waste and the application of Safety and Environmental Detriment (SED) index was considered to evaluate andanalyze potential radiological risks arising from different waste management processes.
3. Summary
Migration behaviors and properties of altered concrete is examined and using the acquired data, a model isdeveloped to predict radioactivity concentrations, and waste management scenarios are evaluated.
Acknowledgements
This study was a part of the outcomes of the “Nuclear science and technology and human resource development program”,“Study on rational treatment/disposal of contaminated concrete waste considering leaching alteration” (2020−2022). Part ofthis work has been performed in the facilities of the Central Institute of Isotope Science, Hokkaido University.

21



W02

Study on rational treatment/disposal of contaminated concrete wasteconsidering leaching alteration:(2) Effect of leaching alteration on 137Cs migration in cement paste
Shinichiro Uematsu1, Kyoya Watanabe1, Keisuke Matsumoto1, Naoko Watanabe1,
Tamotsu Kozaki1, Yuka Morinaga2, Daisuke Minato2 and Toru Nagaoka2
Hokkaido Univ., 2CRIEPI

1

Abstract
Sorption and diffusion behavior of 137Cs in hardened cement paste altered by leaching was studied. Effectsof the leaching alteration of cement paste were found for the sorption coefficient, diffusion coefficient, and theactivation energy for diffusion.
1. Introduction
Radioactive waste management of concrete materials is an important issue in decommissioning of the
Fukushima Daiichi Nuclear Power Station (FDNPS). Concrete materials in contact with water in the FDNPS may
have been altered by leaching. In this study, sorption and diffusion coefficients of 137Cs were experimentally
determined for hardened cement paste (HCP) with and without alteration by leaching treatment to clarify the effectof leaching on migration behavior of 137Cs in HCP.
2. Materials and Methods
Cement paste prepared from ordinary portland cement with the water-to-cement ratio of 0.36 was cured inan equilibrated cement solution for 28 d at 323 K. The leaching of the HCP was conducted by immersing the HCP
in 6 M NH4NO3 solution for 5 d. Apparent diffusion coefficients (Da) of 137Cs were determined from one
dimensional non-steady diffusion experiments at contrasting temperatures
between 288−323 K. The apparent sorption coefficients (Kd) of 137Cs wereTable 1. Sorption coefficients(Kd) of 137Cs on HCP samples.

determined for ground HCP samples by a batch method.

HCP sample

Kd(L kg−1)

3. Results and Discussion

Unaltered

1.0±0.1

Altered (6 M NH4NO3)

17±3

The Kd value was higher for altered HCP than for unaltered HCP as
indicated in Table 1. Fig. 1 shows the temperature dependence of the Da of
Cs in unaltered and altered HCP samples. The Da values were different

137

between the unaltered and altered samples at each temperature except at 313
K. In addition, the activation energy for diffusion (Ea) of 137Cs for the altered
HCP sample was ca. 60 kJ mol−1, which was higher than that reported for
unaltered HCP samples (37 ± 2 kJ mol−1) [1]. Since Ea was determined fromthe temperature dependence of Da, it may include the effect of the sorption
enthalpy. Therefore, our result of the increase of Ea is likely to be caused by(1) changes in sorption behavior by alteration of HCP, (2) changes in thedominant diffusion process as a consequence of alteration, or both.
References [1] Morishita et al., ICONE26‑82570, London (2018).

Fig. 1. Temperature dependenceof the apparent diffusion
coefficients (Da) of 137Cs.

Acknowledgements This study was a part of the outcomes of the “Nuclear science and technology and human resource development
program”, “Study on rational treatment/disposal of contaminated concrete waste considering leaching alteration” (2020−2022) fundedby the MEXT. A part of this work has been performed in the facilities of the Central Institute of Isotope Science, Hokkaido University.

22Quantitative Evaluation of Long-Term State Changes ofContaminated Reinforced Concrete Considering the Actual
Environments for Rational Disposal

W03

Ippei Maruyama1, Kazuo Yamada2, Kazutoshi Shibuya3, Yoshifumi Hosokawa4,
Yo Hibino5, Yasumasa Tojo6 and Yoshikazu Koma7
1
Univ. of Tokyo, 2National Institute for Environmental Studies, 3Taiheiyo Consultant Co. Ltd.,
4
Taiheiyo Cement Corp., 5Nagoya Univ., 6Hokkaido Univ., 7Japan Atomic Energy Agency
Abstract
In the decommissioning of concrete structures at the Fukushima Daiichi NPP, it is important to estimate theamount and concentration of waste. In this study, a prediction method of radionuclide penetration behavior
considering the actual environment is investigated for quantitative prediction of contamination concentration
distribution in concrete components. In this report, the outline of this study is described.
1. Overview of the study
1-1. Examination of condition setting of concrete members assumed in the actual environment
In order to set the condition of concrete members as expected in a real environment, the mesoscale concept
have applied to mechanically evaluate the amount, width, distribution, and depth of cracks in concrete members.
To evaluate the behavior of cracks, data on deformation and moisture movement due to drying and reabsorptionof mortar have been obtained. In parallel, a rigid-body spring model (RBSM) was used to develop a program that
can take into account changes in concrete age and temperature, water, and stress conditions.
1-2. Examination of penetration behavior analysis method considering the condition of concrete members
In order to evaluate the long-term penetration behavior of radionuclides into actual concrete matrices, dataon sorption were obtained and a mathematical model was developed. And to evaluate the penetration behavior ofradionuclides into concrete through cracks, the penetration of Cs and Sr into mortar specimens with different crack
widths at concentrations of various ions equivalent to those of cooling water released immediately after theaccident was evaluated by autoradiography.
1-3. Experimental evaluation of penetration behavior of α-radionuclides on concrete members
In order to investigate thepenetration

behaviorofα-radionuclides into the concrete
matrix,long-term immersion

tests of cement paste specimens
were initiated, and distribution
ratiosofα-radionuclides

toaggregate, paint, and steel bars
were measured.
2. Future Action
Based on the results of theabove studies, we will estimate

Figure 1. Framework of the study

the contamination distribution
and attempt to quantitatively predict the volume of waste and contamination concentration.
Acknowledgement
This work was supported by JAEA Nuclear Energy S&T and Human Resource Development Project throughconcentrating wisdom Grant Number JPJA20P20333545

23

W04

Challenge of Novel Hybrid-waste-solidification of Mobile Nuclei
Generated in Fukushima Nuclear Power Station and Establishmentof Rational Disposal Concept and its Safety Assessment

Masahiko Nakase1, Miki Harigai1, Shinta Watanabe1, Ryosuke Maki2, Hidetoshi Kikunaga3,
Tohru Kobayashi4, Tomofumi Sakuragi5, Ryo Hamada5 and Hidekazu Asano5
1

Tokyo Institute of Technology, 2Okayama University of Science, 3Tohoku University., 4Japan Atomic
Energy Agency, 5Radioactive Waste Management Funding and Research Center

Abstract
The present study aims to establish the rational waste disposal concept of various wastes generated in
the Fukushima Nuclear Power Station by the novel hybrid-waste-solidification concept. The radioisotope insynthesized ALPS sediment wastes, and AREVA sludge wastes, which were secondary wastes generated by water
decontamination, were immobilized in phosphate waste form. They were processed to the stabilization treatment
such as Spark Plasma Sintering (SPS) or Hot Isostatic Pressing (HIP) with well-characterized materials such as
SUS and Zircalloy, which make the long-term stability evaluation and safety assessment possible.
1. Introduction

The decommissioning of the 1F NPP

station is in progress, and the retrieval of fuel debris will also
be started The next issue is the disposal of various wastes,
including some of the important elements such as Iodine, which
is a mobile element and toxic alpha emitters of Actinide. Not
only developing the waste form but also the compatibility withthe actual disposal s essential. Therefore, we started the new
project on hybrid-waste-solidification to connect the waste
form study to the future disposal scenario.
2. Outline of project and goals

Fig.1 Schematics of HIP and SPS treatment

In the project, the research was divided into 5 themes as shown in Fig 2. Intheme 1, some of the wastes such as ALPS, AREVA sediment wastes, and their stabilized waste forms aresynthesized as well as a mobile and difficult element of iodine such as iodine apatite. In theme 2, structural study
on wastes as well as leaching and ,  and  irradiation behaviors are studied. Theme 2 includes both microscopic
and macroscopic analyses such as SEM, TEM, XAFS, etc. In theme 3, first principle and molecular dynamics
calculations were implemented to elucidate the mechanism of
HIP and SPS treatments. In theme 4, synthesis of hybrid
waste form by SPS and HIP and characterization are ongoingby varying the matrix materials. In theme 5, a scenario study
on the disposal of hybrid waste form based on the inventory
study is ongoing, and a safety assessment based on the
corrosion behavior of the matrix material, and migration
calculation are iimplementing.
3. Conclusion

1. Synthesis of primary waste form
2. Structure and characterization study
3. Chemical simulation approach
4. Synthesis of hybrid waste form
5. Scenario and safety assessment
studies of disposal
Fig.2 Research themes of the project

A new research project on hybrid-waste-solidification and safety assessment was started lastyear. The outline of the project and the scope, and some of the latest results will be shown in the poster presentation.
4. Acknowledgment

This work is funded by the JAEA Nuclear Energy S&T and Human Resource

Development Project through concentrating wisdom Grant Number JPJA21F21460873.

