Mon, August 6, 2018
Alios Iwaki Performing Arts Center in Iwaki-city,
Fukushima-prefecture, Japan

Nuclear Damage Compensation andDecommissioning Facilitation Corporation (NDF)

This document is published by Nuclear Damage Compensation and Decommissioning
Facilitation Corporation (NDF). Inquiries about each abstract should be addressed to
ndf-poster@ndf.go.jp by e-mail.

Cover photos: Unit 3 of the Fukushima Daiichi nuclear power station in Japan as at 24 August
2011 and 21 February 2018 (TEPCO, Japan).

Contents
Session A: Plan and Activities related to Decommissioning ・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 1
A01

Overview of Research & Development for Decommissioning of the Fukushima
Daiichi Nuclear Power Station
Setsuo Nakashima, Takuto Miyamoto , Hidenori Nomura and
Takuro Ueda
Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF)

A02

Activities of Review Committee on Decommissioning of Fukushima Daiichi
Nuclear Power Station in Atomic Energy Society of Japan
Noriko Asanuma1, Michitaka Sasoh2 and Hiroshi Miyano3
1: Tokai University, 2: Toshiba Corporation, 3: Hosei University

A03

Latest Activity of AESJ Research Committee on Fission Product Behavior under
Severe Accident
Yosuke Katsumura1, Shunsuke Uchida2, Masahiko Osaka2,
Hidetoshi Karasawa3, Koichi Nakamura4 and Junichi Takagi5
1: Japan Radioisotope Association, 2: Japan Atomic Energy Agency (JAEA),
3: Institute of Applied Energy (IAE),
4: Central Research Institute of Electric Power Industry (CRIEPI),
5: Toshiba Energy Systems & Solutions Corporation

A04

Relationship between the Research and Development of the Project ofDecommissioning and Contaminated Water Management and Connection toPreliminary Engineering
Masaaki Matsumoto, Naoki Kondo and Keishun Nakamura
Mitsubishi Research Institute, Inc.

A05

Advanced Research and Education Program for Nuclear Decommissioning
Toru Obara1, Kenji Takeshita1, Yukitaka Kato1, Hiroshi Akatsuka1,
Hiroshige Kikura1, Takehiko Tsukahara1, Katsumi Yoshita1,
Koichiro Takao1, Jun Nishiyama1, Masayuki Harada1, Nobuyuki Iwatsuki1,
Koich Suzumori1, Gen Endo1, Kenji Kawashima2, Noriko Asanuma3,
Tsuyoshi Arai4 and Naoyuki Takaki5
1: Tokyo Institute of Technology, 2: Tokyo Medical and Dental University,
3: Tokai University, 4: Shibaura Institute of Technology,
5: Tokyo City University

A06

Contribution to Fukushima Daiichi unit 5 and 6 Strategic Decommissioning
Utilizing Experience of Decommissioning in Japan and Oversea
Toyoaki Yamauchi1 and Colin. R. Austin2
1: The Japan Atomic Power Company, 2: Energy solutionsiSession B: Remote Technology and Radiation Tolerant Equipment ・・・・・・・・・・・・・・・・・・・・・ 7
B01

The Fukushima Consortium of Robotics Research for Decommissioning andDisaster Response
Yoshiro Owadano
Fukushima Consortium of Robotics Research for Decommissioning andDisaster Response

B02

Robots Developed by the Members of Fukushima Consortium of Robotics
Research for Decommissioning and Disaster Response
Yoshiro Owadano
Fukushima Consortium of Robotics Research for Decommissioning andDisaster Response

B03

Organization Profile of IRID
Naoaki Okuzumi
International Research Institute for Nuclear Decommissioning (IRID)

B04

Overview of IRID R&D Projects
Naoaki Okuzumi
International Research Institute for Nuclear Decommissioning (IRID)

B05

The Future of Remote Access for 1F Decommissioning (VNS Integrated Nuclear
Remote Access)Akira Ikeda1, Mamoru Numata1, Matt Cole2, Scott Martin2 and Marc Rood2
1: Kurion Japan K.K, 2: Veolia Nuclear Solutions, Inc.

B06

Development of a Remote Concrete Sampling Device
Kiyoaki Kato, Akira Horii and Toshihiko Mizushima
ATOX

Co.,

Ltd.

/International

Research

InstituteforNuclearDecommissioning (IRID)
B07

Usage of Remote Technologies for D&D Operations in Hostile Environment/
Orano’s Experience Serving Fukushima-Daiichi
D. Ogawa, Y. Kikkawa, N. Nakamura and N. Breton
Orano Atox D&D Solutions Co., Ltd. (ANADEC)

B08

Remote Control of a Sensor-less Manipulator for Decommissioning Work
Takahiro Nagai, Masaki Odai, Kishiko Maruyama and Ryosuke Kobayashi
Hitachi, Ltd.

B09

Introduction of Mirion Technologies’s Radiation Tolerant Camera the World’s
Highest Radiation Tolerant Performance Camera
Tadashi Sato, Mikio Katsura and Yuto Tajima
CORNES Technology Limited

B10High Radiation Resistant Fiberscope using Silica based Imagefiber
Tomoaki Toriya, Shinya Taniwaki and Manabu Kudoii

Fujikura Ltd.
B11

High Radiation Shielding Rubber Chips
Kenji Kitazume, Shinji Shimada and Hiroshi Kouda
Toyo Rubber Chip Co, Ltd.

B12

Radiation-Resistant Flexible PVC Material
Makoto Oi, Hidenori Shima and Ryosuke Ogata
Tigers Polymer Corporation

B13

Radiation Shielding Clothing to Facilitate Decommissioning of Nuclear Reactor
Nobuhiro Shiraishi, Koji Suzuki, Yosuke Tanji
Fukushima Midori Anzen Inc.

Session C: Analysis, Measurement and Survey Technologies ・・・・・・・・・・・・・・・・・・・・・・・・・ 20
C01

Development of a Fiber Optic Radiation Monitor Used in Severe Environments
Takahiro Tadokoro1, Shuichi Hatakeyama1, Katsunori Ueno1,
Yuichiro Ueno1, Keisuke Sasaki1, Yoshinobu Sakakibara1,
Toru Shibutani1, Takahiro Ito1, Mikio Koyoma1 and Koji Nebashi2
1: Hitachi Ltd., 2: Hitachi-GE Nuclear Energy, Ltd.

C02

Progress of 1F PCV contaminant analysis by TEM observation
Akihiro Suzuki1, Yuji Kitsunai1 and Ikken Sato2,3
1: Nippon Nuclear Fuel Development Co., Ltd. (NFD), 2: Japan Atomic
Energy Agency (JAEA), 3: International Research Institute for Nuclear
Decommissioning (IRID)

C03

Experimental Study on Neutron Induced Gamma Ray Spectroscopy for
Characterization of Fuel Debris
Yasushi Nauchi1, Tadafumi Sano2, Hirokazu Ohta1, Shunsuke Sato1,
Motomu Suzuki1 and Hironobu Unesaki2
1: Central Research Institute of Electric Power Industry (CRIEPI),
2: Kyoto University

C04

NDA system for Fuel Debris Characterization at Fukushima Dai-ichi
Ryo Ichimiya, Atsuo Suzuki, Mitsuo Takeuchi and Helene Lefebvre
Mirion Technologies (Canberra) K. K.

C05

Application of MTKK Technology for D&D and Fukushima Remediation
Atsuo Suzuki, Ryo Ichimiya, Mitsuo Takeuchi and Helene Lefebvre
Mirion Technologies (Canberra) K. K.

C06

Gamma-ray

Measurements

inHigh-dose

Radiation

Fields

using

TlBr

Semiconductor Detectors
Mitsuhiro Nogami1, Keitaro Hitomi1,Tatsuo Torii2, Yuki Sato2,
Yoshihiko Tanimura2, Kuniaki Kawabata2, Kenichi Watanabe3,
Toshiyuki Onodera4, Nobumichi Nagano1, Seong-Yun Kim1, Tatsuya Ito1,
iii

Keizo Ishii1 and Hiroyuki Takahashi5
1: Tohoku University, 2: Japan Atomic Energy Agency (JAEA),
3: Nagoya University, 4: Tohoku Institute of Technology,
5: The University of Tokyo
C07

Piping System, Risk Management based on Wall Thinning Monitoring andPrediction - PYRAMID
Toshiyuki Takagi1, Tetsuya Uchimoto1, Yutaka Watanabe1, Hiroshi Abe1,
Shinji Ebara1, Atsushi Iwasaki2, Ryo Morita3, Shun Watanabe3,
Ryoichi Urayama1, Hongjun Sun1, Philippe Guy4, Christophe Reboud5,
Bernard Normand6, Nicolas Mary7
1: Touhoku University, 2: Gunma University,
3: Central Research Institute of Electric Power Industry (CRIEPI),
4: National Institute of Applied Sciences (INSA) – LVA,
5: French Alternative Energies and Atomic Energy Commission (CEA),
6: National Institute of Applied Sciences (INSA) – MATEIS, 7: ELyTMax

C08

3D Reconstruction of Unit 3 Primary Containment Vessel Interiors at Fukushima
Daiichi Nuclear Power Station Using Structure from Motion
Xiaorui Qiao, Yonghoon Ji, Atsushi Yamashita and Hajime Asama
The University of Tokyo

C09

Research on Numerical Simulation of MCCI using MPS method
Keigo Fujikawa, Koji Okamoto and Shunichi Suzuki
The University of Tokyo

C10

Structural Deformation Monitoring for Safe Dismantling of Nuclear Facilities
Charles Viano1, Masuo Kado2 and Masao Tajima2
1: OSMOS Group SA., 2: OSMOS Association

Session D: Decontamination Technology and Treatment for Contaminated Water ・・・・・・・・ 30
D01

Development of New Radioisotopes Selective Adsorption Media in EBARA
Takashi Sakuma, Makoto Komatsu, Tatsuya Deguchi,
Kaoru Kikuchi, and Takeshi Izumi
EBARA Corporation

D02

Development of a Treatment System for Radioactive Spent Ion Exchange Resins
Takashi Kawakami, Takaeshi Izumi, Ishiyama Yuji and Hirofumi Inagawa
EBARA Corporation

D03

Inhibition of Oxidative Decomposition of Ion-exchange Resins
Makoto Komatsu, Tatsuya Deguchi and Takeshi Izumi
EBARA Corporation

D04

Alpha nuclide absorbent, TANNIX
Kazuhiko Hamaguchi and Hideyuki Teshimaiv

Mitsubishi Nuclear Fuel Co., Ltd. (MNF)
D05

Study on Application of Laser-Decontamination in Decommissioning
Masato Watanabe1, Hiromitsu Inagaki1, Kazuhisa Fujita2,
Shin-icihro Okihara2, Kazuaki Toyosawa3 and Nobumitsu Maebashi3
1: CHUBU Electric Power Co., Inc.,
2: The Graduate School for the Creation of New Photonics Industries,
3: Toyokoh Co., Ltd.

D06

Sorting Technology for Efficient and Safe Soil Processing
Steve Rima1, Jeffrey Lively1, Masaru Noda2 and Yoshiharu Hashizume2
1: Wood, 2: Obayashi Corporation

D07

Assessment of Future Countermeasures against Contaminated Groundwater inFukushima Daiichi Nuclear Power Station
Tadafumi Koyama1, Yasuharu Tanaka1, Daisuke Tsumune1, Takatoshi
Hijikata1, Mamoru Kamoshida2, Takeki Yamauchi2, Masaki Imazu2 andToshihiko Fukuda2
1: Central Research Institute of Electric Power Industry (CRIEPI),
2: Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF)

Session E:
E01

Progress of Decommissioning at the Fukushima Daiichi NPS ・・・・・・・・・・・・・ 37

Construction of Unit-3 Cover Structure for Fuel Removal
Shinya Okada, Kenichiro Go, Norio Ryoki, Taimei Hida, Miho Miyazaki,
Ippei Matsuo and Kihei Ogawa
Kajima Corporation

E02

Land-side Impermeable “Frozen-Soil” Wall
Atsuhiro Fukada, Isao Abe, Taichi Esaki, Shoji Seno, Teru Yoshida andTorahito Suzuki
Kajima Corporation

E03

Technologies for Dismantling Radioactive Contaminated Water Storage Tanks at
Fukushima Daiichi Nuclear Power Station
Kazushige Kato1, Masahiro Nakamura1, Haruo Nagamine1,
Naoki Maeshiro2, Hiroshi Nakamura3 and Yoshihito Nakahara4
1: TAISEI CORPORATION,
2: Tokyo Electric Power Company Holdings, Inc. (TEPCO),
3: TOYO UNION Company Limited, 4: TAIYO KOGYO CORPORATION

E04

Dismantling Plan of Exhaust Stack for Units 1 and 2
Hiroshi Noda, Yoshihisa Michizuka and Takashi Okutsu
Tokyo Electric Power Company Holdings, Inc. (TEPCO)

E05

Improvement in work environment at Fukushima Daiichi NPSv

Yuta Suzuki, Toshiharu Matsuzawa, Shiro Takahira, Koji Hayashi
Tokyo Electric Power Company Holdings, Inc. (TEPCO)
Session H: Hydrogen Treatment ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 42
H01

Present Status on R&D of Hydrogen Concentration Reduction Technology in
Waste Storage Containers with PAR
Kazuyuki Takase1, Hisayuki Suematsu1, Hitoshi Sugiyama2,
Hirohisa Tanaka3, Masashi Taniguchi4, Isamu Kudo5, Ryutaro Hino6,
Atsushi Terada6, Daijyu Muramatsua6 and Yasuo Nishihata6
1: Nagaoka University of Technology, 2: Utsunomiya University,
3: Kwansei Gakuin University, 4: DAIHATSU Co., Ltd.,
5: ADVAN ENG. Co., Ltd., 6: Japan Atomic Energy Agency (JAEA)

H02

Evaluation of Performance on Hydrogen Concentration Reduction in Storage
Containers with PAR Wetted by Water
Kazuyuki Takase1, Matsmoto Yoshinobu1, Isamu Kudo2, Shogo Taira2,
Keisuke Tanizawa2, Gaku Takase1 and Yoshihisa Hiraki1
1: Nagaoka Univ. of Technology, 2: ADVAN ENG. Co., Ltd.

H03

Hydrogen safety technology utilizing the automotive catalyst
T. Kita1, H. Ono1, K. Takenaka1, S. Kishimoto1, M. Taniguchi2,
D. Matsumura3, Y. Nishihata3, R. Hino3, E-A. Reinecke4, K. Takase5 and
H. Tanaka1
1: Kwansei Gakuin University, 2: Daihatsu Motor Co., Ltd.,
3: Japan Atomic Energy Agency (JAEA),
4: Forschungszentrum Jülich GmbH, 5: Nagaoka University of Technology

Session K: Radioactive Waste Management ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 45
K01

Challenge for the Decommissioning of Fukushima based on the Developed
Technologies
Susumu Tosha, Shohei Sato, Masaki Jimba and Toshiki Fukui
IHI Corporation

K02

Basic Research Programs of Vitrification Technology for Waste Volume Reduction(1) Development of Vitrification Technology for Low Level Radioactive Waste(LLW)
Yasutomo Tajiri, Oniki Toshiro Toshiaki Kakihara andToyonobu Nabemoto
IHI Corporation

K03

Study on Vitrification Technology for Radioactive Waste at Fukushima Daiich
NPS
Toshiro Oniki1, Yasutomo Tajiri1, Toyonobu Nabemoto1, Toshiki Fukui1,
viSeok-Ju Hwang2 and Cheon-Woo Kim2
1: IHI Corporation, 2: Korea Hydro & Nuclear Power Co., Ltd. (KHNP)
K04

Basic Research Programs of Vitrification Technology for Waste Volume Reduction(2) Development of high waste loading glasses
Yoshiyuki Miura1, Tetsuji Yano2, Jun Matsuoka3, Toru Sugawara4 andAtsunobu Masuno5
1: Japan Nuclear Fuel Limited (JNFL), 2: Tokyo Institute of Technology,
3: The University of Shiga Prefecture, 4: Akita University,
5: Hirosaki University

K05

Volume Reduction and Recycling of Contaminated Metal by Melting
Satoshi Karigome1, Kentaro Shibata1, Yasuhiro Miyagoshi2 andNorifumi Onogi2
1: The Japan Atomic Power Company (JAPC),
2: JFE Engineering Corporation (JFEE)

K06

SIAL® : Geopolymer Solidification Technology Approved by Slovak/ Czech Nuclear
Authority
Milena Prazska1 and Marcela Blazsekova1, Hisashi Mikami2,
Nobuyuki Sekine2 and Mika Mochimaru2
1: Wood (Slovakia), 2: Fuji Electric Co., Ltd.

K07

Contribution to Environmental Remediation in Fukushima (2)Takao Ikeda, Yasuhiro Suzuki and Takahiro Kikuchi
JGC Corporation

K08

Radiological Separation of Contaminated Materials -experiences with a highperformance mobile belt conveyor systemF. Langer and M. Sokcic-Kostic
NUKEM Technologies Engineering Services GmbH

K09

Radioactive Waste Management - treatment, storage and disposal of LILW and
HLW K. Büttner and R. Slametschka
NUKEM Technologies Engineering Services GmbH

K10

Tailor-made solutions with proven Technology to Waste Management
Motoki Nishibori
KOBELCO STUDSVIK Co., Ltd.

Session J: Research and Development of the JAEA ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 55
J01

Current Status of the JAEA’s Fukushima R&D Bases for the Decommissioning of
the Fukushima Daiichi Nuclear Power Station
Nobuhiro Shobu, Makoto Ooka, Toshiro Nakai and Shinichi Nakayama
Japan Atomic Energy Agency (JAEA)
vii

J02

Operation Training System for Decommissioning Work Using Immersive Virtual
Reality Technology
Yoshihiro Tsuchida1, Takahiro Ohno1, Nobuhisa Nosaki1 andMitsugu KatoJapan Atomic Energy Agency (JAEA)

J03

Outline of JAEA Okuma Analysis and Research Center
Akihiro Ishimi, Toshihiko Inoue, Ichiro Kosaka, Shinichi Koyama andTomozo Koyama
Japan Atomic Energy Agency (JAEA)

J04

Plasma Heating Test on Fuel Assembly Degradation
Takuya Yamashita1,2, Ikken Sato1,2
1: Japan Atomic Energy Agency (JAEA),
2: International Research Institute for Nuclear Decommissioning (IRID)

J05

Characterization of the Large-scale MCCI Test Products for Fuel Debris Removalfrom the Fukushima Daiichi Nuclear Power Plant
Hirotomo Ikeuchi1, Hiroyasu Hirano1, Tadahiro Washiya1,
Pascal Piluso2, Laurent Brissonneau2
1: Japan Atomic Energy Agency (JAEA)/ International Research Institutefor Nuclear Decommissioning (IRID), 2: French Alternative Energies andAtomic Energy Commission (CEA)

J06

Advanced Study on Remote and In-situ Elemental Analysis for Molten Fuel
Debris in Damaged Core by Innovative Optical Spectroscopy
Ikuo Wakaida1, Hironori Ohba1,2, Yuji Ikeda3, Tetsuo Sakka4 andTakunori Taira5
1: Japan Atomic Energy Agency (JAEA), 2: National Institutes for
Quantum and Radiological Science and Technology (QST),
3: Imagineeing Inc., 4: Kyoto University, 5: National Institutes of Natural
Sciences (NINS)

J07

Application of ICP-MS to Analysis of Samples from 1F site at Radioactive Material
Analysis and Research Facility
Van-Khoai Do1,2, Takuma Horita1,2, Katsue Miura1,2 and Maho Iwasaki1,2
1: Japan Atomic Energy Agency (JAEA),
2: International Research Institute of Nuclear Decommissioning (IRID)

J08

Estimation of Radioactive Nuclide Distribution at NPS site Developing Model
based on Analytical Data of On- and Off-site
Kazuki Iijima1, Yoshikazu Koma1, Tadafumi Niizato1, Terumi Dohi1,
Yoshihito Ohmura2, Takayuki Sasaki3 and Kenji Nanba4
1: Japan Atomic Energy Agency (JAEA),
2: National Museum of Nature and Science, 3: Kyoto University,viii

4: Fukushima University
J09

Development of Corrosion Database under Radiation Environment
Tomonori Sato1, Kuniki Hata1, Yoshiyuki Kaji1, Hiroyuki Inoue2,
Kousei Taguchi3, Hajime Seito3, Eiji Tada4, Hiroshi Abe5, Eiji Akiyama5
and Shunichi Suzuki6
1: Japan Atomic Energy Agency (JAEA), 2: Osaka Prefecture University,
3: National Institutes for Quantum and Radiological Science andTechnology (QST), 4: Tokyo Institute of Technology, 5: Tohoku University,
6: The University of Tokyo

J10

A Prediction of the Dose Rate Distribution in the Primary Containment Vessel ofthe Fukushima Daiichi Nuclear Power Station
Keisuke Okumura and Eka Sapta Riyana
Japan Atomic Energy Agency (JAEA)

J11

Study on Wet Waste Storage - Irradiation Behavior of Carbonate Slurry and Wet
ZeoliteIsao Yamagishi and Takafumi Motooka
Japan Atomic Energy Agency (JAEA)

J12Visualization of Radioactive Substances by Integrating Radiation Measurement,
Remote Technology, and 3-D Optical Measurement
Yuki Sato, Yuta Terasaka, Masaaki Kaburagi, Yuta Tanifuji andTatuso Torii
Japan Atomic Energy Agency (JAEA)