24

W05

SIAL® : Geopolymer solidification technology approved by
Slovak / Czech Nuclear Authority
Milena Prazska and Marcela Blazsekova , Jacobs (Slovakia)
Hisashi Mikami , Isamu Kudo and Nobuyuki Sekine , Fuji Electric Co.,Ltd.

Abstract
We present the features and the performance records of SIAL® geopolymer solidification technology islicensed by both the Slovak (ÚJD SR) and Czech Nuclear (SUJB) regulators, and the technology has been usedsuccessfully for 20 years. More recently, geopolymers have been noted as an immobilization technology andwhich shows potential of immobilizing resins generated by treatments of contaminated water at Fukushima
Daiichi Accident. Reference will also be made to some of the activities being undertaken in Japan todemonstrate its performance.
1. Introduction
The Nuclear Power Plant (hereinafter called NPP) Unit A1 located in Jaslovske Bohunice, which wascompleted in 1972 and had been operated for 5 years until two accidents happened in 1976 and 1977. After thesecond accident (INES level 4), NPP Unit A1 was permanently shut-down for decommissioning. Damaged fuelassemblies and claddings in the accidents caused contamination of strontium-90, caesium-137 and transuranic.
As a result of a long-term corrosion of barrier’s materials, highly contaminated sludge were accumulated, andthe waste could not be effectively immobilized with using conventional methods such as Cementation or
Bitumen treatment due to negative impact on physical-chemical properties and high specific activity(caesium-137) of the waste. This challenge led to developing SIAL® solidification technology. Today, SIAL® isproven and widely used for on-site solidification of radioactive waste streams such as sludge, resins, sorbents
and organic liquids. This is directly applicable to the conditions at Fukushima Daiichi as well as other Japanese
NPPs where the waste streams are not well understood.
2. Feature
SIAL® matrix can provide efficient and practical on-site
treatment of radioactive waste streams at room temperature,and can incorporate times as much wastes as Cement matrix
equivalent on average.
The equipment used to deploy SIAL® solidification
technology is also modular, flexible and versatile.

It canencapsulate waste streams quicker than Cementation, and canbe applied under water.

Figure 1 SIAL® solidified samples

SIAL® solidification technology can realize
higher compressive strength and lower leachability
compared to

Cementation, and posing a low firerisk and excellent physical stability in the presenceof frost and water (no distortion or cracking).

Figure 2 Sludge and slurry waste streams crosssection observation

3. Performance Record Example
About 3×106 kg of radioactive waste streams (resins, sludge and crystalline borates) is successfully
immobilized using SIAL®. This comprehensive scope of works started with licensing processes, solidification,and then were followed by decontamination and cleaning of the workplace post cleanup and transports of all
equipment to off-site.

25

W06

Waste Management Symposia: The Annual Phoenix Radioactive
Waste Conference Exchanging Knowledge from Around the World
Kazuhiro Suzuki1, Gary Benda2 and Akira Ono3
１
WM Symposia Board of Director, ２WM Symposia Deputy Managing Director/
Program Advisory Committee (PAC) Chair and 3 WM Symposia PAC Member

Abstract
WM Symposia’s annual Waste Management (WM) Conference attracts thousands of registrants from around theworld and is widely regarded as the premier international conference for the management of radioactive waste andrelated topics.
1.

Introduction

The WM2023 Conference, the 49th version, will beheld February 26 - March 2, 2023 at the Phoenix Convention
Center in Phoenix, Arizona. Conference theme is “Planningfor the Future: Innovation, Transformation, Sustainability”
with a strong Conference focus on promoting the next
generation of students and radwaste management
professionals, encouraging exchange of expertise to young
professionals around the world. WM2023 will feature over
500 papers and more than 80 panel discussions in over 160
technical sessions, complemented by nearly 200 exhibiting
companies, the industry’s largest.
2.

Poster
The poster provides Conference details and describes theTechnical Panel, Poster and Oral Sessions, Exhibitor,
Student and Sponsorship program, as well as theopportunity to network with over 2,000 industry
specialists and managers from more than 35 countries andlearn of trends and developments from the most seniorindustry managers around the world. WM2023 has special
programs aimed at encouraging participation of theworld’s leading companies in the exhibit hall, including a
program for live product demonstration.

3.

Conclusion

The Conference promotes, among Japanese and professionals fromaround the world, a broad exchange of knowledge in 12 Technical
Tracks, including technologies, operations, safety, security &safeguards, waste management issues, decommissioning anddismantling, environmental remediation, STEM education andadvanced nuclear reactors. The deadline for submittal of Abstracts
for WM2023 is August 26, 2022. Details are shown on the poster.
Reference: www.wmsym.org.

26

W07

Integrated Waste Management and its Application to
Nuclear Decommissioning and Dismantling Projects
Michelle Dickinson, Antonio Guida and Bill Miller
Jacobs UK Ltd., Decommissioning and Regeneration Solutions

Abstract
Integrated Waste Management involves making waste management an integral part of the decision making,planning and execution processes for decommissioning and dismantling (D&D) projects. Waste is the main productof any of these projects, and waste management is often a limiting factor in project schedules. The success of any
D&D project is therefore significantly impacted by the way in which waste management is planned and performed.
Introduction
Waste management has often been an afterthought in D&D projects, and this can result in waste disposition
delaying the overall project completion or significantly increasing cost. Problems include wastes produced without
appropriate characterisation; inadequate sorting and segregation; not meeting waste acceptance criteria (WAC) etc.
With Integrated Waste Management, waste management experts are involved early and throughout the whole D&D
project. Aspects to be considered as part of "integration" can include, depending on the circumstances:•

Integrating decommissioning and waste management
plans to avoid generating wastes without a treatment route•

Integrating strategic planning with regulatory complianceand the design / operation of treatment facilities•

Integrating decision-making across all radioactive waste
streams at national and site to achieve economies of scale•

Integrating treatment and decontamination methods toreduce waste volumes and maximise recycling•

Integrating national and local infrastructure to maximise efficiency and avoid duplication of facilities•

Integrating national and private companies to incentivise improved and cheaper waste treatment methods
Benefits include reduced lifetime costs, regulatory compliance and improved safety of the workforce.

Experience
At Jacobs we apply Integrated Waste Management in D&D projects in the UK, US and other countries. For
example, Jacobs works with every nuclear site in the UK and across the entire lifecycle of radioactive waste:•

Planning, including decision making on Best Available Techniques (BAT) for waste treatment and disposal•

Characterisation, including performing statistical sampling and surveys of buildings and bulk wastes•

Waste retrieval and treatment, including high active tank wastes treated using cement or geopolymer•

Waste packaging and transport, including packaging of wastes to meet formal WAC for disposal•

Interim storage and waste disposal, including preparing operational and disposal safety cases for licensing.

Conclusion
We welcome the opportunity to demonstrate our experience and to explain the benefits of Integrated Waste
Management to D&D projects in Japan that are limited by waste management plans and disposal infrastructure.

27

E01

Prototype of Differential Amplifier Circuits
Based on Radiation hardened H-diamond MOSFET (RADDFET)

Hiroki Fukushima1, Manobu M. Tanaka2, Hitoshi Umezawa3, Hiroyuki Kawashima4, Yusei Deguchi1,
Tadashi Masumura1, Naohisa Hoshikawa1 and Junichi H. Kaneko1
Hokkaido Univ., 2KEK, 3AIST

1

Abstract
A prototype differential amplifier circuit using hydrogen-terminateddiamond MOSFETs (RADDFETs), which have excellent resistance tohigh temperatures and radiation, was fabricated by creating an Ltspicemodel based on the characteristics of the diamond FETs that make up the
RADDFET. A device with 14 MOSFETs mounted on a 4 mm squarediamond substrate was used. The actual circuit achieved 4.7 times
amplification compared to 4.4 times in the simulation.
1. Introduction
After the accident at the Fukushima nuclear power plant, theb)performance required for equipment used in containment vessels to copewith severe accidents, such as operating temperature: 300°C (72 hours),
integrated dose: 5 MGy, etc., has been increasing. In this study, as part ofthe development of electronic equipment for severe accident response, we
designed and fabricated a differential amplifier circuit using diamond
MOSFETs (RADDFET[1]), which are required in preamplifiers, andmeasured its characteristics.
2. Experiment2-1. Preparation of RADDFETc)
D
Intrinsic layers were homo-epitaxially grown on a 4 mm squarediamond single-crystal substrate by microwave CVD and the surface wasShydrogen-terminated. A 40 nm oxide insulating film (Al2O3) was depositedG
as a passivation layer, and 14 diamond MOSFETs (called RADDFETs)
were fabricated. RADDFETs have a proven track record of operation at
Figure1:
3 MGy integrated dose and 300°C. The production yield of thea) Appearance of prototype
RADDFETs was 84%.
circuit
2-2. Circuit Simulationb) 14 FETs fabricated on
An LTspice model was created from the Id-Vd characteristics of
diamond substrate
RADDFETs. A differential amplifier circuit was designed, thec) Enlarged view of FET device
external supply voltage was specified so that all FETs enter thesaturation region, and the input-output characteristics were
investigated by simulation.
2-3. Circuitry
The prototype RADDFETs were mounted on a PCB board
to create the prototype circuit shown in Figure 1. In this circuit,
RADDFETs with similar characteristics are combined toconstruct a differential amplifier circuit.
3. Result
Figure2: Example of input/output
Based on simulations, a voltage amplification of about 4.4
characteristic measurement
times was designed, but in the actual circuit, non-inverting(Real circuit, one-sided amplification)
amplification with an amplification factor of 4.7 times could beobtained by adjusting the external voltage and inverting and non-inverting outputs could be obtained
simultaneously by further adjusting the voltage. This achievement is a major step forward toward the practicalapplication of radiation-resistant, high-temperature-operating preamplifiers.

Acknowledgements
Part of this work was supported by the "SCORE" project of the Japan Science and Technology Agency. A part of this work wasalso supported by the Nuclear Energy Systems Research and Development Program (MEXT).
References[1] T. Yamaguchi et al., Appl. Phys. Lett.118, 162105

28

E02

Optically reconfigurable gate array VLSI without any common
signal
Sae Goto, Minoru Watanabe and Nobuya Watanabe
Okayama University

Abstract
Very large-scale integrations (VLSIs) are vulnerable to radiation. VLSIs used in strong radiation
environments such as space environment and nuclear power plants are always affected by radiation. Soft-errors
and permanent failures on current VLSIs are caused by radiation. Therefore, we’ve been developing radiationhardened optically reconfigurable gate arrays. This paper presents a proposal of a new optically reconfigurable
gate array VLSI without any common signal.
1.

Optically reconfigurable gate array
Optically reconfigurable gate arrayconsists

of

three

components:aprogrammable

gate

array

a

VLSI,

Laser Array

holographic memory, and a laser array. The

Holographic Memory

programmable gate array includes logic
blocks, switching matrices, I/O blocks [1].

Optically Reconfigurable

Although the programmable gate arraystructure

is

same

as

that

of

Gate Array VLSI

field

programmable gate arrays (FPGAs), theconfiguration procedure is different from

Figure 1. Structure of optically reconfigurable gate arrayeach other. The configuration of optically reconfigurable gate arrays can be executed in a perfectly parallel. As aresult, even if a part of the programmable gate array and/or a part of the configuration circuit are broken byradiation, the operation on the programmable gate array can be executed continuously by reconfiguring the
programmable gate array. In order to support such parallel configuration, optically reconfigurable gate array VLSIs
have many photodiode circuits to detect optically-applied configuration contexts generated from the holographic
memory. However, such optical configuration procedure requires two common signals: a refresh (nREF) signal tocharge the junction capacitance of photodiodes and a configuration clock (CCLK) signal for flip-flops. The
common signals were weak points for radiation.
2. Optically reconfigurable gate array VLSI without any common signal
We have developed a new optically reconfigurable gate array VLSI not using such common signals. The
configuration circuit never uses flip-flops so that the configuration clock (CCLK) signal is perfectly removed.
Also, the refresh (nREF) signal is adjusted constantly to an intermediate value which is provided only on a metal
wire without any buffer.
Reference[1] Y. Takaki, M. Watanabe, “Optical multi-context blind scrubbing for field programmable gate arrays,” IEEE Photonics
Journal, Vol. 12, Issue 6, 7801411, Dec., 2020.

29

E03

Total-ionizing-dose tolerance of an optically reconfigurable gatearray VLSI
Kaho Yamada, Takeshi Okazaki, Minoru Watanabe and Nobuya Watanabe
Okayama University

Abstract
Currently available Field Programmable Gate Arrays (FPGAs) use a serial configuration circuit. The serial
configuration circuit is too vulnerable to radiation. Therefore, we have been developing an optically reconfigurable
gate array VLSI with an optical configuration function. This paper presents the total-ionizing-dose tolerance of an
optically reconfigurable gate array VLSI.
1. Introduction
At the Fukushima Daiichi Nuclear Power Station, a decommissioning work is currently being executed.
However, when using a robot in such environment, there is a problem that integrated circuits inside the robot areeasily broken by radiation. Therefore, we have been developing a radiation-hardened Optically Reconfigurable
Gate Array (ORGA)[1][2]. In the ORGA, circuit is configured in parallel by light, so even if a part of theconfiguration circuit is broken by radiation, the configuration of the remaining region can be executed correctly.
2. Total-ionizing-dose tolerance of an ORGA-VLSI
An ORGA consists of a laser array, a holographic memory, and an ORGA-VLSI. Since the semiconductor
devices are vulnerable to radiation, in the ORGA, we use a repairable VLSI concept. If we can use a programmablearchitecture for VLSIs, even if a part of a programmable gate array is broken by radiation, the non-damaged areaof the VLSI can be used continuously so that the radiation-tolerance of the VLSI can be increased. To realize it,
we have introduced parallel configuration architecture onto the ORGA instead of serial configuration architecture.
Since laser arrays and holographic memories are relatively strong for radiation, ORGAs can achieve high-radiation
tolerance. In this test, an ORGA-VLSI was exposed to 290 Mrad gamma radiation. Under the radiation
environment of the Fukushima Daiichi Nuclear Power Station, although the existing integrated circuits will not
operate normally in 10 hours, the ORGA can normally operate for 120 days.
3. Conclusion
In this test, only a 2.4 % performance degradation was confirmed at an absorbed dose of 290 Mrad. The small
degradation can be out of consideration.
References[1]

T. Fujimori, M. Watanabe, “Optically reconfigurable gate array using a colored configuration,” Applied Optics, Vol. 57,
Issue 29, pp. 8625-8631, Oct., 2018.

[2]

T. Fujimori, M. Watanabe, “Parallel light configuration that increases the radiation tolerance of integrated circuits,” Optics
Express, Vol. 25, Issue 23, pp. 28136-28145, Nov., 2017.

30



G01

Development of technology to reduce environmental problems via
innovative water purification agents

Naoki Asao1, Taketoshi Minato2, Natsuhiko Yoshinaga3, Kazuto Akagi3, Joseph Hriljac4,
and Neil Hyatt5
Shinshu Univ., 2Institute for Molecular Science, 3Tohoku Univ., 4Diamond Light Source,

1

Univ. of Sheffield

5

Abstract
An efficient fabrication method to produce granular adsorbents for strontium without granulation process
was developed, and titanates and silicotitanates were obtained as grains with suitable size for column operation in
the decontamination process. They exhibited remarkable ion-exchange properties for removal of strontium ionsfrom simulant contaminated water and behaved as reusable adsorbents without any significant loss of activity.
1. Introduction90

Sr is one of the dangerous radioisotopes because its chemical similarity to Ca promotes the accumulationin bones and teeth of the human body, which causes long-term internal exposure. Although the decontamination
work at Fukushima continues using various adsorbents, it is still desirable to develop highly efficient andinexpensive adsorbents. Furthermore, an adsorption process inevitably generates used adsorbents as secondary
pollutants, which are required to be safely stored under strict control for a long time. Therefore, the suppression ofgeneration of used adsorbents has emerged as an urgent issue. To overcome the above-mentioned problems, thepresent interdisciplinary collaboration research has been launched.
2. Results and discussion
Thefabricatedtitanates

andsilicotitanates were analyzed by SEM, XPS,
and PDF, and these data together withadsorption results were used for data analysis
andtheoretical

calculations,leadingtoaccurate prediction of the improved fabrication
method of adsorbents. The resulting titanatesand silicotitanates exhibited remarkable ionexchange properties for removal of strontium
ions from simulant contaminated water.
Furthermore, they could be reused repeatedly
without any significant loss of activity.
3. Conclusion
Granular adsorbents having high Sr uptake ability were fabricated without granulation process. Further
studies to elucidate the adsorption mechanism is in progress. Additional studies to produce suitable waste forms
to contain the Sr species produced during recycling and for the final spent absorbent are also in progress.

31

G02

Fukushima Forests: Decontamination and Restoration
D.Hildebrand1, S.Suzuki2, H.Asama2, C.Casto3, V.Dudarchik1, V.I.Kislyi1,
Y.Kawashima1, P.Molchanov1, R.Foster1, C.Chow1, V.Boksha1
1

NeuroSyntek, 2University of Tokyo, 3Casto Group

Abstract

Needle/branches(<1%)

Cs levels (kBq m–2)

NeuroSyntek [1] and partners are actively involved into the
Fukushima region revitalization – leveraging our experiences from
Hanford (USA), Chernobyl, Daiichi, and Silicon Valley (USA).
There are many challenges to decontaminate/clean-up Fukushima
forests. Corresponding opportunities are enormous.

Bark (<1%)
Wood (1%)
Litter (5%)
Soil (94%)

Years after accident
137

From Problems to Opportunities

Cs dynamics in Fukushima forests [2]Two thirds of Japan land is forest. More than 40% are

New top layer of fertilized soil

planted forests, such as cedar and cypress; most with

Middle soil layer

mountain terrain. With aging population forestry it is hard

Former top (contaminated) soil

to sustain with traditional human-centric technologies.