J13Development of simulation function of remotely operated vehicle for underwaterreconnaissance tasks
Kenta Suzuki, Shunji Mukaida and Kuniaki Kawabata
Japan Atomic Energy Agency (JAEA)

J14Information Gathering of Working Environment based on 3D Reconstruction from
Plural Images for Decommissioning
Toshihide Hanari and Kuniaki Kawabata
Japan Atomic Energy Agency (JAEA)

J15

Sensor Database for Research of SLAM in Reactor Buildings
Taichi Yamada and Kuniaki Kawabata
Japan Atomic Energy Agency (JAEA)

J16

3-D Sensing Technologies of Radioactive Substances and Environmental
Structure inside the Fukushima Daiichi Nuclear Power Station
Yuta Tanifuji, Yuki Sato, Yuta Terasaka, Hiroshi Usami,
Masaaki Kaburagi, Kuniaki Kawabata, and Tatsuo Torii
Japan Atomic Energy Agency (JAEA)
ix

Session ON: OECD/NEA ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 71
ON01

NEA’s role in addressing International Challenges in Radioactive Waste
Management
Ichiro Otsuka and Gloria Kwong
Organisation for Economic Co-operation and Development/Nuclear Energy
Agency (OECD/NEA)

ON02

The NEA Committee on Decommissioning of Nuclear Installations and Legacy
Management (CDLM): a new platform for international cooperation
Ichiro Otsuka and Gloria Kwong
Organisation for Economic Co-operation and Development/Nuclear Energy
Agency (OECD/NEA)

ON03

NEA’s Support to the Fukushima Daiichi Decommissioning and Addressing
Public Concerns
Ichiro Otsuka, Edward Lazo and Kentaro Funaki
Organisation for Economic Co-operation and Development/Nuclear Energy
Agency (OECD/NEA)

ON04

Strategic Approach to Managing Fukushima Daiichi Waste
Ichiro Otsuka and Gloria Kwong
Organisation for Economic Co-operation and Development/Nuclear Energy
Agency (OECD/NEA)

Session F: France ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 75
F01

IN CAN MELTER for the Conditioning of High and Medium-level and Longlived
Waste from D&D Operations
Christophe Girold1, Christine Georges1, Stéphane Catherin2, Thierry
Prevost3 and Guillaume Lecomte4
1: French Alternative Energies and Atomic Energy Commission (CEA),
2: ANDRA, 3: AREVA NC, 4: ECM Technologies

F02

Development of Laser Cutting and Dust & Fumes Collection Technologies for Fuel
Debris Retrieval
D. Roulet1, C. Georges2, C. Chagnot2, C. Journeau2 and E. Porcheron3
1: ONET Technologies,
2: French Alternative Energies and Atomic Energy Commission (CEA),
3: Institut de Radioprotection et de Sûreté Nucléaire (IRSN)

F03

Potentialities Offered by New Binders in Conditioning of Problematic Low- andIntermediate-level Radioactive Waste
David Lambertin, Christine Georges, Fabien Frizon, andCéline Cau dit Coumes,
French Alternative Energies and Atomic Energy Commission (CEA)x

F04

Foams and Gels for Decontamination of Structure and Concrete Walls
Alban Gossard1, Sylvain Faure1, Célia Lepeytre1, Agnès Grandjean1,
Christine Georges1, Jérôme Rousselle2, Lucien Gallo2, Thierry Varet3 andFabrice Moggia3
1: French Alternative Energies and Atomic Energy Commission (CEA),
2: FEVDI, 3: ORANO

F05

Fission Products, Uranium and Plutonium Measurements in Gases and Corium
after a PWR Severe Accident
L.Boucher, P.G. Allinei, H.Hupont, F.Jallu and C.Roure,
French Alternative Energies and Atomic Energy Commission (CEA)

F06

Sorbmatech Sorbent for Cesium Removal from Contaminated Liquid Effluents
Yves Barre1, Audrey Hertz, Jérémy Mascarade1, Agnès Grandjean1,
Christine Georges1, Eric Louradour2, Didier Dhaller2 and Thierry Prevost3
1: French Alternative Energies and Atomic Energy Commission (CEA),
2: CTI, 3: ORANO

F07

Virtual Reality: a way to prepare and optimize operations in decommissioning
projects
Caroline Chabal, Yves Soulabaille and Christine Georges
French Alternative Energies and Atomic Energy Commission (CEA)

F08

Supply Chain for Decommissioning in France
Christine Georges1, Anne Sophie Defay2, Serge Perez1, Serge Bouffard3 andClaudia Roche4
1: French Alternative Energies and Atomic Energy Commission (CEA),
2: AIFEN, 3: Nucléopolis, 4: Nuclear Valley

Session GB： Great Britain ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 83
GB01

The NDA: Supporting Innovation

GB02

Remote Operations: Under the Water

GB03

Remote Operations: On the Ground

GB04

Remote Operations: In the Air*GB01 to GB04 have been produced by the British Embassy Tokyo in cooperation
with the following organizations.
Nuclear Decommissioning Authority (NDA), UK Atomic Energy Authority(UKAEA), Nuclear Industry Association (NIA), Cavendish Nuclear Ltd.,
Wood PLC,

DBD

International(DBD),

NuclearAdvanced

Manufacturing Research Centre (Nuclear AMRC), Mott MacDonald,
National Nuclear Laboratory (NNL), James Fisher Nuclear Ltd., Createc,
Nuvia, Innovative Physics Limited (IPL), International Nuclear Services(INS), Atkins, PCubed, Rider Levett Bucknall (RLB), Sellafield Ltd., Veolia,xiRolls Royce, NIS Ltd., Goodwin PLC, and Gleeds
Session R: Russia ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 87
R01

Polyfunctional Adsorbent LHT-9: “All-in-One” Platform for the Development ofMulti-Nuclide Removal Systems
Sergey Britvin1, Boris Burakov2, Valery Mararitsa1, Yury Demidov1, Yury
Petrov2, Bella Zubekhina2 and Victoria Gribova1
1: Socium Ltd., 2: V.G. Khlopin Radium Institute (KRI)

R02

Development of Prediction Model of the F1’s Fuel Debris Accelerated Aging
Main approaches for estimation of fuel debris properties changing.
Oleg Bagryanov1, Sergei Poglyad2 and Boris Burakov3
1: TENEX, 2: Research Institute of Atomic Reactors (RIAR),
3: V.G. Khlopin Radium Institute (KRI)

R03

Mobile Robotic Manipulator of Anthropomorphic Design
Sergei Floria and Dmitry Kucherov
RosRAO

Session SK: South Korea ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 90
SK01

A Tritium Measurement Technology of Underwater around Decommissioning Site
Ki Joon Kang, Jun Woo Bae and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK02

Detection Technique for Real-time Monitoring of Tritium in Underwater
Jun Woo Bae, Ki Joon Kang and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK03

The Efficiency Calculation of Plastic Scintillator for in-situ Beta Measurement
System in Groundwater
Woo Nyun Choi, Ukjae Lee, Jun Woo Bae and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK04

Radiation Dose Rate Distribution Monitoring System with Portable Gamma
Detector for Decommissioning Site
Rapid and broad scale of site monitoring
UkJae Lee, Jun Woo Bae and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK05

Criteria Analysis of Autonomous Aerial and Marine Radiation Monitoring System
for Safe NPP Decommissioning
Chanki Lee and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

xii

Session US: United States of America ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 95
US01

AVANTech’s Advanced Polymer Solidification for Radioactive Waste Stabilization:
Case Studies
Tracy Barker, Gary Benda, Jim Braun, Raja Beereddy and
Dennis Brunsell
AVANTech, Inc.,

US02

Monitoring Internal Conditions of Radioactive Waste Storage Containers
Gregory Allan
PentekUS03

UK National Waste Program: A Successful Transformation in Radioactive Waste
Management
Tim Carraway, Dennis Thompson and Max Ehrhardt
AECOM

US04

EnergySolutions Decommissioning Experience
Colin Austin1 and Makoto Kikuchi2
1: EnergySolutions, 2: MKC Consulting

US05

EnergySolutions Waste Management Experience
Colin Austin1 and Makoto Kikuchi2
1: EnergySolutions, 2: MKC Consulting

List of Organizations ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 100

xiii

A01

Overview of Research & Development for Decommissioning of the
Fukushima Daiichi Nuclear Power Station
Setsuo Nakashima, Takuto Miyamoto , Hidenori Nomura and Takuro Ueda

Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF)
Abstract
Overview of R&D for Decommissioning of the Fukushima Daiichi Nuclear Power Station (1F) is described.
From basic and fundamental research to applied and practical research are carried out by various institutions, forexample the International Research Institute for Nuclear Decommissioning (IRID), the Japan Atomic Energy
Agency (JAEA), Universities, etc. The NDF has been undertaking optimization of R&D activities carried out by
each institution for a better effectiveness and efficiency as a whole.
1. Overview of R&D for 1F Decommissioning
Several R&D Projects toward to practical outcome are carried
TEPCO

out by R&D institutions as the IRID. Basic and fundamental

R&D
1F
Institutions Decommissi
oning
as IRID

researches are carried out by the JAEA and basic research

Basic ResearchInstitutions(e.g.
Universities)

institutions (e.g. Universities). The Tokyo Electric Company
Holdings, Inc. (TEPCO) carry out R&D that is highly practical

Practical
Development

JAEA

AppliedResearch

research. The main R&D themes are “retrieval of fuel debris” andBasic Research“processing and disposal of radioactive waste”. Fig.1 is a conceptual
picture of the division of roles among main R&D institutions for the
1F Decommissioning.

Fig.1 Overview of R&D for 1F Decommissioning

2. Structure of R&D for 1F Decommissioning
Whole Picture of structure of R&D for the 1F Decommissioning shows Fig.2. There are variety oforganizations engaged with the 1F decommissioning R&D projects. The NDF has been undertaking optimizationof R&D activities carried out by each institution for a better effectiveness and efficiency as a whole.
Decommissioning R&D Partnership Council conducts coordination and cooperation among each organization.
METI

MEXT

Project of Decommissioning andContaminated Water Management

Human Resource Development andResearch Project for
Decommissioning(throughConcentrating wisdom)Management
Expenses Grant

Funding

JAEA

NDF

CLADS

Support(Members）
NDF,JAEA,TEPCO,IRID,Plant Makers RelatedExperts, METI、MEXT

Information sharing ofTechnical Development
and planing

Administering

Othersmakers, etc（including
foreignenterprises）

IRID

Naraha Remote Technology
Development Center

Commission

Project
Management
OfficeForeign
Governme
ntal
Agencies

Administering/
Collaborative
Research
Administering

Okuma Analysis andResearch crnter
Fukushima Environmental
Safety Center

Management

Request for study
and Support

Commissioning Research Institutes

TEPCO

National R&D Agency,
Plant makers, ElectricPower companies, etc.

Domestic/Foreign Enterprises

ClericalInstruction

Information
sharing

Subsidized Project Operating Entities

Cooperative
Relationship

Decommissioning R&D Partnership
CoucilAssignment
Project
Management Office

Assignment

Decommissioning Strategy CommitteFund for Decommissioning andContaminated Water Management

Grant

Funding

Sector of Fukushima R&D

Fund Establishment Organization

Cooperative
RelationshipExchangeneeds andseeds of
Research

Platform of Basic
Research for
Decommissioning

Acceleration
Program for Reactor
Decommissioning
R&D

Adopted Universities,
etc.

Adopted Universities,
etc.

Participation

Fig.2 Structure of R&D for 1F Decommissioning

1

Human Resource
Development andResearch Programfor Decommissioning

Joint
Research Foreign
Universities/
Research
Institutes

A02

Activities of Review Committee on Decommissioning of Fukushima
Daiichi Nuclear Power Station in Atomic Energy Society of Japan
Noriko Asanuma1, Michitaka Sasoh2 and Hiroshi Miyano3
1

Tokai Univ., 2Toshiba Corp., 3Hosei Univ.

Abstract
In view of long term activities under the well-organized structure, Review Committee on Decommissioningof Fukushima Daiichi NPS is established.

Five subcommittees are installed and each activity are introduced.

1. Introduction
Review Committee on Decommissioning of the Fukushima Daiichi Nuclear Power Station, which is directly
connected under the board of directors of Atomic Energy Society of Japan (AESJ), is established in 2014.

In ourcommittee, the knowledge of academia is widely gathered, and activities applied each expertise is ongoing, forexample extraction of the problems and discussion of the countermeasures.

Based on these activities, helpful

suggestion and proposition can be submitted, that is the aim of the action in our committee.
As shown in Figure 1, 5 subcommittees are installed and each activity are currently accelerated.

Based onthe activities of subcommittees, the actions of our committee are introduced.
2. Activities of Subcommittees
In Subcommittee on Risk Assessment and Management, risk assessment methodology of fuel assembly
retrieval from a spent fuel pool as an example was investigated.

With this result, the risk assessment method of

the decommissioning of the actual nuclear power system is developed.
unresolved issues of the accident are summarized.

In the AESJ investigation report,

In Subcommittee on Follow-up the Proposal for Lessons

Learned and the Unresolved Events, the unresolved issues of the accident and the up-to-date knowledges onthese issues are summarized, and the final report was released [1].

Subcommittee on Robotics is a jointcommittee which is constructed by Robotics Society of Japan and AESJ with cooperation in 2015.

In 2016, robot

technology competition for decommissioning, where
typical environment and task for fuel debris retrieval aredefined, was conducted and creative ideas were collectedfrom public and both society members.

Continuing on,

a new competition on robot idea of reactor inside survey
will be held.

In Subcommittee on Structural

Performance of Building, aseismic soundness of thereactor building No.3 damaged by the hydrogen
explosion was verified and the interim report will be
released.

In Subcommittee on Radioactive Waste

Management, the end state as well as the scenarios are
discussed and evaluated thoroughly and carefully in termsof factors such as feasibility, radiation safety, technology
availability, and consistency with the national framework.
Reference[1] The final report (in Japanese) can be downloaded from

Figure 1. Organization chart of the committee.

“http://www.aesj.net/activity/activity_for_fukushima/public”.

2

A03

Latest Activity of AESJ Research Committee on Fission ProductBehavior under Severe Accident
1

2

2

Yosuke Katsumura , Shunsuke Uchida , Masahiko Osaka ,Hidetoshi Karasawa 3, Koichi Nakamura 4 and Junichi Takagi5
1

Japan Radioisotope Association, 2 Japan Atomic Energy Agency,

3

Institute of Applied Energy, 4 Central Research Institute of Electric Power Industry,5

Toshiba Energy Systems & Solutions Corporation

Abstract
In order to revive research activities related to fission product (FP) behavior under severe accidents (SA), a
research committee on FP has been organized in the Atomic Energy Society of Japan (AESJ). Major targets ofthe committee and its procedures for contributing to reliabilities of operating plants and decommissioning ofFukushima Daiichi NPP are introduced.
1. Introduction
Understanding FP behavior under SAs is one of the key issues for plant reliability based on defense indepth as well as decommissioning Fukushima Daiichi NPP. Unfortunately, as a consequence of long quiet period
in fuel defects and anomaly plant operations, number of researchers and engineers related to FP behavior
decreased and decreased to have apprehension of inadequate technical transfer of knowledge base on FP
behavior for long-term decommissioning. Under the background, the AESJ reorganized the research committee
on FP.
2. Major Targets of the Committee and its latest activities
The committee put its major targets on (1) contribution to Fukushima decommissioning and (2) increaseabilities and reliability of advanced SA analysis codes.
Three working groups (WGs) are collaboratively promoted to establish the targets under supports of the
AESJ divisions, e.g., water Chemistry, Nuclear fuel,ThermalHydraulic,

Environmentand

WG1 Experiments

Primary approaches
related to FP
qualitative and quantitative understandof FP transport phenomena
Final targets

Safetydivisions (Figure 1). Based on the survey results of
the Phébus FP projects [1], latest knowledge on FP

Experimental data
for analysis

in Fukushima Daiichi NPP should be compiled to a

Analytical results
Proposal of experiments

new technical document for technical transfer tonext generations.

WG2 Benchmark
analysis

3. Conclusion

Confirmation of SA codes
based on BM analysis

Over a year, the committee has carried on
several numbers of technical discussion meetings toExperimental data forunderstanding phenomena

Finaltargets
Validation of
SA analysis

Interpolation ofplant data

Plant data

WG3 TS proposal

Speculation of FP
transport phenomena
based on measured data

① Contribution to Fukushima decommissioning② Improvement of SA code

polish up the approach toward the final targets.

Figure 1 Major roles of 3 WGs and their final targets

Necessary modifications on its approaches toward

the targets can be flexibly accepted based on the discussions at the poster presentation.
References[1] AESJ Division of Water Chemistry, “Fission Product Behavior Obtained from Phébus FP Projects – Application for Decommissioning
Plans of Fukushima Plants and Severe Accident Analysis ”, AESJDWC Report#2017-0001, Atomic Energy Society of Japan (May, 2017).

3

A04

Relationship between the Research and Development of the Projectof Decommissioning and Contaminated Water Management andConnection to Preliminary Engineering
Masaaki Matsumoto1, Naoki Kondo1, Keishun Nakamura1
1

Mitsubishi Research Institute, Inc. (Management Office for the Project ofDecommissioning and Contaminated Water Management)

Abstract
The Ministry of Economy, Trade and Industry established a fund since fiscal 2013 and has implemented
the “Project of Decommissioning and Contaminated Water Management" as a grant-funded program by publicbid, which supports research and development with high technical difficulty. In this presentation, we will show abriefing on the series of processes including the link between the various subsidized projects of this program andthe results obtained in the projects until applied to the decommissioning of Fukushima Daiichi NPP.

1. Introduction
To safely and steadily carry out the decommissioning of the Fukushima Daiichi NPP, it is important togather together wisdom in Japan and overseas, and conduct research and development. For this reason, theMinistry of Economy, Trade and Industry established a fund since fiscal 2013 and has implemented the "Projectof Decommissioning and Contaminated Water Management " as a grant-funded program by public bid tosupport research and development with high technical difficulty. In fiscal 2017, 16 subsidized projects were
implemented by subsidized operating entities. The contents of research and development in these projects were
diverse. In order to apply the results obtained in the subsidy projects to the decommissioning of Fukushima
Daiichi NPP, mutual coordination among subsidized projects are essential.

2. Relationship between the subsidized projects of Decommissioning and Contaminated Water
Management program and connection to preliminary engineering
In accordance with the next R & D plan [1], the subsidized projects are divided into “Internal
Investigation”, "Development of debris retrieval method", "Development of Fundamental Technologies for
Retrieval of Fuel Debris", "Improvement of Work Environment for Retrieval of Fuel Debris”, and "Processingof Solid Waste". As a rough flow, Development of Debris Retrieval Method, Development of Fundamental
Technologies and Improvement of Work Environment are carried out based on information inside the reactor
vessel obtained by Internal Investigation. In addition, the results of investigation in Development of Debris
Retrieval Method are reflected in the preliminary engineering by the operating entities including concretizationof fuel debris removal process. Processing of Solid Waste also study in parallel with studying on debris retrieval.
In this way, current projects are organically related, and preparations are being made for engineering by
subsidized entities.

References[1] Ministry of Economy, Trade and Industry, Agency for Natural Resources and Energy, Progress of R & D Project andDirection of Next Plan, Decommissioning and Contaminated Water Management Team Meeting / Project Management
Office Meeting (51st) (Japanese only)

4

A05

Advanced Research and Education Program for Nuclear
Decommissioning

Toru Obara1, Kenji Takeshita1, Yukitaka Kato1, Hiroshi Akatsuka1, Hiroshige Kikura1, TakehikoTsukahara1, Katsumi Yoshita1, Koichiro Takao1, Jun Nishiyama1, Masayuki Harada1, Nobuyuki
Iwatsuki1, Koich Suzumori1, Gen Endo1, Kenji Kawashima2, Noriko Asanuma3, Tsuyoshi Arai4,
Naoyuki Takaki5
1

Tokyo Institute of Technology, 2Tokyo Medical and Dental University, 3Tokai University, 4Shibaura
Institute of Technology, 5Tokyo City University

Abstract
Advanced Research and Education Program for Nuclear Decommissioning has been started in 2014 for thesophisticated education and advanced research for the decommissioning of Fukushima Daiichi NPS.

1. Introduction
Advanced Research and Education Program for Nuclear Decommissioning (ARED) has been started in2014 by Tokyo Institute of Technology with the collaboration of Tokyo Medical and Dental University, Tokai
University, Shibaura Institute of Technology, and Tokyo City University with financial support of the Mistry of
Education, Culture, Sports, Science and Technology (MEXT). The purpose of the program is to perform
sophisticated education of graduate school level for the decommissioning of Fukushima Daiichi NPS and toproceed advanced researches to solve the technical issues in the decommissioning.

2. Educational Activities
Some classes for the graduated school students have been offered by the program, which include
laboratories using nuclear fuel and radioactive materials. The students perform experiments in radioactive
hazard laboratories to obtain the fundamental and scientific knowledge of the materials and the method to tread
them. In addition to that here is a class of laboratory for the robotics and remote sensing. The program includes
lectures for the decommissioning of damaged cores and for the decommissioning of the reactors which are shut
down by the life. The internships and seminars for the decommissioning are also offered.