Complex sorbent, blocking layer

Significant cesium contamination from Daiichi tonorthwest causes very serious concerns about groundwater

Forest soil recovery: layers replacement

as cesium penetrates into the soil and downstream.
Solutions & Products: We suggest Autonomous and Semi-Autonomous Forestry

D

D

Robotic Systems with core AI concept – with unique Intellectual Property set (IP). IP

C

examples: Robotic Machine & Method; Robotic Off-Road Vehicle; System & MethodAof Forest Logging; Cesium sorbent & placement method. We also envision

C

B

A

Training and Gaming capabilities engaging Fukushima RTF (Robotics Test Field),development of Game and Media Ecosystem for Fukushima Reconstruction, including
Forest Clean-up Reality Game. Software Segment & Robotic Contest: A competitiveenvironment is proposed to start development of a remote-controlled autonomous
technology for equipment used in agriculture and forestry. First stage deals with pure
software developments of most critical elements of robotic technologies. At the second
stage the developed technologies are deployed in-the-field on dedicated test robotic
CARRY

EQUIPMENTSENSORS

B

C

vehicles.
A

Acknowledgments: Hajimu Yamana, Yoshimi Ota, Paul Dickman,Haruaki Matsuura, Reconstruction Agency TeamIP examples. Modular
robotic off-road and forest
vehiclesReferences[1] NeuroSyntek Robotics, Gaming & AI, private communications, prototypes & business plans, 2000-2022[2] Onda, Y., Taniguchi, K., Yoshimura, K. et al. Radionuclides from the Fukushima Daiichi Nuclear Power Plant in terrestrial
systems. Nat Rev Earth Environ 1, 644–660 (2020). https://www.nature.com/articles/s43017-021-00198-0

32

J01

JAEA’s Efforts for Human Resource Development andRegional Commitment in Fukushima
Satomi Ito1, Yoshihiro Tsuchida1, Satomi Kakutani1
1Japan Atomic Energy Agency

Abstract

Sector of Fukushima Research and Development, Japan Atomic Energy Agency (JAEA)
conducts human resource development and regional contribution activities along with research anddevelopment.
1. Introduction
Research and development for decommissioning of Fukushima Daiichi NPS and environmental
restoration of Fukushima Pref. are the main missions that JAEA is working on. Considering thatthese activities will be conducted over the medium to long term, the following points are important:a) securing and fostering human resources, b) participation of local companies, and c) technology
transfer to the local businesses and field implementation.
2. Efforts for Human Resource Development
JAEA is working to develop human resources through cooperative agreements with KOSEN anduniversities, and through the implementation of the Nuclear Energy Science & Technology and
Human Resource Development Project (Project name: Mission H), which includes giving lectures
to students, accepting research students, and holding the Creating Robot Contest for
Decommissioning. Through these efforts, the promotion for understanding of decommissioning
and environmental restoration and the development for human resources who will lead the future
together with some educational institutions are being implemented in JAEA.
3. Technology development with local industries in Fukushima
In collaboration with local companies in Fukushima, we are advancing analytical equipment andmonitoring technologies for on-site implementation. By transferring these technologies to the local
companies for practical use, we will contribute to the commercialization of the developed
technologies and the industrial recovery of Fukushima.

33

J02

Summary of JAEA's efforts for fuel debris retrieval
Shin-ichi Koyama, Naoya Kaji, Satomi Kakutani
Japan Atomic Energy Agency

Abstract
Japan Atomic Energy Agency (JAEA) will build a knowledge base that enables safety and risk assessment
through basic research as a sherpa for TEPCO HD and NDF for the retrieval of fuel debris from TEPCO's
Fukushima Daiichi Nuclear Power Station (1F).
OutlineThe efforts implemented by JAEA are classified as follows. Research and development by utilizing new
constructed facilities for fuel debris retrieval work, such as demonstration of fuel debris retrieval device and dose
distribution evaluation for reduction of radiation exposure during the work (①, ②, ③, ④). Facility maintenanceand technological development related to analysis of fuel debris (④, ⑤, ⑥). Basic research related to core status
estimation and safety / risk assessment inside the containment vessel / reactor building (③, ⑦, ⑧, ⑨).
① Support for IRID full-scale mockup test at Naraha Remote Technology Development Center.
Mockup test and operation training for fuel debris retrieval from Unit 2.
② Visualization of environment, dose and radiation source distribution for reducing exposure.
Development of a system that can study the reduction of radiation exposure of workers under high dose without
exposure for the preparation work for seeking an access route in R / B.
③ Database / sharing of radioactive source / dose distribution evaluation and core status in PCV.
Comprehensively evaluate the radiation source and dose rate distribution in the PCV and inside the building,and provide and share information to related organizations. Provide and share the knowledge of the inside of thecore and debris characteristics as a database "debris Wiki".
④ Development of analysis and research facilities for radioactive waste and fuel debris at the Okuma
Analysis and Research Center.
Under construction of a radioactive material analysis / research facility for analysis and research to understandthe properties of radioactive waste and fuel debris on the adjacent to 1F.
⑤ Development of remote, quick and simple analysis technology for fuel debris.
⑥ Development of α-ray imaging detector for evaluation of exposure.
Development of a portable laser-induced emission analysis (LIBS) device with high radiation resistance thatcan be applied to the 1F and succeeded in analyzing the sample collected from Unit 2 on the 1F site. Development
an α-ray imaging detector to quickly evaluate the concentration and particle size of α-ray dust.
⑦ Verification of fuel debris characteristics by simulated experiments and comprehensive evaluation
utilizing 1F site information.
Evaluation of metal debris generation behavior by manufacturing large scale equipment and testing to reproduce
accident.
⑧ Elucidation of basic fuel debris characterization and aging degradation.
⑨ Confirmation of cooling status in the reactor core by thermal analysis of fuel debris.
Collection of knowledge necessary for long-term storage safety evaluation. Evaluation of the temperature
distribution of fuel debris and its surroundings when water is injected and stopped for fuel debris retrieval.

34

J03

R&Ds on Waste Management Technologies for Fukushima Daiichi
NPS with National and International Collaboration
Yoshikazu KOMA, Naoya KAJI 1
1

Japan Atomic Energy Agency (JAEA)

Abstract
Japan Atomic Energy Agency has been conducting research and development on waste management
technologies for Fukushima Daiichi Nuclear Power Station in the field of characterization/analysis, safe storage
and processing/conditioning with help of national and international collaboration.
1. Introduction
Decommissioning and its waste management requires R&D for characterization, safe storage, processing anddisposal in parallel to provide promising technical options for future execution. In order to investigate suchtechnologies for the unconventionally generated waste, knowledge and experiences should be assembled fromnational and international organizations and utilized. JAEA is conducting R&Ds with wide collaboration.
2. Characterization/Analysis
Radioactive nuclides of long half life are important target for analysis due to its effect to disposal safety. The
analysis procedure is complex and time consuming, thus, JAEA collaborate with laboratories of privatecompanies to provide data for related decommissioning activities. Furthermore, Radioactive Material Analysis
and Research Facility Laboratory – 1 has been completed at the Fukushima Daiichi site to start analysis in thenear future. Essential methodologies including analysis planning, which is based on Data Quality Objectives
process and Bayesian statistics [1], as well as inventory estimation [2] are the result of collaboration. International
knowledge is helpful to accelerate R&Ds above and JAEA participated in NEA expert group [3].
3. Safe Storage and Processing/Conditioning
The secondary waste from water decontamination is designated as the target, which hazard should bedecreased. To investigate safe storage of such a waste including adsorbent with highly concentrated radioactive
cesium, collaboration with the equipment provider is indispensable and the result is already utilized [4]. For
conditioning such waste, solidification with cement and alkaline activated material (AAM, or also known as
geopolymer) is investigated in the view of highly radioactive and water-bearing waste [5].
4. Conclusion
Waste management for decommissioning should be efficiently conducted, and JAEA promotes
national/international collaboration for establishing potential technologies.
References[1] C. Pyke et al., Nucl. Eng. Technol., 54(1), 414-422 (2022). [2] D. Sugiyama et al., J. Nucl. Sci. Technol., 58(4), 493–506(2021). [3] Nuclear Energy Agency, NEA/RWM/R(2020)2 (2021). [4] I. Yamagishi et al., J. Nucl. Sci. Technol., 51(7-8), 10441053 (2014). [5] V. Cantarel et al., J. Nucl. Sci. Technol., 58(1), 62–71 (2021).

35

J04

Experimental research related to formation of debris using largescaled test facility in CLADS
Yuji NAGAE

Japan Atomic Energy Agency, Collaborative Laboratories for Advanced Decommissioning Science
Abstract
JAEA/CLADS has conducted two types of large-scaled experiments using the LEISAN (Large-scale

Equipment for Investigation of Severe Accidents in Nuclear reactors) facility at Tomioka-site . In-situ and post
investigation provide all of us updates of understanding for debris formation inside and outside reactor pressure
vessel, combined severe accident analysis.
1. Introduction
Investigation of the accident study and recovery actions for TMI-2 and a lot of tests have given us accident
progression for Pressurized Water Reactor. There is a lack of knowledge for Boiling Water Reactor (BWR), in
which a large amount of metal be used in control blades, control rod drive systems etc. A set of control blades
included rod would be possibly degraded by reaction between stainless steel of control blade and at the early stageof accident. Furthermore, these metallic debris would possibly relocate to lower plenum in accident progression,and by reaction between the metallic debris and structural materials or weldments. We would like to introduce
outline of two kind of tests;
1.

Experiments for metallic debris formation (Control blade degradation behavior)

2.