3. Research Activities
Several advanced researches for the decommissioning are in progress including analysis method ofcontaminated material with microanalysis technique, decontamination and stabilization technology ofcontaminated materials, robotics and remote sensing technology, and criticality safety study in the reloading of
the fuel debris. They are performed by the collaboration of the universities. Seminars about the research
progress are held regularly to exchange information with nuclear industries.

4. Concluding Remarks
The educational activities and researches have been proceeded successfully. Some research topics havebeen moved to the phase of joint study with nuclear industry for the practical use. To continue the sophisticatededucation will improve the skill of researchers and engineers who involve in the decommissioning.

5

A06

Contribution to Fukushima Daiichi unit 5 and 6 strategic decommissioning
utilizing experience of decommissioningin Japanand oversea
12
1

Toyoaki Yamauchi , Colin.R.Austin
The Japan Atomic Power Company, 2Energy solutions

Abstract
The Japan Atomic Power Company(JAPC) and Energy Solutions(ES) can contribute to the safe and efficient
decommissioning of Fukushima Daiichi unit 5 and 6 by utilizing decommission knowledge of both ES, specialized
in decommissioning in US, and JAPC in Japan.
Domestic first decommissioning of commercial nuclear power plant
JAPC has executed domestic first decommissioning of commercial Tokai nuclear plant since 2001. In addition,
since 2017 JAPC has also started decommissioning of Tsuruga No.1 nuclear plant(BWR) efficiency introducing
decommissioning project management by setting Tsuruga Decommissioning Project Center this April.

Removing work of the Turbine and Generator at Tokai nuclear plant

Fukushima reconstruction support with 180 personnels from JAPC group
JAPC concluded the cooperation agreement withTEPCO in March 2015, and currently about 130
personnels of JAPC group support decommissioningof Fukushima Daiichi with TEPCO.
Also, about 50 personnels of JAPC group work forradioactive waste treatment in Fukushima, total about
180 JAPC personnels support Fukushimareconstruction.
JAPC group member at Fukushima Daiichi
Utilizing the decommissioning know-how from overseas
JAPC agreed with ES to obtain and utilize the ES
decommissioning know-how in April 2016, and agreed toutilize that know-how for the domestic decommissioning.
So far, we have conducted exchanging eachdecommissioning members and making efforts for the safeand efficient decommissioning of Tsuruga Unit1 and so on,
considering the joint business in the future.

Figure 3. Workshop at Zion plant in US

Workshop at Zion in US
Summary
JAPC and ES believe that early decommissioning of Fukushima daiichi unit 5 and 6 could contribute to
decreasing the maintenance cost and project risk and utilizing these units for decommissioning of the whole
Fukushima Daiichi.
The key factors for successful decommissioning are the proper decommissioning mind set and proper project
management, especially waste management of the removed equipment. First, the project of decommissioning ofunit 5 and 6, which are similar to normal plants, has to be treated safely and efficiently, after that utilizing theknowledge obtained from that experience for the decommissioning project of other units will have to be conducted,
which could lead to decommissioning of whole Fukushima daiichi safely, steady and efficiently.

6

B01

The Fukushima Consortium of Robotics Research
for Decommissioning and Disaster Response
Yoshiro Owadano1,
1
Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response

Abstract
In order to promote the market entry of manufacturing companies in Fukushima Prefecture to thedecommissioning, decontamination and disaster response robotics fields, Fukushima Consortium of Robotics
Research for Decommissioning and Disaster Response (hereinafter called “the Consortium”) is working toestablish a network among manufacturing industries in Fukushima Prefecture and related organizations, throughseminars and exhibitions.
1. Overview of the Consortium
The Consortium was established to support the manufacturing companies in Fukushima Prefecture which
intend to participate in the robotics for decommissioning of nuclear power plant, and to expand their business todisaster response robotics by utilizing their technologies developed and accumulated in the decommissioning
projects. This Consortium organizes seminars, demonstrations and exhibitions of robots and other products for
networking and information exchange with related organizations.
2. Main activities of the Consortium
2-1. Robots demonstration events and exhibitions
The Consortium exhibits its products at the exhibition where
many related people visit. In addition, the Consortium holds a
robots demonstration events in JAEA to appeal new robots bydemonstrating in front of potential customers.
2-2. Technical seminars and matching events
The Consortium provides its members opportunities toexchange technical information useful at site ofdecommissioning and disaster. It also offers meetings fordiscussing various use of the robots under development.

Figure 1. Robots demonstration events

2-3. Other activities
The Consortium members can be supported by expert
coordinators. Market survey and site visit to related organizations,and information gathering from governmental organizations are
possible. Besides, latest project information of new governments
will be provided timely to our members.
Figure 2. Technology matching events
3. Unique Features of the Consortium
Because the Consortium is managed by Fukushima Technology Centre, a public research and testingorganization established by Fukushima local government, the members can exchange information withcompanies involved in the rolated projects, and also can get the latest information about the decommissioning
projects headed by the national government which is normally difficult for local companies to get.

7

B02

Robots Developed by the Members of
Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response
Yoshiro Owadano 1,
1

Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response

Abstract
Aiming for the entry into the market of decommissioning, decontamination and disaster response robotics,
member companies of the Fukushima Consortium of Robotics Research for Decommissioning and Disaster
Response have developed robots with unique features. Below are some examples.

1. Robots developed by the Consortium members
1-1. Drone for high sensitivity radiation measurement 「NJH950-Ⅱ」（NESI, Inc.）
Drone for high sensitivity radiation measurement 「NJH950-Ⅱ」 canmeasure surface radiation while flying over measuring area without
stopping at the measuring point and instantly visualize the results by

Figure 1. NJH950-Ⅱ

installing nine GAGG high sensitivity scintillators. Also, by increasing
the battery and reducing the weight of the load, we could dramaticallyextend the flight time.
1-2. Autonomous Cleaner ReFRO – β Version –（Haloworld Inc.）
Autonomous Cleaner ReFRO – β Version – can create a 3D map with
SLAM using laser sensors etc.. Also, it runs autonomously based on the
3D map, and if there is an obstacle, it can be avoided. It can clean indoor
Figure 2. Autonomous Cleaner
ReFRO – β Version –and outdoor surfaces.
1-3. Fire Fighting Robot（Kouyou Co., Ltd.）This fire-fighting robot uses water pressure as the remote control
medium. Therefore, there is no electrical component in the main body,and secondary damage doesn’t occur. Also, by stretching the hose, it ispossible to extend water discharge and remote operation distance.

Figure 3. Fire Fighting

2. Others

In addition to the robots introduced above, many other products developed by the Consortium members areavailable.
We are pleased to introduce our member companies capable of helping your new product development. They
can provide proper technical solutions in your business related to the decommissioning and disaster response. So,
if you are interested in such services, please contact the secretariat of the Consortium.

8

B03

Organization Profile of IRID
1

Naoaki Okuzumi1
International Research Institute for Nuclear Decommissioning (IRID),

Abstract
Ever since its establishment in August 2013, the International Research Institute for Nuclear
Decommissioning (IRID) has been fully committed to an urgent challenge—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 and domesticorganizations●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. Overview of the Mid-and-Long-Term Roadmap

9

B04

Overview of IRID R&D Projects
1

Naoaki Okuzumi1
International Research Institute for Nuclear Decommissioning (IRID),

Abstract
Towards the decommissioning of the Fukushima Daiichi NPS, four entities cooperate closely to work together
as one team. IRID is a complex entity consisted of eighteen organizations that play the leading role in R&D forthe decommissioning of the Fukushima Daiichi NPS. IRID is conducting R&D projects; “Preparation of Fuel
Debris Retrieval” and “Treatment and Disposal of Radioactive Waste”. In preparation for determination of fueldebris retrieval methods in 2019, IRID continues to develop technologies for; 1.Intensive investigation on theexistence of fuel debris and damaged conditions inside the reactor, 2.Potential risk management and verification
for nuclear safety, 3.Reliable remote operation under high radiation environments.
1. Progress of Research and Development
As R&D projects for the preparation of fuel debris retrieval, IRID is undertaking research based on threeelements. Firstly, detection technology that enables to directly retrieve fuel debris in the PCV has been developed.
In April 2015, a robot successfully entered the PCV at Unit 1. In FY 2016, the preparation of fuel debris
investigation outside the Pedestal was conducted. At the same time, fuel debris investigation robots inside thePedestal, and remotely operated drilling device for the PCV penetration have been developed to reduce theexposure of workers at Unit 2. Additionally, investigation of fuel debris inside the Pedestal, an underwater
swimming robot for Unit 3, and a new survey device equipped with a telescopic pipe and cameras for Unit 2 havebeen developed. When these robots remotely access fuel debris, it will enable to investigate a distribution andcompositions of fuel debris. Severe Accident Analysis Code is upgraded to identify fuel debris inside the reactor,
and investigations through the cosmic-ray Muon are conducted. The distribution of fuel debris in the reactor isinvestigated from outside the reactor building through the Muon at Unit 1. It resulted that the possibility is smallthat a large amount of fuel remains in the reactor core. The Muon transmission measurement was conducted at
Unit 2 from March to July 2016, and at Unit3 from May to September 2017. Essential technologies such as accessto fuel debris in the RPV or the PCV are ongoing. It is necessary to secure nuclear safety risks.
2. Future Development
With an aim to advance a very tough task of the decommissioning, IRID intends to proceed with R&D,
collecting knowledge and expertise from around the world. Specifically, not only hardware support from overseaorganizations is required for removal and storage of damaged fuel, but also experiences of safety management andprocess.

10

B05

The Future of Remote Access for 1F Decommissioning(VNS Integrated Nuclear Remote Access)Akira Ikeda1, Mamoru Numata1, Matt Cole2, Scott Martin2 and Marc Rood2
1

Kurion Japan k.k, 2Veolia Nuclear Solutions, Inc.

Abstract
Veolia Nuclear Solutions (VNS) has an extensive record tackling the most difficult decommissioning
challenges around the world. Taking advantage of our deep project management experience and best-in-classtechnology, VNS has been – and will continue to be – a trusted solutions provider for 1F. Utilizing the latest
access technology, VNS has already had a positive impact on several challenging areas at 1F.
1. Introduction
Veolia Group is the global leader in optimized resource management. VNS provides the mostcomprehensive range of technologies, expertise and services for decommissioning, facility restoration andtreatment of radioactive waste. VNS is already engaged and cooperating on the 1F decommissioning, and is
optimistic that the company’s successful Access technology can be further deployed to help encourage theeffective decommissioning of 1F. This paper provides a brief introduction to VNS achievements in Japan andthroughout the world.
2. VNS Access technology
2-1. VNS Experience Throughout the World
Project: Rotary Deployment Arm•

Location: Trawsfynydd NPP, Wales, UK Date: 2009 - 2011•

Two (2) remote tank-cleaning arms for four (4) different storage tanks.
Waste forms included liquid, sludge and solid (Fuel Element Debris).

Trawsfynydd NPP

Project: Whiteshell Standpipes and Bunker Retrieval•

Location: Whiteshell Laboratories, Canada•

Access and retrieve waste from 177 buried standpipes and several deep

Date: 2016 – Present

bunkers. Waste must be removed and processed for long term disposal.
Whiteshell

2-2. VNS Experience in Japan
Project: Fukushima Inspection and Repair Manipulator•

Location: Fukushima, Japan Date: 2012 – Present•

Two (2) long-reach manipulator systems to inspect and repair leaks that haveoccurred in the primary containment vessel of the Unit 2 Reactor at 1F.

2-3. VNS Capability in the Future on 1F
Project: D-pit sludge retrieval and Fuel Debris Retrieval in the future
3. Conclusion

Fukushima

VNS couples cutting-edge technologies with deep management expertise to provide bespoke solutions fortoday and the future. VNS will continue to benefit the 1F decommissioning by providing both trusted experienceand Access technology.

11

B06

Development of a remote concrete sampling device
Kiyoaki Kato, Akira Horii and Toshihiko Mizushima

ATOX CO., LTD. / IRID (International Research Institute for Nuclear Decommissioning)
Abstract
In order to safely process and dispose of solid waste generated by the accident at Fukushima Daiichi
Nuclear Power Station, it is necessary to analyze its properties. Therefore, we have been developing a remotesampling device of the floor concrete for characterization of rubbles inside the reactor building under ahigh-dose radiation environment.
1. Introduction
As huge amount of radioactive solid waste is expected to generate at Fukushima Daiichi Nuclear Power
Station, it is necessary to develop technology that can safely process and dispose the radioactive wastes based oninventory evaluation by analysis of radioactive waste. Therefore, the method of sampling the floor concrete

was established and the remote sampling device was developed for characterization of rubbles inside thereactor building under a high-dose radiation environment.
2. Development overview
2-1. Core Bit with wedge of core drilling
The shape of the sample to take from the concrete on the floor was decided as the core, and the sampling
method adopted the core drilling method. In the usual core drilling method, actions of drilling, core cut off andcore removal are needed. However, it might be a complicated method if all actions are performed remotely. In
this study, by providing wedge inside the bit, core cutting off andholding were performed at the same time so that series ofoperations can be completed only by "drilling" operation.
2-2. Remote sampling device for Kobra
In order to confirm the feasibility of the remote core drilling
operation by using core bit with wedge, the remote sampling

Wedge

Figure 1. Core Bit and Sampling Core

device for Kobra was developed, which had abundant performances in
the field work at Fukushima Daiichi Nuclear Power Station.
3. Conclusion
It was confirmed that the core sample can be collected remotelyfrom the floor concrete by this study. In the future, we will promote thedevelopment of remote sampling device that can be mounted on Packbot,
which is widely used in the field work at Fukushima Daiichi Nuclear
Power Station, to expand the collectable area for concrete samples.

Figure 2. Sampling device(For Kobra)

Acknowledgments
This presentation includes some of the results from the works under the subsidy program“Project ofDecommissioning and Contaminated Water Management” by METI.

12

B07

Usage of remote technologies for D&D operations in hostileenvironment / Orano’s experience serving Fukushima-Daiichi
D. Ogawa1, Y. Kikkawa1, N. Nakamura1, N. Breton1
1
Orano Atox D&D Solutions Co., Ltd. (ANADEC)

Abstract
The use of remote technologies is required when performing D&D i operations in high dose rateenvironments. A combination of scenario making, engineering and preparation for on-site operation is needed topropose adequate, reliable and cost effective remote solutions for D&D works. Orano experience in nuclearizing
off-the-shelf industrial equipment to develop remote solution perfectly fitted to each problem is unique. ANADEC
can benefit from this Orano experience to tackle the various challenges in Fukushima-Daiichi (1F).
1. Introduction
Since 2011 the situation of 1F site has greatly improved, but the remaining works are very challenging interms of dose rate, contamination risk and access difficulty. The ALARAii principle, and other processes used by
Orano when implementing remote technologies on challenging D&D sites such as 1F, are presented below.
2. Principles for implementation of remote technologies and equipment for D&D works
2-1. Definition of a scenario based on ALARA principle
The first step of scenario selection is to put forward several
scenarios for a given upper limit of radiation exposure. Only then,through comparison as per key criteria (schedule, cost, dose rate, risk,
waste volume, public acceptance, etc.), one scenario among all is
adopted to serve as guide for engineering work. While remotetechnology is not systematically used, it is often key to limit radiation
exposure in compliance with the ALARA principle.
2-2. Nuclearization of reliable off-the shelf equipment
D&D projects are often singular and with restrained budget, thus,

Figure 1. Example of integrated
remote solution: 3D simulation for
scenario selection
Originalacquiring industrial products allows to propose reliable solutions at low
price with little development. But adaptations for nuclear environments
or on-site access are sometimes required: radiation hardening ofsensitive parts, maintenance of contaminated device, boundary
making, handling of malfunctions, accessibility to target, improvementof visibility, etc.
2-3. Development of dedicated solution
Tailor-made solutions can also be developed to answer specificneeds. Ease of operation and maintenance, repeatability for otherprojects are basic guidelines to design versatile remote technologies fornuclear application.
2-4. Testing and training prior to on-site operation

Figure 2. Example of integrated
remote solution: Nuclearizedindustrial robot MC-TEΩ600
Mock-up and training

A series of tests and training are performed in inactive conditions to verify that the operators will be able tosolve any issue on-site. Tests and training using a simulator and/or full-scale mockup are carried out by stages asnecessary. This step is key to the successful operation in a D&D site.
3. Conclusion
Orano has acquired unique experience in remote solutions for exacting D&D projects. ANADEC benefits
fully from this experience to propose integrated solution for 1F, and its ever more demanding challenges ahead.
i

D&D: Dismantling & Decommissioningii

ALARA: As Low As Reasonably Achievable13

Remote Control of a Sensor-less Manipulator for
Decommissioning Work

B08

Takahiro Nagai, Masaki Odai, Kishiko Maruyama and Ryosuke Kobayashi
1

Hitachi, Ltd., Research & Development Group

Abstract
For a maintenance and decommissioning work of nuclear power plants, a remote control system using stereo
camera as external sensors is proposed for a sensor-less manipulator. The sensor-less manipulator is controlled inreal time by estimating manipulator poses using a state estimation filter. This paper outlines our proposed methodand describes results of an evaluation experiment using a sensor-less hydraulic manipulator.
1. Introduction
A remote-controled robot used in nuclear power plants needs to have radiation tolerance. A sensor-lessmanipulator is preferable in order to suppress radiation effects. In this work, we have developed a remote control
method for a multi-joint hydraulic manipulator.
2. Remote Control of Sensor-less Manipulator
2-1. Proposed method
We proposed the concept of a teleoperation system for the sensor-less manipulator as shown in Fig. 1. The
current joint states of the sensor-less manipulator are estimated by only stereo camera images. In order to reduce
processing load for real-time control, the stereo camera detects only 3-D positions of markers which are attachedon every link of the manipulator. The manipulator poses are calculated by using a state estimation filter based on
the kinematic model of the manipulator.

2-2. Experiment and Results
In the experiment, we evaluated calculation time ofthe pose estimation processing during operation of thehydraulic manipulator. The hydraulic manipulator with
attached markers was set between two stereo cameras. We
obtained the average calculation time, shown in Fig. 2, as
60 ms which is less than the refresh rate of the stereo

Calculation time [ms]Figure 1. Concept of teleoperation system for a sensor-less manipulator
150
100

Ave. 60 ms

50
0
0

100
200
300
Number of steps [-]400

Figure 2. Calculation time of the pose estimation

camera (100 ms). We confirmed that the manipulator pose was estimated in real time.
3. Conclusion
We proposed a remote control system for a sensor-less manipulator using a stereo camera. We confirmed thatthe manipulator pose was estimated in real time and the manipulator was controlled based on the estimated pose.

14

B09

Introduction of Mirion Technologies’s Radiation Tolerant Camera
The World’s Highest Radiation Tolerant Performance Camera
Tadashi Sato, Mikio Katsura, Yuto Tajima
CORNES Technology Limited

Introduction
Mirion Technologies (IST) Ltd. is an industry leader in supply of Radiation Tolerant Cameras and specialist
CCTV & imaging systems for the Nuclear Industry. The Mirion IST-Rees brand of radiation tolerant cameras isrecognized as the market leader in the Nuclear Industry worldwide. With a comprehensive range of products to suit
all applications in every part of the Nuclear fuel cycle, Mirion is able to offer standard or customized solutions tomeet customer needs and expectations.
1. Radiation Tolerant Performance Camera
1-1. R93/R94x Series Nuclear Camera
The R93/R94x series are the most widely used high radiation inspection nuclear
camera systems in the world, offering extreme radiation tolerance (2MGy) in acompact size with a large range of viewing attachments and options.
 Radiation Tolerance: 2MGy (2 x 108 rad), 1kGy/hr dose rate Resolution: 600 TV lines per picture height (center zone) R93 Mk3: Fixed NB Lens (9mm standard, 50 deg horizontal)
R941: NB 8mm to 24mm (57.6 deg to 20.8 deg, horizontal)
R940: NB 22mm to 90mm (23 deg to 6 deg, horizontal)
R942: NB 12mm to 72mm (40 deg to 7 deg, horizontal)
1-2. R981 Compact System Camera
The R981 compact camera is a pan-tilt-zoom camera that has very high radiation tolerance(>1MGy). Mirion has combined the camera and pan and tilt into an integral unit and provided
three positions for additional lights and microphone. The R981 Compact can be configured
into multi camera configurations, or a single camera system can be supplied.
 Radiation Tolerance: 1MGy (108 rad), 1kGy/hr doze rate Resolution: 550 TV lines, horizontal
 6:1 - 12mm to 72mm zoom, 40 deg Wide, 7 deg Narrow
6:1 - 24mm to 144mm zoom, 21 deg Wide, 3.5 deg Narrow
3:1 - 8mm to 24mm zoom, 58 deg Wide, 21 deg Narrow

1-3. SC985 Color Camera Outstation
The SC985 Color Camera outstation has been developed to provide a color image for
Mid-Radiation applications in the nuclear industry. Combining high quality images witha robust pan/tilt/zoom platform for inspection and surveillance applications either
underwater or in air. The option of on-board LED lights provides additional scene
illumination when required.
 SC985HR Radiation Tolerance: 50kGy (5Mrad), 1kGy/hr dose rate
SC985 Radiation Tolerance: 10kGy (1Mrad), 1kGy/hr dose rate
 High Quality Color Images with Resolution of 450 TV lines 10:1 NB clear glass optical zoom lens

1-4. RC911 DotCam Miniature Color Camera
The DotCam is a miniature color camera designed for use in higher radiation environments where the use ofstandard CCD based camera is not cost effective.
The compact size of the stainless steel camera body lends itself to a variety of
deployment methods including fixed and mobile installations. Integral white
LED lighting and a range of lens options increase the Dotcam HR’s versatility.
 Radiation Tolerance: 1kGy (105 rad), 300Gy/hr doze rate Resolution 400 TV lines (NTSC), 450 TV lines (PAL) 4mm standard lens (FOV 90 deg)

15

B10

High Radiation Resistant Fiberscope using Silica based Imagefiber
Tomoaki Toriya1, Shinya Taniwaki1 and Manabu Kudo1
1

Fujikura Ltd.

Abstract
Fujikura developed high radiation resistant fiberscope functional for particularly the high dose rateenvironment such as 10kGy/hr. This paper introduces some evaluation results of prototyped imagefiber.
1. Introduction
Silica glass with high-OH content core fiber is well known to perform good optical transmission against
radiation. However, durability of image observation is limited due to radiation-induced absorption. We
accomplished higher radiation durability by introducing fluorine doping and special treatment technologies.
The sample fiber has been evaluated in image quality and transmission efficiency under the dose rate of
10kGy/hr condition.
40

2-1. Material improvement
The sample imagefiber was made of silica glass
core with a tiny amount of fluorine doping, pixel
number was 6,000. Sample length was 100m andmiddle 20m of 100m was irradiated to Co-60 gammaray at 10kGy/hr.

Induced loss (dB/20m)

2. Radiation Durability Enhanced FiberscopeHigh OH core35

F dope core

30
20
15

10kGy

10

300kGy50400Induced loss peak of fluorine doped silica corefiber was about a quarter of conventional high-OH

Special treatment

10kGy

25

500

600

Wavelength (nm)

700

800

Fig 1. Induced loss after irradiation

content silica fiber at 630nm after 10kGy irradiation.
(Fig.1) However, video image after 10kGy radiation
indicates this material improvement is not enough forvisualization tool under high dose rate. (Fig.2)
2-2. Special treatment for enhancing durability
We installed special treatment to the sample imagefiber
to enhance radiation durability. This treatment inhibited thegeneration of absorption defects in the fiber and was(a) Original(b) 10 kGy irradiation

Fig 2. Video image of original fiber

stabilized transmission performance. It shows slightdegradation on image after 300kGy irradiation at 10kGy/hr.
(Fig.3)
3. Conclusion
This combination technology of fluorine doped silica
glass and special fiber treatment accomplished the highest
radiation resistant imagefiber. Fiberscope with using thisimagefiber will be the effective visualization tool which(a) Initial(b) 300 kGy irradiation

Fig 3. Video image of enhanced fiber

performs under high dose rate such as 10kGy/hr. This technology has possibilities of applying to not only fiberscope but
also single core fiber products for laser beam delivery, spectroscopy and other application.

16

B11

High Radiation Shielding Rubber Chips
Kenji Kitazume, Shinji Shimada, Hiroshi Kouda

Toyo Rubber Chip Co, Ltd.
Abstract
The RS-6 shielding rubber chips are new rubber chips which show excellent
properties for shielding radiation inside a nuclear reactor building and in anenvironment with a high radiation.

（Patent No.6074491）

Usage and method of application of RS-6 rubber chips
1.These were developed for multiple protection (6 layers of walls) for an emergency
situation in an environment handling high radiation such as the inside of a nuclear
reactor building.
2.These are very effective as shielding materials in emergency situation and shielding
construction and removal are easily done.
3,Also,they are ideal for ensuring the safety of workers in an environment handling ahigh radiation.
Characteristics of RS-6 rubber chips
The RS-6 ,the name of high shielding rubber chips developed by us ,are in the form
Of rubber chips(particle sizes of 1 ~25mm) and are processable in any desired forms(sandbags containing 10kg chips ~weather resistant flexible container with 500kg
chips) . Also the RS-6 chips can be molded to various forms by solidifying.
Shielding properties of RS-6 rubber chips
Shielding dosage rate by thickness of sandbag
Thickness(mm)Dosage rate(%)

1bag(170)59

2bag(340)80

3bag(510)

90

Dosage rate(%)
100
90
80
70
60
50
40
30

Testing condition :Blank Gamma-ray Co60
160Gy/h, the thickness of one sandbag as
170mm. With 510mm thickness, 1/10 value
thickness obtained.

20
10
0
1bags(170mm)

2 bags(340mm)

3bags (510mm)

Composition of RS-6 rubber chips
A radiation resistant synthetic rubber with an excellent weather resistance is themain component , and the RS-6 rubber chips are composite materials in which anaturally occurring barium sulfate having a high radiation shielding efficiency
is specially combined to form chips.
References
Test place; measured at Takasaki Advanced Radiation Research Institute ,JAEA .

17

B12

Radiation-Resistant Flexible PVC Material
Makoto Oi, Hidenori Shima and Ryosuke Ogata
Tigers Polymer Corporation

Abstract

Tigers Polymer has developed radiation-resistant flexible PVC material with the durability of
5 MGy of gamma radiation dose.
This material maintains the elasticity and flexibility even after radiation and it can be molded
into hose, tube, sheet, etc. with various shapes and diameters.
Moreover, material property is good and wear resistance is excellent.
The cost performance is satisfactory because of PVC material.

450
400

K508-3B

350

K514-3Tensile strength at break (Mpa)

Tensile elongation at break (%)

1. Gamma Radiation Test

300
250
200
150
100
50

0
0

20
18
16
14
12
10
8
6
4
2
0

K508-3B
K514-30

2
4
6
Total radiation dose (MGy)2
46
Total radiation dose (MGy)

2. Prototype

Tube

Hose

18

Sheet

B13

Radiation Shielding Clothing to Facilitate Decommissioningof Nuclear Reactor
Nobuhiro Shiraishi, Koji Suzuki, Yosuke Tanji

Fukushima Midori Anzen Inc.
Abstract
A new type of radiation shielding clothing designed to facilitate speedy and efficient operation andto improve the comfort and safety of the operators in order to show steady recovery of Fukushima
Prefecture from the Great East Japan Earthquake.
1.Introduction
We developed a new type of radiation shielding clothing using a rubber sheet 1/4 the thickness (75%thinner) of the existing type, while providing equal or better shielding performance as the existingradiation shielding clothing for use by the operators engaged in nuclear reactor decommissioning.
2. Four functions of the radiation shielding clothing
2-1. Thickness of the radiation shielding sheet
The radiation shield sheet of the existing products mainly used in the reactor decommissioning site is made of leadwith a thickness of about 4 mm. Considering the adverse effect of lead on the human body, we developed 1 mm
thick radiation shield sheet made of rubber with tungsten kneaded into the molecular level as the radiation
shielding material, which results in an increase in the amount of tungsten per unit area.
2-2. Shielding effect
Shielding effect of 10% is realized with the thickness of 1 mm.
Shielding performance of the new material was measured and tested by the Korea Atomic Energy ResearchInstitute (KAERI).

Measuring Method inKorea

Radiation shielding clothing

2-3. Feel of wearing and effectFeel of wearing of the operators putting on this protective clothing and the burden of the radiation
shield at the site where the operator must stand up and squat down frequently are greatly
improved by the shielding sheet made of thinner material and integrated upper and lower pieces
design. When compared with clothing designed with separate upper and lower pieces, uniform
radiation exposure can be realized. The lower shielding cover is also integrated that will reduce
exposure of the reproductive organs to radiation.
2-4. Multiple usage as the radiation shielding mat
It is considered that the material of the radiation shielding clothing can also be used as protective
covers of precision devices and as shielding mats to create a radiation shielded space in the nuclear
power plant.
2-5. Technical collaboration
This product is jointly developed in collaboration with a patent registered in Korea and a company
in Fukushima, Japan, where demands are promoting the abolishment of nuclear plants.
3. Conclusion
There are 54 nuclear power generation plants on the islands of Japan located in active seismic
zones where three tectonic plates meet. So, it will be too optimistic that a radiation hazard similar
to the one experienced in Fukushima will never happen. In order to realize recovery of Fukushima
Prefecture from the impacts of the Great East Japan Earthquake, improvements in the operating
efficiency of the operators engaged in reactor decommissioning or in the transport of radioactive
contaminants and commercial production, practical application, and industrialization of technicaldevelopment for new tools and devices are necessary. We intend to integrate the patented
technologies with other necessary technologies and to establish manufacturing and marketing
facilities concentrated in the Fukushima area to contribute to the promotion of the regional
economy.

19

C01

Development of a Fiber Optic Radiation Monitor
Used in Severe Environments

Takahiro Tadokoro1, Shuichi Hatakeyama1, Katsunori Ueno1, Yuichiro Ueno1, Keisuke Sasaki1,
Yoshinobu Sakakibara1, Toru Shibutani1, Takahiro Ito1, Mikio Koyoma1 and Koji Nebashi2
1

Hitachi Ltd., 2Hitachi-GE Nuclear Energy, Ltd.

Abstract
We have been developing a fiber optic radiation monitor that uses a near infrared photon counting method.
The monitor is capable of measuring dose rates from 1x10-2 Gy/h to 6.1x104 Gy/h and dose rate linearity is
within ±4% of full scale after the heat and pressure resistance test. In addition, measurement conditions of themonitor can be checked while it is operating using a semiconductor laser.
1. Introduction
It is necessary to investigate the current plant status of the Fukushima Daiichi Nuclear Power Station (1F)
to provide the planning and the actual decommissioning of 1F. It is also necessary to measure the dose rate in
the primary containment vessel (PCV) in any future severe nuclear accident for monitoring plant conditions.
2. Fiber optic radiation monitor

Optical Fiber Optical Divider
Radiation(~300m)Part
Detector Part

Fig. 1 schematically outlines a fiber optic

PhotonMeasurement
Partradiation monitor. The radiation detector part consistsof a light emission element (Nd:YAG) which generates
1064 nm photons when a gamma-ray or a laser beam

Nd:YAG
Housing

is irradiated to it. In the photon measurement part,

808nm (Laser Beam)
Semiconductor
1064nm (Photon)
Laser Part

each photon is converted to an electron pulse and it iscounted. In the control, analysis and database part,

Control
DeviceDatabase
Control, Analysis
and Database Part

counting rates of photons are converted to the dose
rate. The semiconductor laser part emits the 808 nm

Figure 1. Outline of a fiber optic radiation monitor
with a measurement condition check systemlaser beam for checking the measurement condition of
the monitor.
3. Test results
The dose rate linearity test results are shown in

1012
Photon counting rate
[cpm (counts per minute)]

Fig. 2. We confirmed that the monitor had a capability
for measuring dose rates from 1x10-2 Gy/h to 6.1x104
Gy/h. We also confirmed the dose rate linearity waswithin ±4% of full scale after the 84-h heat (300oC)
and pressure (1x106 Pa) resistance test.
4. Conclusion
We have been developing a fiber optic radiation
monitor that uses a near infrared photon counting

1010
108
106
104
102 -310

method. As a result of the dose rate linearity test, we
confirmed the monitor was able to be used in severe
environments.

20

no light attenuation filter
light attenuation filter×3(×10 3 of measurement value)

10-1

101
103
105
107
Dose rate [Gy/h]Figure 2. Dose rate linearity test results

C02

Progress of 1F PCV contaminant analysis by TEM observation
Akihiro Suzuki1, Yuji Kitsunai 1, Ikken Sato2,3
1

Nippon Nuclear Fuel Development Co.Ltd,

2

Japan Atomic Energy Agency, 3International Research Institute for Nuclear Decommissioning

Abstract
Several contaminant samples from 1F1 primary containment vessel are carefully observed by transmission
electron microscope to find micro-scale uranium-containing particles in their chemical forms of U-richc-(U,Zr)O2 and Zr-rich t-(Zr,U)O2, which are also observed in fuel debris of TMI-2 and Chernobyl NPP-4.
1. Introduction
As the progress of the containment vessel (PCV) survey, contaminants from the PCV have been obtained.
The observation of the contaminants, which may contain trace amount of materials related to fuel debris andchemical condition during the process of debris formation, is important to know chemical characteristics ofdebris, FP and their distribution. In April 2017, contaminant samples were obtained from soil-like sediments at
the PCV bottom of Fukushima Daiichi NPP (1F) unit 1. After the simplified fluorescent X-ray analysis in the 1F
site, the contaminants are transported to Nippon Nuclear Fuel Development Co. Ltd. for further observation by
SEM (Scanning Electron Microscope) and TEM (Transmission Electron Microscope) analysis.

2. SEM/TEM observation
A portion of the contaminants picked by the tip of a skewer
was used for the SEM observation. EDS (Energy Dispersive X-ray
Spectroscopy) analysis of the contaminants showed metallic
elements of Fe, Na, Si, Zn, Al, etc. The analysis also detected small
spots of U, whose distribution was not uniform as shown in Fig.1.
Four spots with high U signals were selected for TEM observation.
The TEM samples, which were picked up from the spots and sliced

Fig.1 Uranium spots on EDS mapping

into ~100nm in thickness, were set for the TEM observation
stand in a focused ion beam (FIB) device. The origins of the U signals were either a particle or a region where
several parts possibly separated from one particle existed. Typical size of each particle or region was from ~300
nm to ~10 μm. Composition and crystallinity of each part or particle were either (U0.8~0.9, Zr0.1~0.2)O2 with aface-centered cubic (FCC) structure, or (U0.1, Zr0.9)O2 with body-centered cubic (BCC) structure. Due to the factthat the U-rich (U, Zr)O2 (UO2:~90mol%) and the Zr-rich (U, Zr)O2 (UO2:~15mol%) were the typical phases ofthe fuel debris in TMI-2 and Chelnobyl NPP-4, these particles are judged to have originated from the melted
fuel and suggests that some of the fuel debris in unit 1 has similar chemical characteristics as those of the former
severe accidents. In the presentation, information from the elements other than U will be also introduced.

ACKNOWLEDGEMENT
This work was conducted as part of JFY2015 supplementary budget "subsidiary fund for Project ofDecommissioning and Contaminated Water Management (Upgrading level of grasping state inside reactor)"

21

C03

Experimental Study on Neutron Induced Gamma Ray
Spectroscopy for Characterization of Fuel Debris
Yasushi Nauchi1, Tadafumi Sano2, Hirokazu Ohta1, Shunsuke Sato1, Motomu
Suzuki1, Hironobu Unesaki2
1

Central Research Institute of Electric Power Industry, 2Kyoto University

Abstract
The neutron induced gamma ray spectroscopy (NIGS) was tested in the Kyoto university critical assembly
facility (KUCA) to develop certification methods of reduction of the infinite multiplication factor (kinf) due tostainless steel (SS) incorporated in the fuel debris and the burn-up depletion of the fuel.
1. Introduction
The fuel debris consists of irradiated fuels, SS, etc. Nevertheless, storage facilities of the fuel debris aredesigned based on an assumption where fresh fuels are arrayed in water in the optimum condition. To enableefficient storage of the fuel debris, the reduction of kinf of the fuel debris due to the incorporated SS and to theburn-up depletion should be measured. In the present work, certification methods were experimentally studied.
2. Measurement for Capture Credit
The reaction rate ratio of the neutron capture of SS to the fission corresponds to the sub-criticality margin
by the capture [1]. In the C-core of KUCA, sub-critical assemblies were mocked up with plates of SS and
93wt% enriched Uranium (93%EU) -Aluminum (Al) alloy. The assemblies were moderated by light water anddriven by a 252Cf neutron source. Fission and capture  rays were counted with a BGO scintillator. The countrate ratio of (6-10 MeV) to (3-5 MeV) is monotonously related to the calculated reaction rate ratio (Fig. 1).
3. Measurement for Residual Enrichment
A reaction rate ratio of fission to 238U(n,) is a promising indicator of the fissile enrichment. However,detection of the low energy rays from the activation of 238U is difficult for irradiated fuels due to background 
rays (E < 3.5 MeV). Then, prompt emission of a 4060 keV  ray from a 238U(n,) reaction was focused on. In
the A-core of KUCA, sub-critical assemblies of 5.4%EU were mocked up. The assemblies were moderated bypolyethylene and driven by the 252Cf source. The capture  rays were measured with a high purity Germanium
detector, simultaneously with prompt fission  rays and decay  rays from short-lived fission products (Fig.
2)[2].
4. Conclusion
For the efficient storage of the fuel debris, experimental studies of NIGS have been conducted. NIGS ispromising to certify reactivity reduction of fuel debris due to SS mixing. Also, the detection of 238U(n,) reaction
0.90
0.80
0.70
0.60
0.50
0.40
0.30
0.20
0.10

3000
count rate

count rate ratio(6-10MeV)/(3-5MeV)

by NIGS is successful. The count rate would be useful for certifications of the residual fissile enrichment.

0

0.2
0.4
0.6
0.8
1
SUS captre/fission (calculation)

1.2

135Te 97Y 95Y
19s
3.8s 10min

1000
800
600
400

238U
capture 90Rb
158s 88Br
16s

continuum part= prompt fission spectrum
3200

3600

4000
4400
pulse height (keV)

4800

Fig.2 Detection of 238U(n,) reactions by NIGS.

Fig. 1 Capture of SS to fission and count rates.
References[1] Y. Nauchi et al., JNST, 52, 1074–1083, 2015. [2] Y. Nauchi et al., Preprints of annual meeting of the AESJ-2018, 1F03.

22

C04

NDA system for Fuel Debris Characterization at
Fukushima Dai-ichi
Ryo ICHIMIYA, Atsuo SUZUKI, Mitsuo TAKEUCHI and Helene LEFEBVRE
Mirion Technologies (Canberra) KK

Abstract
After the Fukushima accident, many projects have been initiated by the International Research Institute for
Nuclear Decommissioning (IRID) for the retrieval of fuel debris from the Fukushima Dai-ichi (1F) reactors. The
IRID and Mitsubishi Heavy Industries, Ltd. (MHI) engaged in collaborative project to develop of techniques forinvestigating inside the reactor pressure vessel, by the 2014 supplementary budget of the Government of Japan’s
Ministry of Economy, Trade and Industry (METI). Under this project, Mirion Technologies undertook afeasibility study for the design of a Non Destructive Assay (NDA) system for fissile mass assessment in corium
samples.

1. Introduction
At the accident 1F reactors, fuel debris may consist of several unknown mixture of concrete, stainless steel andmelted spend fuel. The damaged fuel may also consist of a mixture of various types of fuels. These highly
inhomogeneousmaterials

mean

that

standardlaboratory

analysis

will

lead

to

difficulties

of

non-representativeness of samples. In order to solve this problem, the NDA system for fissile mass assessment incorium samples using collimated gamma spectroscopy with HPGe detectors, neutron measurement system with
Cadmium wrapped 3He tubes in a moderator (HDPE) and D-T pulsed neutron generator was proposed. Thismeasurement will be done in-situ, therefore the dose rate around the NDA system must not exceed the maximum
permissible value.
2. Solution
A bibliography study based on existing designs resulted in thedefinition of a basic design of a system:
Collimated Gamma spectroscopy with HPGe detectors
Neutron measurement system with Cadmium wrapped 3He
tubes in a moderator (HDPE), allowing for Passive (PNCC)
and Active (DDA) measurement.
D-T pulsed neutron generator
The sample capsule is moved into a tube that goes throughthe whole system.
Lead and HDPE shielding will prevent damage to the 3He

Figure 1. Conceptual System drawing

tubes and will maintain a low dose rate than outside the system.
3. Conclusion
Proven NDA system technologies with standard & new proposed data algorithm were combined to show how
the total measurement uncertainty can be minimized. Design was optimized in terms of performance, cost,weight and ALARA principals.
References[1] http://www.canberra.com/literature/case-studies/pdf/C0002_Fukushima-Case-Study.pdf

23

C05

Application of MTKK technology for D&D and Fukushima
remediation
Atsuo Suzuki, Ryo Ichimiya, Mitsuo Takeuchi and Helene Lefebvre
Mirion Technologies (Canberra) KK

1. Introduction
Mirion Technologies Canberra KK (MTKK) has more than 30 year experiences of various radiation
measurements and safety for Decontamination and Decommissioning. These experiences are actually appliedfor D&D and Fukushima remediation. Especially, In-Situ Object Counting System (ISOCS) technology isclearly demonstrated as a very useful tool for a large or complex shaped object. In this presentation, someexamples of the ISOCS application are reported by MTKK.
2. MTKK activities and achievement
2-1. Application for flexible container (Super-Sack: SS) measurement
The ISOCS system was applied at three sites of Fukushima Prefecture.
Six SSs were very inhomogeneous, 10 and 1001 SSs were inhomogeneous.
The ISOCS measurements (shown in Fig. 1) indicate very effective andaccurate results in all cases. The main uncertainty for measuring the SSs isbased on internal inhomogeneity and heterogeneous source distribution. So

Fig.1.

SS

MTKK and JAEA have developed collaboratively of the Scattering Gamma

with ISOCS.

measurement

Equivalent Model for decreasing uncertainty which is derived from thesefactors. This method decreased Total Measurement Uncertainty (TMU).
2-2. Application for rubble and block measurement
The ISOCS was applied for rubble and block measurement for getting
radioactivity balance before and after screening contaminated soils to reuse.
The ISOCS was verified as a strong tool for effective radioactive
measurements.

Fig.2. ISOCS measurement

2-3. Application for in-situ and mobile monitoringin car.