Lower head of RPV failure by interaction between metallic debris and structural materials included
weldments (RPV failure behavior)

2. Outline、Mock-up Assembly Degradation test)
2.1 Control blade degradation behavior (CLADS-MADE test、We have done tests under several steam conditions using test pieces with a size of approximately
1.2m×0.1m×0.1m According to experimental result, control blade including B4C started melting at approximately
1200 degree C, but the melt did not completely melt up to around 1500 degree C. SEM/EDX and Raman
spectroscopy have shown that Cr-rich borides, which formed on the surface of control blade’s rods by reaction
between stainless steel of control blade and B4C of absorber, would protect evaporation of Boron.
、Experiment on Late In-vessel Severe Accident Phenomena)
2.2 RPV failure behavior (ELSA test、
RPV Lower region of BWR has complex structure, for instance a lot of pipes for control rod drive system andweldments between several parts. According to sever accident analysis, main three RPV failure modes have been
given as follows; Thermal damage, Mechanical damage, Eutectic reaction damage. We would like to explain
eutectic reaction, melting and relocation phenomenology. Assuming the conditions for the accident progress of
Unit 2, RPV lower head failure tests have been conducted under inert gas. According to in-site observation andpost analysis, solid metals of mixture of stainless steel and Zr alloy were melted and molten pool was formed
around 1000 degree C. After that, the molten material and weldment was reacted and melted at the temperature of
1100-1250 degree C, and the melts penetrated the pipe of control rod drive system.

36

J05

Advanced application method of optical fiber radiation sensor
towards the decommissioning of Fukushima Daiichi NPS
Yuta Terasaka1

Collaborative Laboratories for Advanced Decommissioning Science, Japan Atomic Energy Agency.

1

Abstract
We have developed the advanced application method of optical fiber towards the radiation distribution
measurement inside Fukushima Daiichi Nuclear Power Station (FDNPS). Position-sensitive optical fiber radiation
sensors based on the photon time-of-flight and wavelength resolving analysis were introduced, and the advantage
of both methods for the nuclear decommissioning application was discussed.
1. Introduction
For the decommissioning operation of the Fukushima Daiichi NPS, measurement of radiation distribution isindispensable for the radiation protection of workers. The optical fiber radiation sensor, which has a nature of highsensor flexibility and long-range measurement performance, is one of the promising sensors for radiation
distribution measurement for nuclear decommissioning. In this study, new application methods of optical fiber
radiation sensor were proposed, which has the characteristic of selective measurement of surface contamination
and wide dose rate range applicability.
2. Advanced application of optical fiber radiation sensor
2-1. Time-of-flight method
In this study, quartz and liquid core type optical fibers were newly applied to the conventional time-of-flightmethod (Fig. 1). The TOF method determines the incident position of radiation and optical fiber from the TOF
analysis of radiation-induced photons.
4XDUW]RSWLFDOILEHU
6WRSVLJQDO
Here, since the low-energy gamma rays
2SWLFDOILEHU)DVW
do not produce Cerenkov photons,
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307
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quartz and liquid core-type optical fibers
)DVW
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7$&
307
are useful for selective measurement of
DPS
90
90
Sr/ Y distribution on the contaminated
6WDUWVLJQDO
surface under the scattered-gamma-ray
dominant environment such as inside theFig. 1 Schematic of a TOF method.
FDNPS reactor buildings [1].
2-2. Wavelength-resolving method
Since the TOF method requires signal readout from both ends of an optical fiber, there may have some restrictions
on installation at FDNPS. Therefore, we have developed a "wavelength-resolving method" based on wavelength
analysis of the emitted light from the
6SHFWUXPXQIROGLQJ,QWHJUDWHG
optical fiber end, which enables the
ZDYHOHQJWKVSHFWUXP
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measurement of radiation distribution VFLQWLOODWLRQILEHU
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by the single-end readout. As shown in
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Fig. 2, this method is promising because
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it can be applied to a wide dose rate
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range from several tens of μSv/h to%DQGSDVVILOWHU
several Sv/h by optimizing the
Fig. 2 Schematic of a wavelength-resolving method.
photodetector [2,3].
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References[1] Y. Terasaka et al., Nucl. Instrum. Methods Phys. Res. A 784 (2021).
[2] Y. Terasaka et al., Ionizing Radiation, 47(3) (2022) (in Japanese).
[3] Y. Terasaka et al., Nucl. Instrum. Methods Phys. Res. A 1034 (2022).

37

J06

Overview of Radioactive Material Analysis and Research Facility
Laboratory-1
Yuki Sugaya, Hideyuki Akutsu, Hiroshi Shibata, Koji Ichitsubo
Japan Atomic Energy Agency (JAEA)

Abstract
The construction of a new facility named “Laboratory-1” at Radioactive Material Analysis and Research Facility(hereinafter referred to Laboratory-1) was completed in June 2022. Laboratory-1 has a key role in obtaining
analytical data needed for treatment and disposal of radioactive wastes generated from Fukushima Daiichi
Nuclear Power Station (1F) of TEPCO. This presentation provides an overview of Laboratory-1, including basic
designs, activities and operational plans.
1. Features of Laboratory-1
1-1. Object of the analysis
Laboratory-1 will receive low level and medium level radioactive wastes. The surface dose rates of them are
up to 1 mSv/h and 1 Sv/h, respectively. The objects of the analysis are mainly rubble which is scattered in 1F site,
incineration ash, secondary waste from water processing and ALPS treated water.
1-2. Equipment
Laboratory-1 has steel-shielded hot cells, glove boxes and fume hoods as shown in figure 1, which are used
for pretreating samples. The pretreated samples are analyzed by analytical instruments such as Ge semiconductor
detector, liquid scintillation counter, ICP-AES and ICP-MS.
1-3. Target analytes
Target analytes of rubble, incineration ash and secondary waste from water processing are radioactive
nuclides which are needed for understanding the characteristics and for safe disposal of radioactive wastes. In
addition, Laboratory-1 will obtain the analytical data of ALPS treated water as a third party.
Steel-shielded hot cells (4 units)

Fume hoods (56 units)

Laboratory-1

Three-story building (Height : 25 m)

Glove boxes (10 units)

Gross floor area : approx. 9,700 m2

Analytical instruments

Figure 1. Overview of Laboratory-1
2. Operational Plans
Laboratory-1 is currently operated without radioactive substances in preparation for maintenance andanalysis towards practical use. In September of this year, we will designate a radiation controlled area atLaboratory-1. After that, Laboratory-1 will begin the analysis with radioactive materials, including validation ofanalytical methods, research and development.

38

J07

Development of Exposure Reduction Technologies by Digitalizationof Environment and Radioactive Source Distribution-Current Status of Development Project-

Masahiro Suzuki, Takashi Yamaguchi, Masahiko Machida, Kuniaki Kawabata, Rintaro Ito,
Koji Okamoto
Japan Atomic Energy Agency

Abstract
In the decommissioning of the Fukushima Daiichi NPS (1F), prior to the full-scale implementation of fuel debris
removal operations, it is necessary to improve the on-site environment in order to safely and efficiently constructaccess routes in the reactor building (R/B). Then, JAEA is now developing a prototype system that estimatesradioactive source distributions by using on-site dose rate data and examines countermeasure effects of such as
decontamination, shielding with estimated contamination (radiation sources) on the estimated sources incyberspace. This report presents overview of the development project and its current status.
1. Introduction
In order to improve the environment at the R/B, it is important to identify the distribution of contamination basedon the on-site structural data and dose rate data, and to evaluate the worker’s exposure by decontaminating andshielding the contaminated radioactive sources.
To evaluate the effect of exposure reduction, therefore, JAEA is developing the prototype system to identify theintensity and location of high-dose radiation sources based on the structural data and spatial dose rate data in the
R/B, and to simulate the change of dose rate by decontamination of radiation sources and installation of shields incyberspace. In order to confirm the effectiveness of the system, analysis and evaluation are carrying out using thedata of JMTR, where the radiation sources (source locations) have already been known.
2. Overview of the development project
The development project is planned for two years since Jun. 2021, and is proceeding by three themes shown
below.
a. Development of technologies for inverse estimation of radiation source: (a) source inversion estimation, (b)spatial dose rate estimation, and (c) development of interactive visualization technologies.
b. Development of basic technologies for digitizing real environmental data: (a) development of technologies forbuilding an environment of data preparing, storing and utilizing for source inverse estimation, and (b) research anddevelopment of basic technologies for measuring real environmental data.
c. Development of technologies for field application: (a) verification tests for prototype, (b) research anddevelopment of data update technology based environmental change, and (c) basic study of connectivity andexpandability with other systems at 1F.
3. Main developments to dateThis section introduces the status of the implementation of the key technology in the project, that is the"Development of Technologies for Inversion Estimation of Radiation Source". The validity assessment of source
estimation is carried out by the following steps.
(1) The structural data of the site for the estimation is set (simultaneously, mesh model of the structure is created).
(2) Using the "Observation Point Indication Tool" developed by the project, the number of observation points
and locations of the points are calculated for the success of the source inverse estimation, and measurements
are carried out based on their data.
(3) Assuming that there is a source at each mesh, the radiation contribution from each mesh to the observation
point is calculated by PHITS (Monte Carlo transport calculation code), and from the radiation contribution(matrix) and the measured data, inverse estimation of the source based on LASSO regression is performed.
(4) The validity of the method is examined by comparing the calculated dose rate calculated from the source
distributions obtained by inverse estimation using PHITS with the measured data at the arbitrary observation
points.
Through successive operations from the inverse estimation to the calculations of air dose rate distributions withusing JMTR data, the key technology in the project can be evaluated.
Acknowledgements
This work was partially carried out in a subsidy program of “Project of Decommissioning and Contaminated
Water Management”, entitled “Development of Technologies for Work Environmental Improvement in Reactor
Building (Development of Exposure Reduction Technologies by Digitalization of Environment and Radioactive
Source Distribution)”.