The ISOCS and Response Matrix method (RM) were applied for
emergency mobile and in-situ monitoring. Both methodologies for mobile
monitoring are very unique, so MTKK is collaborating with Fukushima
University, NIFS and Shizuoka Prefecture about it. (Fig.2)
2-4. Application for regular D&D
The ISOCS is expected use for regular D&D, so MTKK is
collaborating with some Electric Power Company. In 2017, MTKK
collaborated with Chubu EPC using the ISOCS application for Reactor

Fig.3. ISOCS measurement

Building wall and floor screening at Hamaoka NPP. The results show thatfor Reactor building wallthe ISOCS possibly contribute to cost reduction and accurate evaluationand floor.

more than conventional method (Sampling and so on). (Fig.3)
3. Conclusion and Future work
According to a lot of validation tests between the ISOCS measurements and the conventional methods, theISOCS results were found to be more effective and pragmatic than those of conventional methods in all cases.
Therefore, the ISOCS is considered to be utilized D&D and Fukushima remediation. MTKK and our user will
try to show the feasibilities of the ISOCS for getting an approval from the regulator.

24

C06

Gamma-ray measurements in high-dose radiation fields
using TlBr semiconductor detectors
Mitsuhiro Nogami1, Keitaro Hitomi1, Tatsuo Torii2, Yuki Sato2

Yoshihiko Tanimura2, Kuniaki Kawabata2, Kenichi Watanabe3, Toshiyuki Onodera4
Nobumichi Nagano1, Seong-Yun Kim1, Tatsuya Ito1, Keizo Ishii1, Hiroyuki Takahashi5
1

Tohoku Univ., 2JAEA, 3Nagoya Univ. 4Tohoku Inst. Tech. 5Univ. of Tokyo

Abstract
Our group fabricated small-volume thallium bromide (TlBr) semiconductor detectors for gamma-ray
measurements in high-dose radiation fields. The performance of the detectors was evaluated using a 137Cs checking
source at room temperature. Simulation study of the TlBr detectors were performed with PHITS code. Tlbrsemiconductor detector exhibited detector performance suitable for Fukushima Daiichi NPP decommissioning.
1. Introduction
TlBr is a compound semiconductor with high atomic numbers (81 and 35) and high density (7.56 g/cm3). The
photon stopping power of TlBr is higher than that of other semiconductor materials such as cadmium telluride(CdTe), silicon, and germanium. Small-volume TlBr semiconductor detectors were fabricated for gamma-ray
measurements in high-dose radiation fields such as Fukushima Daiichi NPP.
2. Fabrication of TlBr semiconductor detectors
TlBr semiconductor detectors were fabricated from TlBr crystals. The TlBr
crystals were cut into wafers with the dimension of approximately 5 mm × 5 mm
× 2 mm. The surfaces of the crystals were polished mechanically. Thallium
electrodes were deposited on the wafers by the vacuum evaporation method. The
device had a planar electrode (3.5 mm × 3.5 mm) on the cathode surface, and apixel electrode (0.5 mm × 0.5 mm) and surrounding guard electrode on the anodesurface. Fig.1 shows a TlBr semiconductor detector fabricated in this study.

Fig. 1 Small volume TlBr
semiconductor detector.

3. Simulation and Experiment
Deposit energy in the TlBr detector irradiated with 662 keV gamma-ray wascalculated by PHITS code [1]. Fig. 2 shows the result of the simulation. The 662
keV photo peak was observed from simulation results
The small-volume TlBr semiconductor detectors were tested with a 137Cs
checking source at room temperature. The cathode and the anode pixel wereconnected to charge-sensitive preamplifiers. Negative bias voltage of 400 V wasFig. 2 Simulation results.

applied to the cathode. The output waveforms from the preamplifier connectedto the anode pixel were recoded with a multi-channel digitizer. Fig. 3 shows a
137

Cs gamma-ray spectrum obtained from the small-volume TlBr detector using

the near cathode events. As can be seen from the Fig. 3, the detector exhibited aclear full-energy peak for 662-keV gamma-rays despite the small active volume.

Fig. 3 137Cs spectrum obtained

The energy resolution of 4.7% FWHM was obtained from detectors.

from a small-volume Tlbr

References

semiconductor detector.

[1] Tatsuhiko Sato, Yosuke Iwamoto, Shintaro Hashimoto, Tatsuhiko Ogawa, Takuya Furuta, Shinichiro Abe, Takeshi Kai, Pi-En Tsai, Norihiro Matsuda, Hiroshi Iwase, Nobuhiro Shigyo, Lembit Sihver and Koji Niita
Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02, J. Nucl. Sci. Technol. 55, 684-690 (2018)

25

C07

Piping system, risk management based on wall thinning monitoringand prediction - PYRAMID Toshiyuki Takagi1, Tetsuya Uchimoto1, Yutaka Watanabe1, Hiroshi Abe1,

Shinji Ebara1, Atsushi Iwasaki2, Ryo Morita3, Shun Watanabe3, Ryoichi Urayama1, Hongjun SUN 1,
Philippe Guy4, Christophe Reboud5, Bernard Normand6, Nicolas Mary7
1

Touhoku Univ., 2Gunma Univ., 3Central Research Institute of Electric Power Industry,
4

INSA – LVA, 5 CEA LIST, 6 INSA – MATEIS, 7ELyTMax

1. Introduction
In the decommissioning of Fukushima Daiichi Nuclear Power Station, a flow with a high concentration

debris of various kinds (concrete, corrosion, metallic…) occurs in piping when removing fuel debris.
Pipe wall thinning by Slurry Flow induced Corrosion (SFC) (Fig.1) under solid-liquid two-phase has been
anticipated. We aim at developing new tools and techniques to quantify pipe wall thinning, and provide a risk
management system based on prediction-monitoring of pipe wall thinning due to SFC in piping systems.
2. Description of the research
2-1. Clarification of wall-thinning for SFC by experiment and numerical simulations
Wall-thinning evaluation model based on the wall-thinning rate evaluation in consideration of a masstransfer coefficient evaluated under solid-liquid two-phase flow is developed. SFC is elucidated with experiment
due to the disturbance of concentration boundary layer by repeated contact of particles, the reaction rateevaluation, and the tribo-corrosion effects. The effects of solid particles on fluid factor of flow accelerated
corrosion in water piping with numerical flow simulation is elucidated. As a result, the location wherewall-thinning is the most severe and maximum wall-thinning rate are revealed by experiments and numericalsimulations.
2-2. Development of EMAT Monitoring System
A system that can monitor pipe wall thinning by
SFC with high accuracy is developed. Also, focusing
type EMAT is applied to the monitoring system. It
improves accuracy of measurement.
2-3. Evaluation of Engineering Risk
A quantitative evaluation method for engineering
Fig.1 Pipe wall thinning phenomena induced SFC

risks associated with SFC in power plant is proposed.

Specifically, PoF (Probability of Failure) evaluation in consideration of the various errors (inspection / damage
progress and applied force) by Bayesian estimation is considered. Moreover, the reasonable plan by integrativeevaluation of factors using a risk matrix is considered.
3. Summary
This project is carried out by an international collaborative research project between Japan and France
planned for 3 years since November 2017. Currently, the research is progressing towards the goal.
Acknowledgement
This study is the result of “Piping System, Risk Management based on Wall Thinning Monitoring and Prediction”
carried out under the Center of World Intelligence Project for Nuclear S&T and Human Resource Development by the
Ministry of Education, Culture, Sports, Science and Technology of Japan, and ANR of France.

26

C08

3D Reconstruction of Unit 3 Primary Containment Vessel Interiors at
Fukushima Daiichi Nuclear Power Station Using Structure from Motion
Xiaorui Qiao1, Yonghoon Ji1, Atsushi Yamashita1, and Hajime Asama1
1

The University of Tokyo

Abstract
The 3D interior structure can be reconstructed from image sequences captured by underwater robot forinternal inspection of Unit 3 Primary Containment Vessel (PCV) based on Structure from Motion (SfM) technique.
This recovered structure provides a better understanding of PCV interior environment and information on how
best to retrieve fuel debris.
1. Introduction
To advance the plant’s cleanup, fuel and other debris submerged in the coolant must be located and mapped.
In July 2017, an underwater robot was sent to Unit 3 reactor for investigation [1]. The video of interior environment
was recorded. We use this video to reconstruct the 3D structure of Unit 3 PCV interiors based on SfM technique.
2. Method
It is impossible to generate a 3D model of the whole scene in one go due to the huge size of the video.
Moreover, the scene continuity is interrupted by floating particle noise. Thus, we propose following steps toreconstruct the 3D model.
2-1. Whole video segmentation
The whole video is divided into several clips based on the scene. The scene filled with floating particles are
removed from the video.
2-2. Image extraction
Better quality images are extracted from each clip at three frames per second. The extracted images can keep
the view is continuous also certain view disparity for 3D reconstruction.
2-3. Chunk 3D model generation
Extracted images of each clip are utilized to generate chunk 3D model by PhotoScan [2]. Furthermore, therelative trajectory of robot can be estimated.
2-4. Merging into whole 3D model
Separated chunks are merged together as a whole 3D model. Each location of chunk model is determined byarranging the whole robot trajectory as a smooth one. Figure 1 shows a Unit 3
interior model by merging two chunks.
3. Conclusion
The 3D model of PCV interior can be recovered by SfM technique using
only image sequences taken by the robot. This reconstructed 3D model may
give a better understanding on interior environment of PCV.
Acknowledgements
The authors would like to thank International Research Institute for Nuclear
Decommissioning (IRID) and TOSHIBA for providing the image data.
References[1]1http://www.world-nuclear-news.org/RS-Tepco-study-of-unit-3-containment-

Figure 1. An example of part of Unit

vessel-under-way-2107176.html

3 PCV interior 3D model[2] http://www.agisoft.com/

27

C09

Research on Numerical Simulation of MCCI using MPS method
Keigo Fujikawa, Koji Okamoto, Shunichi Suzuki
Department of Nuclear Engineering and Management, The University of Tokyo

Abstract
Because the status of fuel debris depends on progress of MCCI, numerical simulation of MCCI is neededfor decommissioning of Fukushima NPP. In this research, the gas model was focused as a factor affecting theshape of concrete ablation and it was shown that gas model influenced the temperature distribution in corium
pool and this affected ablation geometry.

1. Introduction
Results from CCI experiment and simulation of CCI suggested that gas content of concrete and whether
gases were generated from bottom or sidewall of concrete cavity influence concrete ablation in MCCI [1,2]. Then,in this research, the numerical simulation of CCI experiment using MPS method was conducted treating locationof gas generation as a parameter. In addition, its effect for ablation shape was discussed.
2. Simulation method
MPS method is one of particle methods and employs Navier-Stokes equation and equation of continuity as
governing equation. In this simulation, particle distance was 0.01m and the number of particles was 6992. To
simulate forced circulation caused by gas generation, corium particles neighboring the concrete whose
temperature reached the point of dehydration or decarbonation had smaller density, which was shown in Fig. 1.
3. Results
Fig. 2 shows ablation depth in axial and lateral directions. The results show that gas model in bottom
enhanced axial ablation and the model in sidewall enhanced lateral one.

Fig. 1: Image of gas modelFig. 2: Simulation results

4. Conclusion
Location of gas generation affected ablation depth. It was supposed that gas generation enhanced ablationof concrete which was close to the generating spot by changing temperature distribution.
References[1]

M. T. Farmer et al., “OECD MCCI Project 2-D Core Concrete Interaction (CCI) Tests: Final Report”,
OECC/MCCI-2005-TR05 (2006).

[2]

Penghui CHAI, Masahiro KONDO, Nejdet ERKAN and Koji OKAMOTO, “Numerical simulation of MCCI based on
MPS method with different types of concrete”, Annals of Nuclear Energy, vol. 103, pp. 227-237 (2017).

28

C10

Structural deformation monitoring for safe dismantling of nuclear facilities
Charles VIANO1 , Masuo KADO2 and Masao TAJIMA2
1

OSMOS Group SA., 2OSMOS Association

Abstract
The dismantling work of Fukushima Daiichi Nuclear Power Plant requires a strict safety control due to its
huge weight structure and radioactive waste comprised in it. Recently a monitoring system has been developed
to monitor social infrastructures such as roads, allowing more efficient maintenance management and a longer
service life of the structure. We propose this technique to ensure safe dismantling work of nuclear reactor
facilities.
1. Introduction
Because of its huge weight structure and radioactive waste in it, the dismantling work of nuclear reactor
facilities requires a stricter safety control than any other field and necessitates before its completion a long
period of time, during which some structural instability could occur and persist. We propose a monitoring
system which makes it possible to monitor remotely static and dynamic behavior of aimed structure as structuralsafety management tool for dismantling work of nuclear reactor facilities.
2. Structural monitoring system
The monitoring system that we introduce, called OSMOS system, is already widely used worldwide inparticular in Europe. Figure 1 and Table 1 show the outline of its system: the data measured by the sensors is
unified, treated and shared in the Cloud, from which the user can obtain any piece of information in the form ofprocessed data such as graphs or frequency analysis. There are two types of sensors: one is autonomous wirelesssensor in which all the functions, such as data measurement, collection and communication are integrated. The
other is cable type in which several sensors are independently connected to the data acquisition unit. Both types
have an alert function: once the preset threshold value exceeded, an alert is sent to the preselected recipients by
mail or SMS with the measured value. The sensors of this system can be installed in a short time. The wirelesstype sensor is particularly suitable for monitoring of structure in an irradiated environment where the working
time is limited.

Fig-1 Monitoring system

2m or 1m4mm(2m),3mm(1m)50Hz-10℃to40℃
0.02mm1.5to2years
Static and dynamic+/- 15 degree
0.009 degree
10Hz-10℃to40℃
Battery drive
minimum 1year

Measurement

StaticWired
Optical Strand

LengthMeasurement rage
Sampling frequency
Temperature renge
Resolution
Battery life
Measurement
Measurement rage
Resolution
Sampling frequency
Temperature renge
Acquistion unit
Battery life

Data acquistion unit

Wireless
Tiltmeter

Wireless
Optical Strand

Table-1 Main characteristics for sensor and system
LengthMeasurement rage
Sampling frequency
Temperature renge
Resolution
Measurement

2m or 1m4mm(2m),3mm(1m)
100Hz-20℃to60℃
0.02mmStatic and dynamic

Ambient temperature
CaminetSampling frequency
Temperature renge
Measurement
power supply-10℃to52℃
unsert 19",in height
100Hz-10℃to52℃
Static and dynamic
100V or 260v ca
Op.sentor:12
Non Op.senor:18

Connectable sensors

References[1] https://www.osmos-group.com/sites/default/files/safeworks/OSMOS-General-documentation-2018.pdf[2] http://committees.jsce.or.jp/opcet/system/files/monitoring_interim_report_final_1.pdf

29

D01

Development of New Radioisotopes Selective
Adsorption Media in EBARA
Takashi Sakuma1, Makoto Komatsu1, Tatsuya Deguchi1,
Kaoru Kikuchi1, and Takeshi Izumi1
1

EBARA Corporation

Abstract
EBARA has been developing radioisotopes selective adsorption media for treatment of contaminated water
since 2011. During this period, 900 kinds of adsorption media were supplied from some manufacturers andthe radioisotope removal efficiency was tested. These media have been successfully used at NPS in Japanto process radioactive contaminated water and contribute the settlement of severe accident occurred in
March 2011. This paper introduces the media which has high removal efficiency of radioisotopes.

1. Radioisotopes Selective Adsorption Media
1-1. Advantages
These media have big advantages as follows, 1) large decontamination factor (DFs), 2) increases media
troughputs, 3) reduction of waste generation, 4) reduction of radiation exposure to operators by reducing
the frequency of media exchange.
1-2. Adsorption MediaEBARA media consist of organic or inorganic material. These media can remove Cs,Sr,I,Sb and Ru.
EBARA also developed the media which can remove Cs and Sr at the same time.
1-3. Test Results
The batch test results are shown in Figure 1 and 2.

Figure 1 Cs/Sr Removal efficiency underseawater condition

Figure 2 Cs/Sr Removal efficiency under diluted
seawater condition

2. Conclusion
EBARA media removes specific ions from contaminated water efficiently and economically, areavailable in several process to accommodate the needs of any project. We are going to develop theradioisotopes selective adsorption media for farther waste water treatment such as decommissioning.

30

D02

Development of a Treatment System for Radioactive Spent Ion
Exchange Resins
Takashi Kawakami, Takaeshi Izumi, Ishiyama Yuji and Hirofumi Inagawa
EBARA Corportion

Abstract
A new ion exchange resin solidification system, featuring the combining of resins with paraffin wax,
has been developed. The resulting solid matters, consisting of resins and paraffin wax, were able to beincinerated easily by our own test incinerator and the actual incinerator in nuclear power plants (NPP).
Therefore, there is no longer a need to store spent ion exchange resins, as low-level radioactive waste that
had been generated from water purification systems at NPP.
1. Introduction
In NPP, a great amount of bead and powdered ion exchange resins are widely used for condensate
demineralizers, condensate filters, reactor water clean-up systems, fuel pool clean-up systems, and radioactive
waste liquid treatment systems.

After usage, spent ion exchange resins are stored in NPP and the amount ofradioactive spent resins is gradually increasing. Due to the limited storage capacity, spent resin storage issue isbecoming one of the major concerns for stable plant operation. At some plants, spent ion exchange resins areincinerated with other burnable wastes such as paper wastes in existing incinerators. However, due to the effort
in reducing burnable wastes in the plants for rational purpose, the lack of the burnable wastes is causing somedifficulties to incinerate spent ion exchange resins. Consequently, in order to dispose spent ion exchange resins
continuously and stably, we have been developing a new solidification system by combining ion exchange resins
with paraffin wax. These solid matters were able to be incinerated easily in our small-scale test incinerators andthe actual incinerator in NPP.
2. The solidification system
Spent resins are mixed with paraffin wax, calcium hydroxide and stabilizing
additive for the continuous solidification with mixing equipment. After mixing, themixture is heated and extruded by extruder and cut into the suitable length, andproduced solid matters are collected in products tank. Finally, these solid matters areincinerated in existing incinerator.
3. Conclusion

1) Ion exchange resins were able to be solidified stably and continuously
combining with the paraffin wax by the addition of the stabilizing additive
using extruder.
2) The solid matters were able to be incinerated without burnable materials in small-scale test
incinerator and the actual one in NPP. The concentration of sulfur oxide in off gas was kept very
low by adding calcium hydroxide.
3) The amount of residual ash after the incineration was very small and the weight and volume
reduction ratio were both more than 95%.
References[1] H. Inagawa, EPRI 2009 Condensate Polishing Workshop, 61-80 (2009)

31

D03

Inhibition of Oxidative Decomposition of Ion-exchange Resins
Makoto Komatsu1, Tatsuya Deguchi1 and Takeshi Izumi1
1

Ebara Corporation

Abstract
Hydrogen peroxide (H2O2) generated by the radiolysis of water exists in nuclear power plants and it
accelerates the oxidation decomposition of the ion exchange resins[1], and finally, it shortens the resin life. The
application of Palladium (Pd) doped resins which can decompose H2O2 has been considered.
1. Introduction – Pd doped resins
Hydrogen peroxides can be easily decomposed by the catalytic reaction of platinum group metal such as
Pd. Pd doped resins are developed and manufactured by Lanxess Company and are strongly basic, gel-type,palladium-doped anion exchange resins.
2. Experimental
2-1. Decomposition behavior of hydrogen peroxide – Long-term column operating test
To evaluate the stability of the H2O2 (2 mg/l) decomposition performance by Pd doped resins, long term
column (inner diameter: 16 mm, bed height: 10 mm) operating test was carried out (flow rate: 3.5 ml/min,
LV=1 m/h). For 4 years in cold tests, Pd doped resins showed the stable H 2O2 decomposition performance.

Figure 1. Lifetime of Pd doped resins
2-2. TOC leaching behavior – Inhibition of oxidative decomposition of ion-exchange resins
In order to confirm the improvement effect of the ion-exchange resin degradation tendency by the Pd
doped resins, TOC leaching behavior evaluation tests were achieved by using a closed-loop test apparatus (innerdiameter: 25 mm, conventional resin bed height: 150
mm, LV=40 m/h, tank volume: 5 liters, H2O2: 5 mg/l).
Figure 2 shows the results of TOC leaching behavior
evaluation test. In case of the loop without Pd doped
resins,TOC

concentration

showedan

upwardtendency from 500 hours later. On the other hand, incase of the loop with Pd doped resins (26% overlay),
the TOC concentration little rose and it was less than
1mg/l at 2500 hours later.

Figure 2. TOC leaching behavior

3. Conclusion
It is expected that Pd doped resins contribute to the extension of the resin life and the reduction inradioactive waste generation.
References[1] J. R. Stahlbush, R. M. Strom, Reactive Polymers, 13, 233-240 (1990).

32

D04

Alpha nuclide absorbent, TANNIX
Kazuhiko Hamaguchi and Hideyuki Teshima
Mitsubishi Nuclear Fuel Co.,Ltd. (MNF)

Abstract
MNF developed insoluble tannin absorbent, TANNIX, and it has been successfully applied for alphanuclide absorption. TANNIX at a nuclear power plant and a variety of nuclear facilities. TANNIX has shown
its excellent adsorption capacity and distribution coefficient (Kd) of well above 103 against alpha nuclides. It
can contribute to Fukushima-Daiichi decommission when alpha nuclide needs to be removed from the coolingwater.