39

F01

Overview of current CEA developments for characterization
techniques
Olivier Gueton1, Christophe Roure1 and Magali Saluden2

1CEA, DES, IRESNE, Cadarache, 13108 Saint-Paul-lez-Durance, France
2CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

Abstract
The prerequisite to any definition of waste management routes is to characterize physically,
chemically and radiologically the waste to be packaged. It involves taking samples, which can becomplex depending on the operational environment, and carrying out radiochemical and/or physicochemical analyses, or even specific characterizations if necessary. The strategy adopted aims toestablish standard spectra of waste and to carry out non-destructive/destructive nuclear
measurements, chemical analyses, etc.
Today the characterization technologies have reached a satisfactory level of maturity(detection/quantification limits, evaluation of uncertainties). However, further developments areneeded to meet the following two issues.
Issue 1. Making advanced technologies accessible to the nuclear industrialists: by conveying thecharacterization devices on site, making them easy to use (adapted software, training, etc.);
Active Neutron Interrogation (INA) is the most effective technology for the quantification of fissile
mass in radioactive packages, an inseparable objective of waste management optimization. X-ray
tomography is an imaging technique that allows non-destructive characterization of the contentsof radioactive waste containers to support knowledge of the internal constitution of thesecontainers.
The CEA TOMIS (in situ Tomography) (2017-2023) and ANAIS (in situ neutron activationinterrogation) (2022-2026) projects supported by the French government “Booster Economy
Initiative”, innovate the nuclear measurement field by developing transportable and adaptable
technologies. Thus, these technologies can be relocated as close as possible to the production orinterim-storage sites of waste: They authorizes the pooling of costs, increases the non-destructive
characterization rates without having to transport wastes over long distances to a dedicatedfacility.
The project structuration consists in the design and realization of a prototype, as well as radiation
protection studies to design the modular parts that can be deployed on site.
Each subset will be developed almost independently, the system being integrated in the final phaseof the project on a pilot site before being tested
TOMIS and ANAIS propose similar concepts based on complementary technologies. In the longer
term, a mobile measurement platform combining both systems (ANATOMIS) would be possible for
a complete characterization of the package contents.
Issue 2. Developing breakthrough characterization solutions for large volume packages (500/870
Liter) to measure and quantify the fissile mass and detect the presence of materials prohibited for
disposal;
To date, there is no non-destructive technique to probe the center of large packages (500/870 Liter).
To achieve this, the CEA started to invest in a 15 MeV accelerator and, now, proposes a dedicated
R&D program to quantify fissile material and detect the presence of materials prohibited for
disposal (toxic compounds, pyrophoric materials, etc.). This technology will not only meet the CEA
operational needs, but will also offer to all nuclear licensees who will have to manage their
historical packages a technological solution that does not exist today.

40

F02

Cement-based irradiation-resistant materials
David Chartier1, Stéphane Perrin1, Maxime Fournier1 and Magali
Saluden2
1CEA, DES, ISEC, DE2D, Univ. Montpellier, France

2CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

Abstract
The production of hydrogen from the irradiation effect of cementitious materials leads toconstraints for the conditioning of radioactive waste with conventional cements (Portland-type).
The objective of the MATRICE project, completed in 2021, was to allow for the incorporation ofsignificantly higher activities (typically a factor of 10), while maintaining acceptable radiolysis gas
releases with regard to current limitations (interim or final-storage environment) and evendecreasing them.
Three experimental tracks were assessed:- The minimization of the water content in conventional cements (silico-calcium cements) byadding specific compounds (superplasticizers/water reducers for example);- The use of “alternative” cements, compared to conventional cements (Portland), such assulpho-aluminous and phospho-magnesian cements; these cements have higher chemical
demands for water than conventional cements (thus less "radiolysable" residual water,which can produce hydrogen);- The implementation of processes, such as vibro-compaction, to facilitate the shaping of lowwater concrete.
The major MATRICE project results are:- First, the decrease of the H2 rate production by factor of 2 with the using of superplasticizers
compounds to reduce the amount of water;- Second, the decrease of the H2 production by factor of 3 to 10 by addition of NaNO 3 to thecement formulation with a low exothermicity of this formulation allowing its
implementation on large size packages;- Third, the decrease of the H2 production by factor of 2 with phospho-magnesium cement
formulation and satisfactory implementation of these formulations up to the 200 L
container-scale.
- Then, the implementation of vibro-compaction process is especially interessing for theconditioning of small porous solid waste hydrated with little water.
From these reference results for the formulation of cement-based irradiation-resistant materials,
the aim will be to develop H2 predictive modeling for cemented-waste packages.

41

F03

PACH3 (PACkage for H3-waste) projectDavid Lambertin1, Stéphane Perrin1, Etienne Tevissen2 and Magali
Saluden3
1CEA, DES, ISEC, Marcoule, 30207 Bagnols sur Cèze, France

2CEA, DES, IRESNE, Cadarache, 13108 Saint-Paul-lez-Durance, France
3CEA, DES, DDSD, Marcoule, 30207 Bagnols sur Cèze, France

Abstract
PACH3 (PACkage for H3-waste) project (2022-2026) aims to develop a low degassing package fortritiated wastes with no available disposal option. It is based on a two-pronged approach:The qualification of a composite cement matrix, consisting of a cement mortar (the natureof which is selected so as to minimize chemical interactions with the waste) and a mineral tritium
trap type -MnO2/Ag2O;
The development of a high sealing joint for metallic containers.
The technical challenges are (i) to demonstrate the irreversibility of the chemical trapping HT/T 2,(ii) the time evolution of the trap, (iii) the evolution of the properties of the metallic seal toguarantee a high sealing of the container in time.
The "major" milestones of the PACH3 project are the demonstration of the irreversibility of tritium
trapping, the durability studies of the materials and the realization of 50 L package mock-ups.
By developing low-degassing packages, this work will ultimately reduce the storage time of tritiatedwaste, thus reducing capital expenditure, operating costs, and reducing tritium releases to theenvironment.

42

F04

Remediation of contaminated soils in post-nuclear accident
situations: the DEMETERRES MOUSSE projectAnne Fornier1, Maxime Fournier2, Malvina Rennesson3, Marc Messalier4,
Jérôme Guillevic5, Christelle Latrille6, Sylvain Faure2, Pierre Chagvardieff7
CEA, DES, DDSD, DTPI, Saclay, France
CEA, DES, ISEC, Univ. Montpellier, Marcoule, France
3
Orano, DPS2D, Prisme, 125 avenue de Paris, 92320 Châtillon, France
4
Veolia Nuclear Solutions, Bagnols sur Cèze, France
5
IRSN, PSE-ENV/SEREN, 92262 Fontenay-aux-Roses, France
6
Université Paris-Saclay, CEA, SECR, 91191 Gif-sur-Yvette, France
7
Experconnect, avenue Franklin Roosevelt, 75008 Paris, France
1

2

Abstract
DEMETERRES MOUSSE (2022-25) aims to industrialize the Flotation Foam Technology. Such technology
enables to isolate the most contaminated soil particles, allowing waste volume reduction and the reuse of
decontaminated soil. The project gathers CEA (French Alternative Energies and Atomic Energy Commission),
Orano, Veolia, and IRSN (French public expert in nuclear and radiological risks). DEMETERRES MOUSSE has
been selected as a “main impact project” in the “Booster Economy Initiative” of the French government.
1. Presentation of the Flotation Foam Technology
The Flotation Foam Technology is a non-intrusive technology for contaminated soils remediation; it isselective for radionuclides attached to clay particles, mainly cesium. The use of Flotation Foam Technology
enables to reuse decontaminated soils while preserving their fertility and agricultural properties. The first tests
carried out in Japan in 2017 (Figure 1) with soils demonstrated by the Fukushima nuclear accident attested theproper operation of the technology in real conditions. The performances obtained were a reduction of waste volume
by a factor of 3.6 to 7.4 and a decontamination factor of 1.5 to 3.5. Laboratory-scale tests have shown that thesevalues can be significantly improved.

Figure 1: First tests of the Flotation Foam
Technology in Okuma in 2017 (©CEA/J.-L. Sida)

Figure 2: Principles of Flotation
Foam Technology

2. Goals of the DEMETERRES MOUSSE project
The industrial optimization of the Flotation Foam Technology remains to be achieved: this is thetechnological goal of the DEMETERRES MOUSSE project. Indeed, the capture of clay particles by foam bubbles(Figure 2) is a complex phenomenon governed by a succession of processes involving different types of physicochemical interactions which have to accommodate composition fluctuations related to the soils origin. Thereby,
DEMETERRES MOUSSE will implement a new pilot system: improvements will be made to enable optimal
screening and dispersion of soils at the process inlet. Moreover, an industrial sizing and design of the flotation
column will enable more refined particle selectivity. Finally, the efficiency of the separation between soils andwater at the bottom of the flotation column will be improved.
Experimental data from the DEMETERRES MOUSSE project will be used to improve models and Decision
Supporting Tools developed for the French authorities to react effectively in the event of a radiological accident.