1. Introduction
MNF developed insoluble tannin absorbent, TANNIX, originally to remove Uranium from waste water of
MNF Re-conversion facility processing from UF6 to UO2 powder. TANNIX has processed more than 100,000
m3 at MNF by using industry size unit. The major features of TANNIX are 1) its high absorption rate for alphanuclides and 2) almost “0” secondary waste volume after burning.
TANNIX has been used at Paks-2 NPP and serval facilities in Japan such as JAEA plutonium fuelfabrication facility [1]. In this paper, TANNIX performance at Paks-2 NPP is summarized because it would be
useful to evaluate its ability to Fukushima-Daiichi cooling water.

2. Joint research and result at Paks-2 in Hungary
During the chemical cleaning of corrosion products on the surface of fuel cladding in Paks-2 VVER-440
reactor overheated to severely damaged [2] . The waste water including some radionuclides and about 20g/dm3
boric acid (pH=4.1) was generated.
Budapest University of Technology (BUTE) and MNF jointly researched to treat the waste water.
BUTE examined equilibrium test with real radioactive waste.

1.E+01including at Paks-2 NPP processing the water with nuclear fuel and

Cm242

Cm244

Pu239+Pu240

1.E+00Pu238+Am241

TANNIX, developed by MNF, has been used at several facilities

1.E+02Ce141

3. Conclusion

1.E+03

Ce144

successfully disposed.

1.E+04

Cs137

10 . After testing, TANNIX was used and the waste water was

1.E+05Cs134

3

1.E+06
Distribution coefficent :Kd(cm3/g)

The distribution coefficient (Kd) of TRU element was well above

Kd Databoric acid.

MNF thinks that TANNIX can be one of the strong candidates of alpha nuclide absorbent for
Fukushima-Daiichi cooling water. It is desirable to confirm the applicability of TANNIX to the cooling water.

References[1] Science and Technology In Japan, Vol.16 No.62 1997.
[2] Inter National Journal of Nuclear Energy Science and Technology, Vol.2, No.4 2006

33

D05

Study on Application of Laser-Decontamination in Decommissioning
Masato Watanabe1,Hiromitsu Inagaki1, Kazuhisa Fujita2,
Shin-icihro Okihara2,Kazuaki Toyosawa3 and Nobumitsu Maebashi3
1

CHUBU Electric Power Co., Inc.,

2

The Graduate School for the Creation of New Photonics Industries,
3

Toyokoh Co., Ltd.

Abstract
Laser decontamination is applicable to the decommissioning of nuclear power plants as a decontamination
technology, because secondary waste is expected to be considerably reduced.
1. Introduction
In the field of civil engineering and construction, portable laser
paint removal devices (Fig. 1) for steel bridges may be applicable asradioactive

decontamination

technology

fordecommissioning

measures.
We undertook a study on the feasibility of application of laser
decontamination in decommissioning, with confirmation tests on theFig. 1 Portable laser paintremoval device

removal rate of the stainless steel surface and on the dust debris(primary waste) recovery apparatus.

6.E+08

2. Confirmation test of decontamination speed

5.E+08To estimate the laser decontamination processing speed,
we irradiated the surface of steel materials using a 500 W fiber
laser, and measured the amount of stainless steel surface
material removed. (Fig. 2)

voume(um^3)/s<--Optimum condition

4.E+0825mm/s50mm/s100mm/s

3.E+08

Scan Speed
25 mm/s
50 mm/s
100 mm/s

2.E+081.E+08Under optimal irradiation conditions, when estimating

Focal Position
0.E+00the processing speed for a thickness of 10 μm, the removal is 5

0

2,000

4,000

6,000
8,000
Fluence (J/cm^2)

10,00012,000Fig. 2 Laser fluence dependency of
surface volume removed

hours per m² with the 500 W laser.
3. Confirmation test of the debris collection apparatus

Even with laser decontamination, it is necessary to collect the decontaminated radioactive debris. In orderto prevent dust scattering, we realized a decontamination method in water (Fig. 3).
The mechanism for exhaust and collection of the debris in water isprovided at the tip of the laser irradiation part, and the removed debris
is recovered together with water and removed with a filter.
4. Results and future work
This study revealed that laser decontamination is applicable todecommissioning as a decontamination technology.
Presently, we are working on the selection of decontamination

Fig. 3 Prototype underwater laser
decontamination device

targets (systems, equipment, and parts) and the specification ofdevices appropriate for this, which will become the key to practical application.

34

D06

Sorting Technology for Efficient and Safe Soil Processing
Steve Rima1, Jeffrey Lively1, Masaru Noda2, and Haru Hashizume2
Wood1 and Obayashi Corporation2

Abstract
Obayashi and Wood have been working on the application of sorting technology for radioactively
contaminated soil after Fukushima incident. That includes the adaptation and upgrade of well-proven systemin the US to the scopes and specifications for Fukushima offsite cleanup. Its adaptability is further expected toextend into onsite work demands in any challenges. This paper with oral presentation was made at American
Nuclear Society Joint Topical Meeting on Remote Control and Robotics in Pittsburgh in August 2016.
1. Conceptual ISF Process Train Design
The sheer volume of solid radiologically impacted materials that need to be processed through the ISFs
challenges capacity and throughput of the existing, ready-to-deploy ScanSortSM system. The conceptual designfor the ISF process train includes processes other than simply assaying and sorting for radioactive content. An
appropriate scale for the process train needed to be determined and the electro-mechanical function of theprocess train needed to be engineered as the ScanSortSM system could not function as a “stand-alone” process.
The Obayashi teams design for the ISF process train considered a number of factors to arrive at the design
specification. It was determined that a network of nine independent and parallel process lines would be needed
to meet the required productivity of 180 cubic meters per hour. Some of the constraining parameters included
the limitations for material handling and feeding of 1 cubic meter bags, the desire for built in redundancy andflexibility of scale, as well as the expected volume of material and the proposed time line for ISF operations.
2. Generation II ScanSortSM Design
In order to achieve the desired endpoint of a single point of command and control from which an entire
process train’s operation would be controlled, Wood has need to reengineer the ScanSortSM system. The
reengineering efforts fall into two major categories:• Decoupling of the sensor systems from the electrical power distribution system, the I&C system, andthe electro-mechanical control components, and• Redesign of the ScanSortSM operating system software to support the decoupling of the I&C systemand the redistribution of command and control functions to non-dedicated electro-mechanical modules
Wood has completed the redesign of the ScanSortSM operating system software. In addition to the changes
necessitated by the decoupling of sensors system, the redesigned software will provide for the distribution andremote display of real-time system information to a centralized command and control station.
Obayashi and Wood are collaborating on the specifications for the engineering effort required to produce
new standardized interfaces for electrical power, I&C signals, and modular mechanical attachment of the
ScanSortSM system’s components to the mechanical systems sourced completely in Japan.
3. Modification to Process Heavy Clay-Like Soils
Obayashi has previously designed a conveyor, called an “extruder,” to enable efficient processing andsorting of heavy, clay-like materials. The extruder doesn’t simply place material onto the Survey Conveyor belt
like the original ScanSortSM system, but “squeezes” it out onto a consistent layer on the belt using a roller. For aproject in the U.S. with this type of material, Wood has built and deployed an extruder based on Obayashi’s
design. The extruder is in current use and has been proven to handle and sort very wet, heavy, clay-like soils
with no clogging or other issues that have been experienced by other sorting systems. The addition of an
extruder to a ScanSortSM system for use in Japan will enable sorting of soils in Japan to be done efficiently while
preventing the problems encountered by other sorting systems.

35

D07

Assessment of Future Countermeasures against Contaminated
Groundwater in Fukushima Daiichi Nuclear Power Station
Tadafumi Koyama1, Yasuharu Tanaka1, Daisuke Tsumune1, Takatoshi Hijikata1
Mamoru Kamoshida2, Takeki Yamauchi2, Masaki Imazu2 and Toshihiko Fukuda2
1

CRIEPI, 2NDF

Abstract
Analytical study to evaluate radiological risk, residual contamination and amount of major consumptions was
carried out for comparing future countermeasures against contaminated groundwater in Fukushima Daiichi.

1. Introduction
The highly contaminated water in turbine buildings is stably confined by countermeasures, and will be dried-up
completely before fuel-debris retrieval. The groundwater around reactor buildings is far less contaminated, andprotected by the same countermeasures, however, it is important to evaluate the future radiological risk and assess
appropriate countermeasures because groundwater will remain even after debris retrieval.
2. Analytical method
Future distributions of H-3, Sr-90 and Cs-137 in groundwater and soil around reactor buildings were calculatedwith new 2-D countermeasure model applied to FEGM code.[1] Public radiation dose by ingestion of sea foods
was calculated from radioactivity permeated through sea-side

Table 1. Effectiveness comparison

impermeable barrier. Consumption of adsorbers (assumed
properties of IE-96/A-51) to clean pumped groundwater was

3. Results and Discussions
Analysis was carried out for the future with reference
countermeasure①that consists of subdrain and sea-side barrier
after debris-retrieval. The obtained radiation dose, as low as
1E-5 mSv/y, showed the effectiveness of this measure, while
adsorber consumption is huge as shown in ① of Figure 1.
Hence, comparative study was carried out for estimating
possible performance of alternative countermeasures: ②
subdrain, frozen soil barrier and sea-side barrier, ③sea-sidebarrier with permeable reactive barrier to trap Sr, ④subdrain,
adaptive pump&treat for known plumes and impermeable

1600
1400
consumption of adsorber/arbit. unit

estimated from radioactivity transported to subdrain pumps.

( → same ↘ down ↓much down )countermeasures
①

1200

②
③

1000

④

800

⑤
⑥

600
400
200
0
2000

2020

2040
Year

2060

2080

Figure 1. Cumulative adsorber consumption

barriers, ⑤subdrain, soil flushing and impermeable barriers, ⑥subdrain, injection of chemicals to immobilize
Sr-90 and impermeable barriers. All countermeasures were found to keep public radiation dose as low as 1E-5
mSv/y, while reduction of soil contamination or consumption of adsorbers depend on each measure. As shown in
Table 1, ④ and ⑤ are the most effective measures to reduce soil contamination. On the other hand, total
consumption of adsorbers can be drastically decreased to 1/6 of ① by ③ or 1/3 by ② as seen in Figure 1.
References: [1] T. Koyama et al.,” Environmental risk analysis for radionuclides transported in groundwater of Fukushima
daiichi”, poster K02, 2nd Int’l Forum on the Decommissioning of Fukushima Daiichi NPS, July 3, 2017.

36

E01

Construction of Unit-3 Cover Structure for Fuel Removal
Shinya Okada, Kenichiro Go, Norio Ryoki, Taimei Hida,
Miho Miyazaki, Ippei Matsuo, Kihei Ogawa
Kajima Corporation

Abstract
We have successfully decreased the radiation dose rates by way of rubble removal, decontamination andradiation shielding that we incorporated ten remote control technology and we have completed construction
work of Unit-3 cover structure for fuel removal in February, 2018. This paper summarizes the challenges andaccomplishments of the project over seven years.
1. Reducing the radiation exposure of workers
 We have developed remote operation system for heavy equipment. Ten pieces of heavy equipment wereoperated at the same time controlled from an operation room located approximately 500m away.
 We have also developed a labor-saving joint of steel column members for assembling the base of the cover
structure.
2. Removal steel frame rubble collapsed due to hydrogen explosion
 The modeling of the steel flame rubble was performed using 3D leaser scan, and the behavior of steel
frame rubble during dismantling was investigated using finite element analysis.
 Various crane attachment to cut and grip steel frames by remote operation were developed.
3. Reduction of dose rates for manned work indispensable for cover structure construction
 Various remote control devices for rubble removal and decontamination were developed for the various
floor damage and contamination condition.
 The shielding steel plates were installed with the maximum weight within supporting capability of eachfloor by remote control.
4. Design of cover structure The dorm-shape steel frame were employed for the cover structure in order to secure a large space with arelatively lighter weight.
 The vertical and horizontal loads to the reactor building were reduced as much as possible.
 A labor-saving joint was developed for steel column members in order to reduce manned activities.
5. Safe and efficient manned works on the operating floor
 Amount of manned activities at 1F site were reduced by practicing fabrication outside 1F site.
 Practicing fabrication of the cover structure outside 1F site could eliminate problems in advance andincrease the worker’s learning to a higher level.
6. Reducing risk of damaging fuel
 The spent fuel pool was covered with protection covers not to damage fuels during rubble removal.
 Cover structure units were set in place by sliding them to avoid construction work over the spent fuel pool.

BeforeRubble Removal

After
Rubble Removal

Radiation shielding
plate installation

37

Fuel handling machine
girder installation

Completion of cover
structure

E02

Land-side Impermeable “Frozen-Soil” Wall
Atsuhiro Fukada, Isao Abe, Taichi Esaki, Shoji Seno,
Teru Yoshida and Torahito Suzuki
Kajima Corporation

Abstract
The land-side Impermeable Wall with the method of artificial frozen soil, the one of the longest frozen soil
wall ever made, was constructed around the damaged reactor buildings, as a measure against the increase ofcontaminated water. It had been serious issue that the major volume of groundwater had become thecontaminated water due to mixing up with the radioactive contamination in the damaged reactor buildings. The
Frozen-Wall cuts off this groundwater flow toward those reactor buildings, and prevents from the further
increase of contaminated water.

1. Background
Since the accident, the inflow of the groundwater into the reactor
buildings was daily estimated up to 400 m3/day, and it had been

Ground Water

required to shut off and reduce this major volume of the flow. To copewith this issue, the Committee on Countermeasures for Contaminated
Water Treatment, organized by METI, decided a basic policy in May,

Frozen-soil Walls

2013, i.e. the installation of the Frozen-soil Walls.
Figure 1. Concept of the
Frozen-soil Wall
The construction process to create seamless Frozen-soil wall consists of two phases; 1) the installation of

2. Construction

the required facilities and 2) the freezing operation. The facilities-installation was commenced in December
2013. One of the most important facilities in-situ are 1,568 vertical freezing pipes installed into the ground,approximate 30m deep. The following phase for freezing soil has been operated since March 2015, to circulate
brine solution through the freezing pipes at -30 degree Celsius. In addition, Grouting method was also
complementally and partially applied to advance the development of the Frozen Wall, reducing the velocity ofgroundwater flow. The Committee and TEPCO announced the completion of the Frozen-Wall in March 2018.

3. Result and EffectIt is recognized that the ground water level inside of the Wall is 5m lower than the outside. Furthermore,
the growing-amount of the contaminated water is reduced to 140m3/day, compared with the amount before closeof the Frozen-soil wall, 520m3/day. According to these results, The Committee evaluated that the Frozen Walleffectively shut off the groundwater flow and prevent from the increase of contaminated water.

References[1] http://www.meti.go.jp/english/earthquake/nuclear/decommissioning/index.html[2] http://www.tepco.co.jp/en/decommision/planaction/landwardwall/index-e.html[3] http://www.kajima.co.jp/tech/c_frozen_soil_wall/index.html (in Japanese)

38

E03

Technologies for dismantling radioactive contaminated water
storage tanks at Fukushima Daiichi Nuclear Power Station
Kazushige Kato1, Masahiro Nakamura1, Haruo Nagamine1,
Naoki Maeshiro2, Hiroshi Nakamura3, and Yoshihito Nakahara4
1

TAISEI CORPORATION, 2Tokyo Electric Power Company Holdings, Inc.,
3

TOYO UNION Company Limited, 4TAIYO KOGYO CORPORATION

Abstract
The developing technologies for controlling radioactive dust, when the radioactive contaminated water
storage tanks are dismantled at Fukushima Daiichi Nuclear Power Station (1F) are described in this paper.

1. Introduction
The bolted-type tanks for storing radioactive contaminated water constructed at the early stage ofdecommissioning of 1F have been replaced by the welded-type tanks to avoid water leakage due to aging
degradation of sealing material. Since radioactive dust is adhered to the inside of tanks, reducing scattering to
the environment and worker’s radiation exposure are highly demanded during the dismantling the tanks.

2. Technologies for dismantling bolted-type tank2-1. Remote spraying device for controlling dust
Before dismantling a tank, the device is inserted from a top of tank toa bottom and the two spray nozzle arms are expanded to right and left. The
viscous liquid for retarding dust discharge is sprayed from the nozzles on
the inside walls of tank by rotating about the center rod of device from

Preparation work
Spraying
a dustscatter
retarderSpraying
liquidon thewalls

bottom to top of tank by remote control (see Figures-1 and 2).

2-2. Balloon cover for reduce scattering of dust
During dismantling a tank, the balloon is covered at a top of tank after

Dismantlinga flange tanktank
Dismantlinga bolted-type

daily work in order to reduce scattering of dust as well as to prevent

Balloon coveradmitting rain into a tank (see Figure-1 and 3). Since installation andde-installation of cover is daily work, it is crucial to make working hours
shorten. The developed balloon cover is frameless and is light enough toinstall by small crane in a short time. The cover is removable by tightening

Covering/transportation

belts of cover with mounting brackets on a top of side wall at about one

Figure-1 Tank dismantling flow

meter interval. Thus, the cycle time of the work is to be about 40 minutes.

3. Conclusion
The availability of these technologies was
confirmed at the site, and the worker’s radiation
exposure was kept minimum in consequence.
Figure-2 Spraying white liquid

39

Figure-3 Balloon cover

E04

Dismantling Plan of Exhaust Stack for Units 1 and 2

Hiroshi Noda1 ,,Yoshihisa Michizuka1, Takashi Okutsu1
1
Tokyo Electric Power Company Holdings, Inc.

Abstract
Damage has been confirmed in some parts of the exhaust stack by inspection after the earthquake.
Currently the exhaust stack has no function. For this reason, we decided to dismantle the upper half of
exhaust stack for the purpose of improving the earthquake resistance. Emphasizing on reducing worker
exposure dose, we plan dismantling work with unmanned operation using a remote dismantling device.
1. Introduction
The exhaust stack consists of a rectangular steel tower and a cylinder stack which is supported by thesteel tower. Height of the cylinder stack is 120 m and inside diameter of it is 3.2 m.
The steel tower part mainly consists of main pillars, diagonal braces and horizontal parts. We plan topartially dismantle the exhaust stack. The dismantling range is from around 59 m to 120 m above theground as the diagonal nine braces around 66 m and 45 m above the ground are damaged.
2. Exhaust Stack Dismantling Plan
We plan to dismantle the stack and steel tower on a block-by-block basis using a large crane which isused in fuel removal works. Saving labor is taken in by using dismantling device which has cutting andholding functions.
First the stack which is protruding is dismantled and then sequentially the dismantling of steel tower andthe stack is repeated.

Figure 1. Outline of dismantling step
3. Remote dismantling device
Remote dismantling device is held and fixed with two kinds ofholding devices after inserting the disassembling tool into the cylinder
stack. The cylinder stack is cut with a chip saw from the inside of thecylinder.
Accessories (such as ladders) outside the cylinder stack thatinterfere at the time of cutting the cylinder stack are removed with ahexaxial arm robot.
In order to prevent the radioactive dust to scatter when cutting the
Fugure 2 dismantling device image
cylinder stack, fixing agent sprayed inside the stack with another device.

40

Improvement in work environment at Fukushima Daiichi NPS
E05 YutaSuzuki, Toshiharu Matsuzawa, Shiro Takahira, Koji Hayashi
Tokyo Electric Power Company Holdings, Inc.

Abstract
For workers at Fukushima Daiichi NPS (1F), we have been engaging in the improvement in work
environment, light-duty protective clothing, and the reduction of radiation exposure.
1. Introduction
Because the radiation environment at Fukushima Daiichi NPS was extremely changed by its nuclear
accident, it was necessary to improve the working environment in order to steadily advance the long-term
decommissioning work. Here we would like to introduce how we work on such improvements.
2. Improvement of work environment (dose reduction)
In order to reduce the exposure of workers, we have decontaminated by asphalt pavement, stripping
surface soil, shielding, etc. By the end of FY 2015, we could achieve the targeted dose rate of 5 μSv/h* by suchdecontamination. (Fig. 1).
* The targeted dose rate is set to average of 5 μSv/h at chest position. The Evaluation of dose rates at groundsurface level was done by collimator in case the place was affected by direct rays from the plants.
3. Light-duty Protective clothing and finer zoning
In order to keep the decontaminated area at 1F site clean and to reduce the protective gear burden on
workers (ex. changed from coveralls to normal work clothes), the controlled area must be divided into somezones according to contamination levels. And it must also carefully controlled to prevent from spreading thecontamination from highly contaminated areas, such as inside the Unit 1~4 buildings and the tank dismantlingarea.
Currently, we can work with normal working clothes and disposable dust mask at about 96% of 1F (Fig. 2).
4. Conclusion
While grasping the radiation environment after the accident, we have taken appropriate measures forradiation protection, and have been continuously improving radiation environments and optimizing radiation
management. In order to steadily proceed the decommissioning, we will keep on making our best efforts toimprove work environment, light-duty protective clothing and reduce radiation exposure.