43

F05

French facilities in support of Fukushima Daiichi
fuel debris cutting and retrieval R&D

Christophe Journeau1, Arthur Denoix1, Jules Delacroix1, Nicolas Minazzo1, Viviane Bouyer1, Ioana
Doyen2, Romain Garnier3, Emmanuel Porcheron4, Antonin Bouland4, Romain Berlemont5
1

CEA, DES, IRESNE, DTN, SMTA, Severe Accident Experimental Lab., Cadarache, 13108
Saint-Paul-lez-Durance, France
2
Université Paris-Saclay, CEA, Service d'Études Mécaniques et Thermiques, 91191 Gif-surYvette, France
3
CEA, DES, ISAC, DE2D/SEAD, Simulation and decommissioning techniques laboratory,
Marcoule, 30207 Bagnols-sur-Cèze, France
4
Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192 Gif-surYvette, France
5
ONET Technologies CN, 950 - 970 chemin des agriculteurs, 26700 Pierrelatte, France
A French consortium is being contracted by Japanese partners since 2014 to study several aspects of Fukushima
Daiichi fuel debris cutting and related aerosols. This R&D is based on a unique set of experimental facilities thathave been adapted to these issues: fabrication of prototypic samples (VULCANO), laser cutting (DELIA),
mechanical cutting (FUJISAN, CAPIMIF), thermal release of aerosols (VITI), dust mitigation (TOSQAN),
supervision and testing of robotic dismantling tasks (HERA). KROTOS facility could also be adapted for laser
and mechanical cutting in 1:3 scale pedestal geometry.

Mechanical cutting and aerosol
sampling in FUJISAN

44

Robotic arm supervisionin HERA

sampling in VITI

Laser cutting and aerosol
sampling in DELIA

Heating and aerosol

Fuel debris fabrication
in VULCANO

Layout of KROTOS for
UO2 sample cutting in
PCV mockup

of radioactive aerosols dispersion during laser cutting
F06 MitigatMitigation
ion of radioactive aerosols dispersion during laser cutting
1

1

1

2

2

Damien Roulet , Rémi Delalez , Thomas Da Silva , Emmanuel Porcheron , Thomas Gelain , Yohan
2
3
4
4
Leblois , Ioana Doyen , Christophe Journeau , Christophe Suteau
1

ONET, Engineering Business Unit, 26700, Pierrelatte, FranceInstitut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, Gif-Sur-Yvette, 91192, France
3
Université Paris-Saclay, CEA, Service d’Etudes Mécanique et Thermiques, 91191, Gif-Sur-Yvette, France
4
CEA, DES, IRESNE, DTN, SMTA, LEAG, Cadarache, F-13108 Saint-Paul-lez-Durance, France
2

This study is related to the demonstration of feasibility for the use of laser cutting techniques forthe fuel debris retrieval in the damaged reactors of Fukushima Daiichi. Besides laser cutting
techniques development, ONET Technologies has been leading a team with IRSN and CEA todevelop new technologies for dust mitigation with local collection and spray scrubbing systems ofaerosols generated during laser cutting and mechanical cutting operations.

Laser cutting in DELIA

Core boring on a
block of ex-vessel
fuel debris simulant

Spray in TOSQAN

45

F07

French Expertise and Technologies applied to
Fukushima Daiichi Fuel Debris Sorting

Damien ROULET1, Yvan LALLOT1, Thomas GRUNENBERGER1, Daphné OGAWA2,
Nicolas DEWYSE2, Pierre STROCK2, Benoit TASTET2, Frédérick CARREL3, Adrien
SARI3, Roberto DE STEFANO3, Aly ELAYEB3
1 ONET Technologies, Pierrelatte, France
2 ORANO

3 Université Paris-Saclay, CEA, List, F-91120 Palaiseau, France

One major challenge of Fukushima Daiichi decommissioning is the safe management of Fuel
Debris (FD) from retrieval to storage. Several types of waste will be generated during the large
scale retrieval operations and using the best management routes will have a decisive impact on
the final cost for storage and disposal. To overcome this challenge in the optimal way, a key stepis waste characterization and sorting. Several criteria are necessary, one of them being themeasurement of the fissile material contained in retrieved FD, which is deemed extremelycomplex due to the variability of the FD in its physicochemical and radiological content.
ONET, ORANO and CEA have started undertaking the above challenge, on the basis of existing
technologies and French expertise as well as operational feedback throughout the whole chain forthe management of nuclear waste from the retrieval to the final disposal.

FD simulant produced by CEA (left) and representation of
FD showing complexity (right)

Example of types of waste identified
inside the PCV on TEPCO picture© Orano© OranoOrano La Hague ACC facility: solid waste
canisters (CSD-C) are measured andcontrolled in a fully remote hot cell, prior tobeing processed (drying and compaction inthis case).

ECC is a fully remote and automated facilityin Orano La Hague where hundreds CSD-C
are temporally stored in a safe and organized
way, until final disposal site is available.

46

F08

Orano expertise to propose a comprehensive solution for Large-Scale
Retrieval and Interim Storage of Fuel Debris at 1F
Daphné Ogawa1, Nicolas Dewyse1, Vincent Janin1, Arnaud Rollet1, William Rocher1
1
Orano

Abstract
Orano 50-year experience for the design, operation & maintenance, decommissioning & waste management of highactive facilities related to nuclear fuel cycle is unequalled worldwide. Orano gathered experts from in engineering of
Hot Cells, robotics, decommissioning scenarios, reactor dismantling, nuclear measurement, nuclear packages, andproject management to propose a conceptual study for Large-Scale Retrieval and Interim Storage of Fuel Debris for
1F Unit 3.
1. Introduction
In 2020, TEPCO entrusted Orano to perform a study on a comprehensive solution to perform safe retrieval of Fuel
Debris (FD) from 1F Unit 3, from the retrieval of the FD located inside the PCV through to its interim storage onsite. The Fuel Debris Retrieval and Storage scenario was divided into 6 processes:● Developing access routes into the PCV● In-PCV processes to retrieve FD● In-cell processes to treat FD● On-site transport● Interim storage on-site● Dismantling and removal of FD retrieval facility and equipment at project end
2. Outline of the proposed scenario
2.1 Retrieval sequence(1) FD inside and outside pedestal are retrieved through a side-access while (1’) top-accessis prepared in parallel. (2) Removal of other internal equipment accessible from side-accessand at last (3) FD inside RPV upper part is removed by the top-access retrieval.
Retrieval sequence numbered

2.2 Main technological solutions proposed
SIDE ACCESSFD treatment cells are designed
based on Orano high activity
engineering principles to ensure safe
operation, maintainability and later
dismantling.

Retrieval is done through X6 and X1-B penetrations with robotic means.
STORAGENUHOMS® canisters and concrete modules
are licensed storage solutions for Spent Fueland Rad Waste, used to store TMI fueldebris. For 1F, the MATRIX two-stories
system is proposed for a reduced footprint.

TOP ACCESS
A new rotating platform is
installed on-top of the RPV.
Cutting and handling tools
are set on the platform.
3. Conclusion
The concept proposed by
Orano is: based on high
TRL solutions, costcompetitive, modular anddesigned to be operated andmaintained remotely.
The study was performed
under given conditions (ex.
in-air) with little knowledgeof the actual plantconditions. Further studies
are required to precise thebest scenario and associatedsolutions.

47

F09

Innovative solutions for Fuel Debris Retrieval (FDR)
New Lateral Opening for RPV access and cells conceptual designfor Large Scale FDR through PCV side access
Laurent David1, Daphné Ogawa1, Vincent Bessiron² and Pascal De Vito2
Orano, 2Framatome

1

Abstract
On the one hand, a lateral opening toward the Reactor Pressure Vessel is made possible by integrating water
jet cutting and demolition tools combined with a ferro-concrete debris suction system. On the other hand, cells
conceptual design for Large Scale FDR through PCV side access is completed. Both solutions are designed for 1F
needs based on French technology.
1. New Lateral Opening for RPV accessA new access path from the outside of the reactor building toward the inside of
the Reactor Pressure Vessel is performed by a tool head that combines several
technologies such as Abrasive Water Jet Cutting (AWJC), Water Jet Demolition (WJD)
and a ferro concrete debris suction system (see figure 1). The access allows internal
investigation and possibly sampling of Fuel Debris laying inside the RPV.