Figure 1 History of dose reduction

Figure 2 Map after finer zoning

41

H01

Present Status on R&D of Hydrogen Concentration Reduction
Technology in Waste Storage Containers with PAR
Kazuyuki Takase1, Hisayuki Suematsu1, Hitoshi Sugiyama2, Hirohisa Tanaka3,
Masashi Taniguchi4, Isamu Kudo5, Ryutaro Hino6, Atsushi Terada6,
Daijyu Muramatsua6 and Yasuo Nishihata6
1

Nagaoka Univ. of Technology, 2Utsunomiya Univ., 3Kwansei Gakuin Univ.,

4

DAIHATSU Co. Ltd., 5ADVAN ENG. Co., Ltd., 6Japan Atomic Energy Agency

Abstract
In the decommissioning of Fukushima Daiichi nuclear power plant (1F), hydrogen which is the flammablegas is generated in a long-term waste storage containers by decomposition of water due to radiation. Then, R&D
on reduction of hydrogen concentration using a storage container with catalysts has been begun.

1. Introduction
In the decommissioning of 1F, the long-term management of fuel debris is necessary. Hydrogen is
gradually accumulated in the storage container with time. To reduce the hydrogen concentration, R&D on thelong-term radioactive waste storage container with a passive autocatalytic recombiner (PAR) has been begun.
The objective is to confirm experimentally and analytically effectiveness of the storage container with PAR.
Kwansei Gakuin University

2. Present Status of R&D

・Catalyst designand evaluation

The R&D is performed by six organizations
since 2017 and Nagaoka University of Technology

Daihatsu Motor Co.,Ltd.
・Prototyping and improvementof monolith-type catalyst

Nagaoka Universityof Technology

CVis a main organization. Research topics are shown in

Advan Eng.Co.,Ltd.

・Verification experimentand data analysis

Fig. 1. Until now, two kinds of prototyped catalysts
were developed: the monolith type catalyst made of・Development of new
catalysts made fromlow cost

Japan Atomic Energy Agency

ceramics; and, the geopolymer catalyst based on・Construction of catalytic
reaction model・Catalytic reaction experimentwith SPring-8

geopolymer, alumina and precious metals. The

Utsunomiya University
・Modeling and analysis ofnatural convection behavior

catalytic reaction experiment was conducted forFig. 1each catalyst, and it was confirmed that it has a good

Research topics of each organization
Without geopolymerhydrogen recombination reaction.

With geopolymer

Figure 2 shows a preliminary result on hydrogen reduction in asmall-scale vessel with PAR. This vessel simply simulates a storage
container. The number of particulate geopolymer is a parameter. The
reduction of hydrogen concentration depended on the contact area of
geopolymer catalysts with which hydrogen contacts. Moreover, aConcentration of hydrogen (%)

2

simulation method and numerical models are also being developed.

20
40

1.5

1

0.5

0
-5

0

5

10

Time (s)(s)

Fig. 2
3. Summary

Numbergeopolymer
Number ofof geopolymer

Change of hydrogen concentration
in a small-scale vessel with time

It was found that storage containers with PAR are effective for reducing hydrogen concentration. In the
future the present result will be confirmed under a wide range of experimental and analytical conditions.

42

H02

Evaluation of Performance on Hydrogen Concentration Reduction
in Storage Containers with PAR Wetted by Water
Kazuyuki Takase1, Matsmoto Yoshinobu1, Isamu Kudo2, Shogo Taira2,
Keisuke Tanizawa2, Gaku Takase1 and Yoshihisa Hiraki1
1

Nagaoka Univ. of Technology, 2ADVAN ENG. Co., Ltd.

Abstract
In the decommissioning of nuclear reactors, hydrogen which is the flammable gas is generated in storage
containers by decomposition of water due to radiation. Then, in order to reduce the hydrogen concentration in along-term waste storage container, use of passive autocatalytic recombiners (PAR) [1] was considered.

1. Introduction
160 mmThe objective of this study is to confirm experimentally that PARfunctions effectively, even if a part of the PAR is the wet condition.

2. Evaluation of PAR on Hydrogen Concentration Reduction
In this experiment, the condition that much water is mixed with

500
mm

radioactive waste into a container was assumed. Figure 1 shows anexperimental apparatus which is a cylindrical container with a diameter of 160
mm and length of 500 mm. Particulate formed geopolymer catalysts are
installed in the cylindrical container. The geopolymer catalyst made by
ADVAN ENG Co. Ltd. consists of a geopolymer as the base material, alumina
as the support material and platinum as the catalyst metal. Hydrogen touches
platinum and combines with oxygen, and then water is generated.

Fig. 1 A cylindrical container

As one of preliminary experimental results, time change ofhydrogen concentration in the cylindrical container is shown in

1

Fig. 3. Initially, the inside of the cylindrical container is filledwith air with the hydrogen concentration of 1%. Geopolymer
catalysts were set in the container on the condition which got

0.9

wet in water. The hydrogenconcentration decreases with
0.8

time, and it reaches around0.75 at 200 minutes.

3. Conclusion

0.7
0

The performance of thewet conditions was confirmed.

Fig. 2

References

100

150

200

Time (min)10 mmgeopolymer catalyst under

50

Particulate formed
geopolymer catalysts[1] J. Henrie, et al., GEND 051, 1985.

43

Fig. 3 Time change of hydrogen concentrationin the cylindrical container with the geopolymer
catalysts which were set under wet condition
initially

H03

Hydrogen safety technology utilizing the automotive catalyst
T. Kita1 , H. Ono1, K. Takenaka1, S. Kishimoto1, M. Taniguchi2, D. Matsumura3 ,Y. Nishihata3, R. Hino3, E-A. Reinecke4, K. Takase5, H. Tanaka1

1. Kwansei Gakuin University, 2. Daihatsu Motor Co., Ltd. 3. Japan Atomic Energy Agency,
4. Forschungszentrum Jülich GmbH, 5. Nagaoka University of Technology
Abstract
We are aiming to establish hydrogen safety technology. It is a serious problem that water is decomposed bystrong radiation to generate hydrogen in a container for long-term storage of radioactive waste collected bydecontamination, as well as decommissioning nuclear reactors. An intelligent automotive catalyst showed
excellent activity to recombine hydrogen and oxygen without external power supply. Moreover, it is found thatthe monolithic catalyst oxidized a hydrogen from -25 °C.
1. Introduction
It is necessary to store radioactive waste related to decontamination and decommissioning reactor safely andlong term. In the container of nuclear radioactive waste isolated, water is decomposed by radiation to producehydrogen. The passive autocatalytic recombiner (PAR) which needs no electric power supply and heating
sources has been developed. We are aiming to apply the intelligent automotive catalyst to PAR.

H2 conversion (%)

100
2. Research and Development
2.1. Catalyst preparation
80
Three types of monolith catalysts (900, 100 and 30 cpsi) wereprepared using a mass-produced automotive “Intelligent
60
Catalyst”. The catalysts consisted of Pt, Pd and Rh-perovskites.
40
2.2. Laboratory scale catalytic reaction (forced flow)
30 cpsi
100 cpsi
The catalytic activities were observed from -25 °C (Fig. 1).
20
900 cpsi
Higher density cell catalyst showed better activity. When the0
activity of the catalyst arose, the temperature rose, with no-50
050
electric power supply and heating sources, due to hydrogen
Temperature ()oxidation and the hydrogen conversion efficiency exceeded 90%
Fig. 1. Cryogenic catalytic activity.
rapidly.
2.3. Large scale catalytic reaction (natural convection)
One piece of the monolithic catalyst (Φ70 x L10 mm = 38.5 mL), shown in Table. 1, was set in REKO-4
reactor (5,330 L) and exposed the mixture gas (6%-H2, Balance-Air) at a room temperature (Fig. 2). Hydrogen
reaction and natural convection velocity were improved 3 times by the use of a chimney. When the cell density
decreased, the amount of hydrogen treatment doubled. In the cell structure with small gas flow resistance,
natural convection and the hydrogen oxygen recombination reaction accelerated.

Table. 1. Experimental conditions.
Catalyst

Condition

H2 throughput [g/h]30 cpsi

w/o Chimney

4.16

30 cpsiwith Chimney

11.7

100 cpsiwith Chimney

5.79

3. Conclusion
The intelligent automotive catalyst showed high hydrogen and oxygen
recombination activity.
・ In the forced flow, the catalytic activity increased with increasing celldensity.
Fig. 2. REKO-4 reactor.
・ The catalytic activities were observed from -25 °C under the forced flow.
・ Quite contrary, in natural convection, the catalytic activity increased as the cell density decreased.
・ In a container for storing radioactive waste, low density cell catalyst showed better activity.
Our laboratory continues the study more in order to apply our catalyst into the practical PAR.
Acknowledgement
A part of this study is the result of “R&D on technology for reducing concentration of flammable gases generated inlong-term waste storage containers” carried out under the “Center of World Intelligence Project for Nuclear Science andTechnology and Human Resource Development” by the Ministry of Education, Culture, Sports, Science and Technology ofJapan (MEXT).

44

K01

Challenge for the Decommissioning of Fukushima based on theDeveloped Technologies
Susumu Tosha1, Shohei Sato1, Masaki Jimba1 and Toshiki Fukui1
1

IHI Corporation

Abstract
To contribute to the decommissioning of Fukushima Daiichi, IHI develops the technologies ofdecontamination, remote dismantling and vitirification based on 60 year experience in the nuclear field.
1. Introduction
IHI achieves the experience in the nuclear field over 60 years. IHI has supplied not only the primaryequipment such as the reactor pressure vessel and contaminant vessel, but also the construction and maintenanceto the nuclear power plant. In the nuclear fuel cycle and radwaste management field, IHI supplied the service ofdesign, fabrication, installation, and trial operation. Based on these experiences, the technologies ofdecontamination, remote dismantling and vitirification are being developed for the contribution to thedecommissioning of Fukushima Daiichi.
2. Experiences in the Nuclear Field
IHI accumulated the experiences of the design, fabrication, installation and maintenance for the primaryequipment of the nuclear power plant through the supply of reactor pressure vessel and contaminant vessel tothe BWR all over the world including the Fukushima Daiichi unit 1 to 3. IHI is Tier-1 contractor of the highly
active liquid waste vitrificaiton facility and vitrified waste storage facilities in Rokkasho Reprocessing Plant.
Due to extremely high dose rate of vitrification cell, all works inside the cell, such as operation, maintenance,
repair, replacement must be performed by remote handling. Based on these experiences, IHI could contribute to
the decommissioning of Fukushima Daiichi.
3. Technical Development
IHI develops the state-of-art technologies based on the experience indicated in the chapter two in this paper.
For the decontamination technology, Ultra-High Pressure Liquid Nitrogen technology “NitroJet®” which caneliminates the cross contamination and secondary waste streams are developed. For the remote dismantlingsystem, the laser cutting technology and remote handling simulation technology with 3D-CAD, which were
installed in the dismantling facilities for vitrification melter in Rokkasho Reprocessing Facility are developed toapply to the decommissioning of Fukushima Daiichi. The cold crucible induction melter technologies are also
developed to stabilize the variety of radioactive waste including organic substances based on the experience ofthe development of Liquid Fed Ceramic Melter in the Rokkasho Reprocessing Facility.
4. Conclusion
IHI develops the state-of-art technologies of decontamination, remote dismantling and vitirification basedon the 60 year experience in the nuclear field to contribute for the earliest possible decommissioning ofFukushima Daiichi.

45

K02

Basic Research Programs of Vitrification Technology for Waste
Volume Reduction (1)Development of Vitrification Technology for
Low Level Radioactive Waste(LLW)
Yasutomo Tajiri1, Oniki Toshiro1 Toshiaki Kakihara and Toyonobu Nabemoto1
1

IHI Corporation

Abstract
The vitrification tests for the LLW generated from nuclear power plant and reprocessing facility were
performed. The simulated wastes of resin, ash and nitrate sodium liquid were vitrified in the alumina crucible,
and the homogeneity and durability of vitrified glass was confirmed. Furthermore, the continuous vitrification
with the simulated wastes which simulated the physical forms of above wastes was tested in the small furnace,and evaluated the molten condition, environmental applicability of off gas and volume reduction capability.
1. Introduction
The vitrificaion technology is one of the processing methods of radioactive waste for stable disposal. To
confirm the applicability of vitrification technology to LLW, the stability of vitrified glass and vitirification
operation must be confirmed. In this research, the homogeneity and stability of glass were evaluated by the testin the crucible. Furthermore, the molten glass condition and off gas of the simulated wastes were evaluated bycontinuous vitrification in the furnace.
2. Evaluation of glasses
Firstly, the glass composition of resin, ash and nitrate sodium liquid were set
based on the glass database, the state diagrams of slags and so on, and the additives

Fig.1. produced glassof each wastes for vitrification were calculated based on the glass composition of each waste. Secondly, thesimulated waste and its additives were melted in the crucible at 1100 degrees Celsius(Fig.1). As a result of
vitrificaion, the homogeneity was confirmed for each glass according to the SEM. The stable results wereobtained according to the PCT-leaching test. The glass compositions that didn’t satisfy the evaluations wereregulated again. The PCT-leaching rate of the sodium silicate glass of nitrate sodium liquid was improved byadding Al or Ti to glass. This mechanism has been investigated with some universities.
3. Evaluation of operations
To evaluate the stable operation, the simulated wastes which simulated the physical forms were vitrified
continuously in the furnace. As a result, the stable operations were confirmed by observing molten conditions
that the simulated wastes were mixed in glasses continuously(Fig.2). Also, the high volume reduction wasobserved for resin glass compared to cement solidification. This is because themolten glass

resin was decomposed in the process of vitrification. On the other hand, the 96%
of sulfur components of resin was volatilized. It is assumed that the treatment
system is necessary to remove gas of sulfur in the vitrification process of resin.

resin

Fig.2. Molten condition

4. Conclusion

The applicability of vitrification for resin, ash and nitrate sodium liquid was confirmed by the vitrification
tests. The electric conductivity and viscosity of molten glass of LLW must be obtained to evaluate theapplicability to various type of glass melting furnace after this.
5. Acknowledgements
This study was carried out as a part of the basic research programs of vitrification technology for waste
volume reduction supported by the Ministry of Economy, Trade and Industry, Japan.

46

K03

Study on Vitrification Technology forRadioactive Waste at Fukushima Daiich NPS
Toshiro Oniki1,Yasutomo Tajiri1, Toyonobu Nabemoto1, Toshiki Fukui1,
Seok-Ju Hwang2, Cheon-Woo Kim2

1

IHI Corporation, 2 Central Research Institute, Korea Hydro & Nuclear Power Co.,Ltd

Abstract
The process of various types of radioactive wastes generated in Fukushima Daiichi Nuclear Power Station(NPS) site is one of the issues to be solved. The purpose of this study is to stabilize these wastes by vitrification
technology using CCIM, especially wastes with high level of radioactivity such as secondary wastes generated
from contaminated water treatment systems.
1. Introduction
The process of various types of radioactive wastes generated in Fukushima Daiichi Nuclear Power Station(NPS) site is one of the issues to be solved. IHI started research on the treatment of waste from the Fukushima
Daiichi NPS, and a study of the application of vitrification technology to such waste after the Great East Japan
Earthquake of 2011. The purpose of this study is to stabilize these wastes by vitrification technology, especiallywastes with high level of radioactivity such as secondary wastes generated from contaminated water treatment
systems. Cold Crucible Induction Melter(CCIM) is selected for vitrification process of these wastes by IHI.
CCIM is developed in collaboration with KHNP in this study, zeolite, slurry and sludge are selected as target
wastes.
2. Experiment and Preliminary Conceptual DesignFirstly, laboratory scale tests were performed toconsider appropriate glass compositions for each waste
in order to evaluate the waste loading and durability.

Inside view of CCIM

Physical properties, such as PCT leaching rate,
viscosity and electrical conductivity of candidate glass
compositions were evaluated in order to optimize theglass compositions. Secondly, demonstration tests
using CCIM mock-up facility were performed to

Pouring the glass melt

evaluate the applicability for each waste. The main purpose of

Figure 1. Schematic diagram of CCIM

demonstration test is to confirm the throughput and operation condition during feeding of simulant and glass frit.
The behavior of cesium into the CCIM system was evaluated during feeding period. In addition, homogeneity of
product glass by SEM/EDS analysis and PCT leaching rate were confirmed.
Furthermore, preliminary concept of process flow of CCIM system and arrangement of equipment werestudied for processing the slurry waste.
3. Conclusion
IHI has assessed the applicability of vitrification technology to waste from the Fukushima Daiichi NPS,
and it has selected a CCIM as the glass melter and confirmed that the melter is capable of processing those
wastes such as zeolite, slurry and sludge on mock-up test.

47

K04

Basic research programs of vitrification technology for waste
volume reduction(2) Development of high waste loading glasses

Yoshiyuki Miura1, Tetsuji Yano2, Jun Matsuoka3, Toru Sugawara4 and Atsunobu Masuno5
1

Japan Nuclear Fuel Limited, 2Tokyo Institute of Technology, 3The University of Shiga Prefecture,
4

Akita University, 5Hirosaki University

Abstract
In japan, the advanced vitrification technologies which could more stably immobilize the more low level
wastes and high level liquid wastes (HLLW) for waste volume reduction will be needed in near future. The
technologies for HLLW have been developed from 2014. As a result, we could develop the new glass formulations
having higher waste loading. This paper describes progress of the development.
1. Introduction
In this research, the developmental works such as the high waste loading glasses, the alternate glasses ofcurrent borosilicate glasses including glass-ceramics and the minor actinide adsorbent glasses have been entrustedwith some organizations. JNFL is primarily concerned with the high waste loading glass focused on solubility ofmolybdenum which may cause the formation of undesirable secondary phase (yellow phase) in the case of HLLW
vitrification. JNFL have collaborated with various universities to study the above programs. Furthermore, JNFL
have uniquely investigated the effects of several constituents to keep a good chemical durability with high wasteloading, which are conflicting with each other, of simulated HLLW glasses.
2. Experimental
There are some development approaches for investigation of high waste loading glasses. The one of theapproaches is to modify the composition of the borosilicate glasses which currently used at Rokkasho
Reprocessing Plant without any further additives. In this approach, the effects of concentration of network formers,
intermediates and network modifiers which are fundamental elements of glasses were investigated.
Through this investigation, JNFL had obtained candidate glass composition ranges until FY2016. Currently,
JNFL are undergoing a narrowing the candidate glass composition ranges from the point of view of physicalproperties, chemical durability and melter operation.
And furthermore, the mechanisms of improvement of molybdenum solubility have been investigated fromthe point of view of not only chemical reaction in the cold cap but also glass network structure.
3. Conclusion
In this research, the candidate glass composition ranges for high waste loading were obtained. In the future,
it will be farther optimized by modifying the compositions of formers, intermediates and modifiers from the pointof view of chemical durability, MoO3 solubility and physical properties.
This work was carried out as a part of the basic research programs of vitrification technology for waste
volume reduction supported by the Ministry of Economy, Trade and Industry, Japan.

48

K05

Volume reduction and recycling of contaminated metal by melting
Satoshi Karigome1, Kentaro Shibata1,
Yasuhiro Miyagoshi2 and Norifumi Onogi2

1

The Japan Atomic Power Company (JAPC), 2JFE Engineering Corporation (JFEE)

Abstract
JAPC has an extensive record of nuclear power plant (NPP) decommissioning and assessments forclearance of materials from NPPs. JFEE contributed to disposal of disaster wastes in Fukushima Pref. following
the Great East Japan Earthquake. Utilizing this experience, we will contribute to volume reduction and recyclingof radioactively contaminated metals.
1. Background of JAPC First decommissioning of a commercial

Radiation Shield

Bench

Block

Intrusion

NPP (Tokai NPP) in Japan. (Figure 1)
 Assessment of contaminated wastes fromdecommissioned NPPs for clearance andFigure 1. Dismantling of Fuelprevention block
handling Machine Figure 2. Recycled products

recycling. (Figure 2)
2. Background of JFEE Construction and operation of temporary
incinerators for disaster waste, including
radioactive substances, and design ofvolume reduction facilities for incineration
ash from the temporary incinerators inFukushima Pref. (Figure 3)
 Design and supply of ash melting furnaces
for general waste incineration ash.
3. Melting of radioactive contaminatedmetal Volume

reduction

of

scrap

metal

contaminated with radioactive cesium (Cs)by melting, and separation of radioactive

Figure 3. Incineration plants for radioactive
disaster wastes in Fukushima Pref.
 Assessment of decontaminated metal for clearance and recycling.
Cs from the metal.

Radioactive Cs in metals is distributed to slag and dust
Slag separation using difference in specific gravity
Contaminated scrap metal with radiation level of less thanapprox. 1 mSv/h can be used as clearance metal

Figure 4. Process flow of radioactive Cs separation

49




Enclosed structure to prevent dispersal ofradioactive Cs
Production of ingots for recycling use

Figure 5. Conceptual design ofthe equipment

K06

SIAL® : Geopolymer solidification technology approved by
Slovak / Czech Nuclear Authority
Milena Prazska and Marcela Blazsekova , Wood (Slovakia)
Hisashi Mikami , Nobuyuki Sekine and Mika Mochimaru , 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 practicalon-site treatment of radioactive waste streams at room
temperature, and can incorporate four 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, andFigure 1 SIAL® solidification treatment process

can be applied under water.
®

SIAL solidification technology can realize higher
compressive strength and lower leachability compared to
Cementation, and posing a low fire risk and excellent physicalstability in the presence of frost and water (no distortion or
cracking).
Figure 2 SIAL® solidified samples
3. Performance Record Example
About 650 m3 of radioactive waste streams (resins, sludge and crystalline borates) stored in 14 tanks
situated in auxiliary building of NPP Unit V1 in Jaslovske Bohunice were treated and large capacity tanks were
emptied and cleaned. The purpose of the project was to retrieve the waste streams from operational tanks and its
solidification in 200 liter drums.