Fig 1. Framatome

The combination of the three technologies in one tool allows to avoid changing

combined tool

the tool each time the material to be cut changes (from concrete to steel or from steel
to concrete), to minimize the quantity of abrasive to inject (only water is used for
concrete demolition), to recover most contaminated debris, abrasive, and water, fortreatment and recycling. This tool head is positioned at the end of a mechanical system
to guide the tool inside the reactor building (see figure 2).
Fig 2. Orano-Framatome facility

Tests were conducted to confirm the feasibility of the system.
By optimizing the functions, a miniaturization of the system could beachieved, making the site implementation easier.
2. Cells conceptual design for Large Scale FDR through PCV
side
The conceptual design includes cells (see figure 3) ensuring
Full Scale FDR robot arm access to PCV, maintenance of the robot
arm, extraction, sorting and packing of and Fuel Debris in cans andcanisters and of Low-Level Waste in dedicated containers.
Contamination control, decontamination systems, maintenancesolutions including emergency situations are anticipated, based on

Fig 3. Fuel debris retrieval facilities overview

proven solutions with operating experience feedback.
Conclusion
Orano and Framatome join forces to develop solutions adapted to 1F situation based on their experience inrobotics, remote operations, hot cell design and waste management.
Those studies were performed in the frame of IRID “Development of Technology for Gradually Increasing Retrieval Scaleof Fuel Debris” and “Development of Technology for Investigation inside Reactor Pressure Vessel (RPV)”

48

F10

DEM&MELT In-Can Process for Fukushima Daiichi Nuclear Power
Station Water Treatment Secondary Waste
Régis Didierlaurent1, Daphné Ogawa1, Thierry Prevost1, Laurent David2,
Maxime Fournier3, Caroline Michel3, Jean-François Hollebecque3, Kohei Shibata4

Orano Recyclage, Le Prisme, 125 avenue de Paris, 92320 Châtillon, France
Orano Projets, Place de Wicklow, Montigny le Bretonneux – 78182 Saint Quentin Yvelines, France
3
CEA, ISEC, DES, DE2D, Univ. Montpellier, Marcoule, France
4
ANADEC, 4-11-3 Shiba, Minato-ku, Tokyo 108-0014 Japan
1

2

Abstract
The DEM&MELT In-Can vitrification process is a robust, simple and versatile in-situ vitrification process.
It is designed for intermediate and high-level waste and is compact enough to be implemented in an existing facility
or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of
secondary waste and to minimize investment and operating costs. The applicability of the DEM&MELT
technology to water treatment secondary waste arisen at Fukushima Daiichi Nuclear Power Station (1F waste) was
studied and tested. High rates of waste incorporation and good performances of the wasteform especially withregards to its durability were verified.
1. Introduction
The DEM&MELT In-Can vitrification technology is considered as a mature solution that could be applied
for the treatment and conditioning of waste resulting from the D&D and remediation works in Fukushima Daiichi
Nuclear Power Station. As a result, the DEM&MELT technology was tested From FY2018 to FY2021 on a wide
range of 1F waste in the frame of the R&D funded by the Japanese Ministry of Economy, Trade and Industry as
The Subsidy Program “Project of Decommissioning and Contaminated Water Management”.
2. Evaluation of the In-Can vitrification technology for 1F waste – Main results
Orano and CEA contributed to the successful achievement of the following
results to demonstrate the applicability of the thermal treatment throughDEM&MELT:- High waste loadings were achieved (up to 80 wt.% oxide) for a wide
range of waste (zeolites, CST and ALPS slurries), allowing significant
volume reduction.
- Very low radionuclides volatilities were obtained during the treatment.
- Demonstrations of the capabilities to treat 1F waste in mixture or oneby-one were performed.
- Technological and process demonstrations were performed at pilot scale
and at full-scale, with challenging waste (Figure 1).
- The wasteforms produced allow long-term immobilization of theradionuclides. The material performances such as chemical durability
have been assessed and are very promising.
- Two compact facilities were conceptually designed to process all thewaste from ILW to HLW. A safety study revealed the prospect that the
DEM&MELT process can meet the safety requirements of the Japanese
regulations.

Figure 1: Vitrified
ALPS slurries(270 kg at 42% WLox)

3. Conclusion
The tests carried out confirmed the process ability to treat a wide variety of waste with easy and reliable
operation. The technology makes it possible to achieve high rates of waste incorporation into the matrix while
guaranteeing good performances of the wasteform — especially with regards to its durability, in order to limit therelease of radionuclides into the biosphere — and ensures very low volatility of radionuclides during thermal
treatment.

1
49

F11

Chameleon and Anemone tools - Innovative gripping
technologies capable of sampling and recovering fuel debris
D. OGAWA, K. LE FLANCHEC, V. TOULEMONDE, A. COUDRAY
ORANO

Abstract
Orano develops various tools for remote investigation and sampling, designed to be remotely operated andmaintained in harsh environment. Since 2018, Orano is manufacturing two types of end-tool dedicated to thesampling and recovery of fuel debris from Fukushima-Daiichi unit 2 RPV.
 The Chameleon Tool, called as Flexible Gripper Type Sand-like debris sampling Tool, at the industrial
prototype stage is ready to undergo combination tests with the robot arm.
 The Anemone Tool, which detail design is conducted during JFY2022 with the aim to manufacture the final
tool by 2024.
The end-tool are bio-inspired, meaning they are inspired by nature/animal life.
1. Chameleon Sampling Tool

Figure 2 - Chameleon Sampling Tool
parts

The Chameleon Sampling Tool (or CST) has beendeveloped to catch small pieces of Fuel Debris, up to
100g. The system is based on a silicone membrane
which enwrap the target piece by vacuum suction. The
name “Chameleon” was given as the grasping
principle was inspired from the capability of the
Figure 1 animal’s tongue to catch its target. Elementary tests Chameleon head
were performed on active corium simulants in 2019,and since that Orano has been developing an industrial prototype. The CST
consists of a sampling head (see figure 1) and a sliding table (see figure 2)
allowing accurate positioning of the tool from the sampling target, tocompensate the robot arm positioning uncertainties. The CST prototype
will soon be tested in Europe in inactive conditions in combination withthe robot arm, before being transferred to Japan for additional testing.

2. Anemone Retrieval Tool
The second gripping tool is the Anemone Retrieval Tool (ART) and is inspired
from the sea anemone behavior. After the investigations conducted inside unit 2
PCV, it appeared that the size of the Fuel Debris pebbles to be collected are bigger
than planned. Orano suggested the use of an alternative gripping technology, withhigher gripping force. The Anemone solution has been developed for nuclear
application in France. It has successfully collected graphite samples from a legacy
waste pit in April 2022. For its application at Fukushima, tests
were performed in 2021 with samples up to 1kg in air andunderwater. The good results confirmed that for bigger Fuel
Debris pieces, the Anemone head combined with the sliding
table for positioning would be a good solution, and Orano is
Figure 3 - QR code now conducting the design and manufacturing of the tool forto Anemone video a future use on-site.

Figure 4 - Anemone tool
retraction principle

Conclusion
Both gripping tools designed by Orano are bio-inspired, thus easy to be deployed on-site in humid or dry condition,even with poor visibility thanks to its dedicated positioning system. The heads are pneumatic-actioned thus lesssensitive to radiation than other electro-mechanical tools.
This achievement includes the results of research and development which MHI, as a member of IRID (International Research Institute for Nuclear
Decommissioning), has had implemented by the subsidy of the projects of METI (Ministry of Economy, Trade and Industry) on the Decommissioning andContaminated Water Management.

50

F12

Tooling
Tooling Solutions
Solutions forfor Nuclear
Nuclear Dismantling
Dismantling Projects
Projects

Heiko Herbell1, Holger Damies1, Frank Kostroun1, Christoph Stiepani1, Norio Terada2, Vincent
Bessiron2
Framatome GmbH, 2Framatome Japan KK

1Abstract
Framatome proposes high technology products and services for the D&D of reactors worldwide and is alreadyinvolved in the Fukushima Dai Ichi project with engineering activities for Fuel Debris Retrieval.
Additionally, decommissioning of nuclear sites demands dedicated tooling solutions ensuring reliabledismantling operations fully complying to all safety requirements. Application of specific manipulators or
remotely-operated tools is mandatory especially in high-radiation areas. Beside segmentation technologies likemechanical cutting the D&D tooling also covers decontamination or water filtration systems, handling devices,support structures or dedicated manipulators. The poster provides an overview on tooling concepts applicable atthe Fukushima plant.

Tooling Solutions related to Plant Decommissioning Strategy
Many nuclear dismantling projects are following a decommissioning strategy from hot to cold or from inside to
the outside which enables an early removal of the radiological inventory. The overall dismantling sequencebasically consists of:- Full System Decontamination- Radiological evaluation of the systems using Sampling and/or inline radiological measurements- Segmentation and packing of the RPV internals pieces into waste containers- Segmentation and packing of the RPV pieces into waste containers- Removal of other large components (e.g., steam generator, pressurizer, …)Decontamination prior to decommissioning can significantly reduce volume of radioactive waste, minimize
personnel exposure to radiation and thereby improve safety. Full System Decontamination (FSD) is an efficientand reliable chemical decontamination process of primary cooling circuit and main auxiliary systems. The
technology is easily adaptable to any NPP to- Minimize activity inventory- Generate metallically clean surfaces- Optimize gamma/alpha ratio- Reduce generation of waste
The process has been extensively tested and features a high degree of material compatibility.
For the segmentation of large components mechanical cutting techniques (e.g. sawing, nibbling) are preferredsolutions from different reasons. Besides being reliable and allowing easy maintenance they are generating few
aerosols, hydrosols and are therefore not linked to specific heat or secondary waste constraints.
Typically, a mechanical segmentation tool set is containing- Sawing solutions including disc saws, band saws as well as complete sawing racks and cutting stations- Handling tools like various grippers including connecting rods, specific couplings and hoisting devices /traverses- Special tooling as nozzle cutter, drilling or milling devices also for contingency operations- Auxiliary tooling as water filtration systems, shielding covers, cameras
In terms of efficient dismantling execution, a reliable sequence planning is needed based on the radiologicalconditions but also the specific state of the nuclear site and the main components. As an example robotics andremotely-operated tools like in-pipe manipulators can be applied for visual inspections of hardly accessible highdose areas providing valuable information on plant conditions.
There are several such technologies that have been already applied in recent projects with a proven performanceand continuous improvement. It can be adapted to be applicable at the Fukushima plant. There are also new
solutions under development addressing future D&D needs.

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