50

K07

Contribution to Environmental Remediation in Fukushima (2)Takao Ikeda, Yasuhiro Suzuki, Takahiro Kikuchi
JGC Corporation

Abstract
To reduce the final disposal volume of removed soil and incineration ash from decontamination work, wehave demonstrated the dry pyro-treatment technology for removal of persistently adhering radioactive cesium
from the object materials. Subsequently, we have been developing, with Kyoto University and Taiheiyo Cement,a geopolymer treatment method to safely dispose of the secondary waste generated from pyro-treatment.
1. Dry Pyro-Treatment Technology
The key technology of dry pyro-treatment is tovaporize and remove radioactive cesium in the object
materials using some reaction accelerant in a rotary
furnace at high temperatures of around 1,350 °C.
Through this treatment, the cesium content of the object
materials is reduced to levels that allow unconditional
recycling. In contrast, fly ash with concentrated cesium isgenerated as a by-product. The process flow diagram isshown in Figure 1.

Figure 1 Process flow of dry pyro-treatment process

2. Geopolymer Solidification of Fly Ash
Since fly ash is in the form of chloride, cesium is easily soluble and cannot be expected to remain confinedeven if it is cement solidified. We have been developing a geopolymer solidification process as an optional
technique to solidify fly ash.

1

Leaching ratio of Cs◆Geopolymer▲OPC cement

0.1

0.01

0.01

0.1
1
Leaching ratio of K
Figure 2 Experimental equipment and solidified geopolymer Figure 3 Comparison of leaching ratio between K and Cs
3. Conclusions
The solidified geopolymer of simulated incineration ash was produced, and a compression strength of 20
N/mm2 or more was obtained. It was confirmed that the G value is lower than that of the cement solidified body
because the moisture content is reduced by vacuum heating treatment.
A leaching test in accordance with the IAEA standard leaching test method shows that the leaching ratio ofCs was selectively low as compared with Na and K, and it was observed by XRD and XAFS analysis that
cesium was immobilized in the pollucite structure (CsAlSi2O6).

51

K08

Radiological separation of contaminated materials-experiences with a high performance mobile belt conveyor systemF. Langer1, Dr. M. Sokcic-Kostic1
1

NUKEM Technologies Engineering Services GmbH

Abstract
The existing, containerized, mobile belt conveyor system offers radiological characterization of soil andpossible other materials of approximately similar size with potential contamination. Based on radiologicalcharacterization, the material is segregated and directed to respective containers, thus reducing the volume
for subsequent handling. The system also obtains information and generates documentation to accompany
the characterized material in accordance with regulatory requirements.
Positive experiences with a mobile belt conveyor scanning system (FREMES)
Photos: courtesy of FBFC International

1. Purpose of the belt conveyor scanning systemRadiological characterization of bulk material

Classification of contaminated bulk material

Segregation and filling of the materialVolume reduction for subsequent handlingGeneration of sophisticated documentation per packageComparison with regulatory limits

Figure 1. On-site installation of FREMES

2. Achieved benefitsVolume reduction regarding expansive disposal routes

No secondary waste generated

No need for process media



Detection of Alpha-, Beta- and Gamma- emitters



Evaluation of spatial activity distribution

Proven toughness for construction site applicationFilling in containers according to customer’s requirements Figure 2. Feeding via trucks

3. Technical Data

Throughput up to

100 tons per hour

Sorting size

150 kg Detection limit< 0.005 mSv/h

3x 40’ transport containers

Installation size

Figure 3. Removal of free release material
3. Conclusion
This mobile belt conveyor system can automatically segregate materials into three different material streams(free release, limited release/utilization on site, radioactive waste) and can be optimized tocustomer’s needs.
References:

www.nukemtechnologies.com; jochen.petermann@nukemtechnologies-si.de

52

K09

Radioactive Waste Management- treatment, storage and disposal of LILW and HLW K. Büttner1, R. Slametschka1
1

NUKEM Technologies Engineering Services GmbH

Abstract
Appropriate conditioning of all kinds of waste reduces waste volumes thereby saving storage space andhelping to limit the costs of interim and long-term storage, making waste less hazardous and ensuring a safe
final disposal. NUKEM offers customized solutions for all waste problems which rank amongst the world’s
leading technologies for the treatment of radioactive waste, e.g.:- Waste treatment concepts and systems, as well as turnkey waste treatment centers;- Interim storage and final disposal facilities.

1. Waste treatment technologies and systems


Evaporation

Concentration



Microwave dryingNuclide separation

Ultrafiltration / Reverse osmosisBiological water treatment

Cementation

Figure 1: Waste Treatment Center



Vitrification

Reference Project:High-force compaction

Industrial Complex for Radwaste Management at Chernobyl NPP,

Incineration

Ukraine (throughput over 3,500 m3 waste per year).



Pyrolysis / Pyrohydrolysis

Sorting / Segregation

2. Waste storage and disposal facilities

Temporary storage for Lowand Intermediate Level Waste

Near-surface disposal facility

3. Conclusion
Depending on customers requirements and the types
and quantities of waste NUKEM Technologies offers
already proved waste treatment concepts up to
turn-key waste treatment centers and storage andfinal disposal facilities.
Schematic View of the long-term cover proposed for the Kozloduy
repository

Figure 2: Waste Storage and Disposal Facility (schematic)
References:

www.nukemtechnologies.com ; jochen.petermann@nukemtechnologies-si.de53

Tailor-made solutions with proven Technology
to Waste Management

K10

Motoki Nishibori
KOBELCO STUDSVIK Co., Ltd.
KOBELCO STUDSVIK Co., Ltd. was formed on July, 2016, jointly by Kobe Steel Ltd. andStudsvik AB, focusing on delivering design, engineering to provide innovative radioactive waste
management solutions to the Japanese nuclear industry. KOBELCO STUDSVIK will provide
technologies and processes to reduce radioactive waste safely and recycle metals generated fromthe decommissioning effectively for the Fukushima Daiichi Nuclear Power Station.

1. Metal Recycling

In EU, where the effective reuse of resources has socially taken root, metal materials of nuclear
power equipment are safely recycled. Studsvik is the organization who developed the metal
recycle model in 1980’s. The technology can be used for the increasing demand for metal recyclingin Japan. In Fig. 1, metal recycle system flow and, in Fig. 2, photos of track record in Berkeley are
shown respectively.

Fig. 2 Photo of 1,550t Boilers in Berkeley (UK)

Fig. 1 Metal Recycle system flow

2. THOR technology

THOR (THermal Organic Reduction) is a proventhermal treatment technology for LLW and ILW,
with pyrolysis and steam reforming mechanism.
Fig. 3 shows THOR process flow.
(1) Wastes
Ion exchange resins, rubbers, activated carbon,
plastics etc. are treatable wastes.
Fig. 3 THOR process(2) Reformed Residue
Radioactive wastes are converted into dry and stable
solid residue product (Reformed residue).
The reformed residue in which radioactive
nuclides are captured is insoluble and leach-resistant.
(3) Volume reduction
When spent ion exchange resin is treated,
Before(resin)
After(residue)
its volume is reduced up to 1/15, depending on
waste type and final waste form.
Fig. 4 Reducing up to 1/15
Fig.4 shows the Sample resins before-and-after treatment.

54

J01

Current Status of the JAEA’s Fukushima R&D Bases for theDecommissioning of the Fukushima Daiichi Nuclear Power Station
Nobuhiro Shobu, Makoto Ooka, Toshiro Nakai and Shinichi Nakayama
Japan Atomic Energy Agency

Abstract
The Japan Atomic Energy Agency (JAEA) is operating and constructing R&D bases at the coastal area of
the Fukushima Prefecture for the decommissioning of the Fukushima Daiichi Nuclear Power Station (NPS) of
the Tokyo Electric Power Company Holdings, Inc. (TEPCO).

1. Introduction
R&D for the environmental restoration of Fukushima and the decommissioning of the Fukushima Daiichi
NPS is a prioritized mission of JAEA. JAEA is carrying out the R&D activities at the Collaborative Laboratories
for Advanced Decommissioning Science (CLADS) in Tomioka Town and the Naraha Center for Remote Control
Technology Development in Naraha Town (Naraha Center), and constructing the Radioactive Material Analysis
and Research Facility in Okuma Town (Okuma Analysis and Research Center: Okuma Center).

2. JAEA’s R&D bases for the decommissioning of the Fukushima Daiichi NPS
2-1. Collaborative Laboratories for Advanced Decommissioning Science
CLADS is the core of JAEA’s R&D for the decommissioning of the Fukushima Daiichi NPS, strongly
linked with each infrastructural base of Naraha Center and Okuma Center. CLADS began the operation in April
2017 of the Main Building in Tomioka Town, where the evacuation order was lifted in April 2017. CLADS is
expected to act as both an international research hub and a symbol of the shared international interest in
decommissioning the Fukushima Daiichi NPS.
2-2. Naraha Center for Remote Control Technology Development
JAEA is working on remote control technology development. The Naraha Center for Remote Control
Technology Development, which began operation in April 2016, are available for external users who areinterested in R&D on remote-control devices (e.g., robots) necessary for decommissioning work. The external
users can conduct demonstration/element-test and operation training here. The center is not only fordecommissioning related work, but also for developing robots and their demonstration tests for disaster
responses.
2-3. Okuma Analysis and Research Center
The Okuma Analysis and Research Center is supposed to analyze and characterize the radioactive wastes
and nuclear fuel debris for development of long-term waste management. The center consists of three buildings:
an Administration building, Laboratory-1 and Laboratory-2. Administration building was opened in March 2018.
Laboratory-1 is under construction, and Laboratory-2 is at the designing phase.

55

J02

Operation Training System for Decommissioning Work Using
Immersive Virtual Reality Technology
Yoshihiro Tsuchida1, Takahiro Ohno1, Nobuhisa Nosaki1 and Mitsugu Kato

1

Naraha Center for Remote Control Technology Development, Japan Atomic Energy Agency

Abstract
The Naraha Center for Remote Control Technology Development is a shared facility to conduct
development/demonstration tests of remote controlled devices such as robots used for the decommissioning ofFukushima Daiichi Nuclear Power Station. In addition to the test equipment for such devices, the center provides
operation training system for decommissioning work using immersive virtual reality technology.

1. Introduction
In order to carry out safely and reliably the decommissioning work, it is important to prepare a work
procedure after sufficiently understanding of the site condition in advance and pre-training. However, it is difficult
to do that on site due to high radiation dose rate. Therefore, we developed and installed the operation training
system using immersive virtual reality technology in the center.

2. Specifications and functions of the system
The system is composed of a computer cluster system to visualize
the 3D model data, device for measuring trainee’s eyes, projectors,
screens and a console for a supervisor. The immersive area configured
with CAVE style screens has a height and a width of 3.6 meters and a
depth of 2.25 meters, and makes it possible to display structures in a fullsize. Consequently, trainee can look up at the ceiling direction and obtain

Traineea high immersive feeling.

Fig. 1 View of the system

Supervisor

Data on the first and the basement floors of the reactor buildings
for unit 1 - 3 have been equipped in the system thus far. The trainee canfreely move within the virtual space and train decommissioning work by
combining with functions such as indication and calculation of airradiation dose rate, lighting simulation, distance measurement ofarbitrary two points, import of 3D CAD data of equipment.
Fig. 2 Training in the system
3. Summary
Utilizing this system enables trainee to experience the realistic work environment repeatedly, consider andverify work plans such as a traffic line of a human, a loading/unloading route and an installation site of equipment,
and so on. As a result, we believe that this system contributes toward carrying out safely and reliably thedecommissioning work, and improving the decommissioning working efficiency. Our future works are to expand
the data based on needs of candidate users, and to develop a mechanism for applying ever-changing on-sitesituation to the system.

56

J03

Outline of JAEA Okuma Analysis and Research Center
Akihiro Ishimi, Toshihiko Inoue, Ichiro Kosaka, Shinichi Koyama, Tomozo
KoyamaJapan Atomic Energy Agency

Abstract
Japan Atomic Energy Agency (JAEA) is constructing Okuma Analysis and Research Center near Tokyo
Electric Power Company (TEPCO), Fukushima Daiichi Nuclear Power Station (1F) site. Outlines of the Okuma
Analysis and Research Center are introduced at the poster.
1. Outline of the Okuma Analysis and Research Center
JAEA works on various missions for decommissioning of the 1F, and the research and development facility“Okuma Analysis and Research Center” is designed and has been constructed near 1F. As shown in Figure-1, theOkuma Analysis and Research Center consists of three buildings; an administrative building, Laboratory-1 andLaboratory-2. The Okuma Analysis and Research Center will be used for research and development to ascertain
characteristics of radioactive wastes, and technological development to treat and dispose fuel debris.
2. Administrative building
The Administrative building provides office space and meeting rooms for the researchers, and an apparatus
mock-up space “Workshop”. The simulated steel hot cell, Glove Box, hoods and some apparatus such as
master-slave manipulators were installed at the workshop, and the mock-up test and training of the analysis will
be performed with these apparatus. The construction of this building began in October 2016, and an operation
started in March of 2018.
3. Laboratory-1
Laboratory-1 will perform radioactive analysis, chemical and mechanical characterization of low andmedium level of radioactive wastes such as radioactive rubble and secondary wastes to establish strategy andmethodology for treatment and disposal of them. The schematic layout is shown in Figure-2. Laboratory-1 will
be equipped with 10 glove boxes, and 56 hoods for low level radiochemical analysis. For the handling of
medium level samples, 4 steel cells will be installed. The construction of Laboratory-1 began in April 2017 andplans an operational start in fiscal year 2020.
4. Conclusion
JAEA started an operation of Administrative building, and is moving forward with the construction ofLaboratory-1 and the design of Laboratory-2 for the analysis of the fuel debris.

Steel cell
Hood

Glove box
Figure-1 Okuma Analysis and Research Center

Figure-2 Laboratory-1

57

J04

Plasma heating test on fuel assembly degradation
Takuya Yamashita1,2, Ikken Sato1,2
1

JAEA, 2IRID

Abstract

A large uncertainty is present in understanding of BWR accident progression behavior. A series of tests
were conducted to provide experimental data that address uncertainties. As a results, useful information on corestate just before slumping was obtained. This information will help us to comprehend core degradation in BWRs
like those in the Fukushima Dai-ichi nuclear power plant (1F).
1. Introduction
Core degradation in a BWR like the 1F has not been comprehensively understood. This uncertainty, which
was clarified by the MAAP-MELCOR Crosswalk [1], cannot be resolved with existing experimental data andknowledge. In this study, a core-material melting and relocation (CMMR) experiments were conducted toprovide information to answer the uncertainties related to BWR-design conditions so that understanding on
accident progression behavior in 1F can be improved.
2. CMMR experiment design
The CMMR test equipment mainly consists of the plasma heating system, and thesimulated fuel assembly. The plasma heating system used for the tests were a Phoenix
Solutions Company commercial-grade non-transferred torch (model PT200). The simulated
fuel assembly was composed of simulated a fuel rod bundle within a channel-box of BWR.
In this test, ZrO2 pellets were installed instead of UO2 pellets.
3. Test conditions
The test conditions were selected with intention to simulate the state of 1F Unit 2
just before slumping. Oxygen concentration kept less than 0.1%. Maximum temperature
was close to the melting point ZrO2 pellets.
3. Conclusion(a) Macroscopic gas permeability of the core approaching ceramic-fuel melting was
confirmed under the condition simulating the 1F Unit2 condition (See Fig.1).
(b) The hot core fuel remained as columns in these tests suggesting possible actual BWR
scenario where weight of core upper part would result in collapse of fuel columns
basically keeping axial profile of fuel temperature (though shortened) rather thanpellet-wise relocation from the hottest part (See Fig.1).
(c) Fuel debris without melting would keep its permeability even after collapse andrelocation

to

lower

plenumallowingenhancedcooling

capability

whenheat-transferring fluid and heat sink is maintained.
Acknowledgements

The content of this material includes output of the METI subsidy for Project ofDecommissioning and contaminated Water Management (Upgrading for Identifying
Comprehensive Conditions inside the Reactor)

Fig.1 Post-test
appearance of
the CMMR-4

Reference

test piece[1] EPRI,”Modular Accident Analysis Program (MAAP) – MELCOR Crosswalk, Phase 1 Study”, Nov.2014

58

J05

Characterization of the Large-scale MCCI Test Products for Fuel
Debris Removal from the Fukushima Daiichi Nuclear Power Plant
Hirotomo IKEUCHI1, Hiroyasu HIRANO1, Tadahiro WASHIYA1,
Pascal PILUSO2, Laurent BRISSONNEAU2
1

JAEA/IRID, 2CEA

Abstract
A large-scale test of the molten core-concrete interaction (MCCI) was performed to investigate heterogeneity ofcharacteristics of the MCCI products. The results of the post-test analysis confirmed importance of appropriate
selection of defueling tools according to the location and the characteristics of MCCI products.
1. Introduction
In the Unit 1 of the Fukushima Daiichi Nuclear Power Plant (1F), the molten core materials are expected to haveflowed into the bottom of the primary containment vessel, causing the molten core-concrete interaction (MCCI)at the pedestal floor. Clarifying the characteristics of the MCCI products is important for the removal of thosematerials from the damaged floor. Especially, characteristics such as solid phases and mechanical properties
could be locally changed due to spatial variation of the material composition and temperature history, making
appropriate selection of defueling tools much more complicated. To achieve insights on this “heterogeneity” of
the MCCI products, a large-scale MCCI test was performed in collaboration with CEA, France.
2. Method
Conditions for the large-scale test were constructed specifically taking the average composition of core materials
and heating conditions for the 1F Unit 1. Approximately 50 kg of the simulant core materials (mixture of UO 2,
Zr, ZrO2, and stainless steel) were melted in the crucible made of
the basaltic concrete. Post-test analyses were performed on thesolidified samples taken from various representative zones, e.g.,bulk region, interface between molten cores and concrete, etc, to“Hard” porous oxide(Si-rich oxide)

characterize the solid phases and micro-hardness.
3. Conclusions
The post-test analysis of the samples revealed stratified structureof the solidified melt which mainly consisted of hard porousoxide (Si-rich), hard dense oxide (Si-rich), and soft metallic
block (Fe-rich), as shown in Figure 1. The difference incharacteristics will necessitate the appropriate selection ofdefueling tools according to the location and the characteristics“Soft” metallic block(Fe-based alloy)“Hard” dense oxide(Si-rich oxide)

Figure 1. Typical appearances of thesolidified melt after the large-scale
MCCI testof MCCI products.
Acknowledgment

This paper includes the results obtained under the research program entrusted to the International ResearchInstitute for Nuclear Decommissioning (IRID) by the Agency for Natural Resources and Energy, Ministry of
Economy, Trade and Industry (METI) of Japan.

59

J06

Advanced study on remote and in-situ elemental analysis for molten
fuel debris in damaged core by innovative optical spectroscopy
Ikuo Wakaida1, Hironori Ohba1,2, Yuji Ikeda3, Tetsuo Sakka4, Takunori Taira51Japan Atomic Energy Agency/CLADS, 2QST/QuBS, 3Imagineeing Inc., 4Kyoto Univ., 5NINS/IMS

Abstract
Radiation resistant optical fiber based laser induced breakdown spectroscopy (Fiber LIBS) is developed forin-core and in-situ elemental analysis of debris in damaged core, and its activity is examined under highradiation field of about 10kGy/h and under water. For higher sensitivity, long pulse laser, microwave assisted
LIBS are performed, and additionally, Microchip laser for longer light delivery and flexible use is introduced inhigh radiation condition.
1. Introduction
For the decommissioning of “Fukushima Daiich Nuclear Power Station(F1NPS)” which contained
damaged or melt downed core, development of rapid, easy, onsite and especially in-situ remoteanalysis/surveillance techniques under the severe environments such as extremely high radioactive condition
will be strongly required. In order to accomplish these requirements, the concept of “Probing by light andAnalyzing by light” with radiation resistant optical fiber will be one of the simple, powerful and applicablechoices as the innovative technical development based on LIBS technology.
2. Current activities of LIBS
2-1. Conventional fiber LIBS
As for R&D of onsite, in-situ
monitoring under the severe
environments, we have beendeveloping the Optical Fiber LIBS
Probe using with radiation resistant
optical fiber for laser and plasma
Fig.1 Schematics of Fiber LIBS system and its application for F1NPS
emission light delivery. The concept
ure 1. Cross section of BWR
of Fiber LIBS and its application for in-situ monitoring are shown in Fig.1. The transmission of the optical fiberin near infrared range will be confirmed to have no damage by radiation rate of 10kGy/h and total dose of morethan 2MGy by the 60Co irradiation test. Characteristics of plasma generation, emission under strong radiation
have been checked and no influence in specific emission line under strong radiation will be confirmed.
2-1. Microchip laser LIBS
As for the requirement of longer
application in remote analysis, we are now
developing “Microchip laser LIBS system” inwhich the laser source will be installed at the
top of the LIBS Probe Head as shown in Fig.2.
Fig.2 Basic concept of Microchip Laser Fiber LIBS system
Unlike the Fiber LIBS system in which
breakdown laser delivered though optical fiber, laser power and focusing characteristics will be improved, and it
may have better performance such as longer focusing length and possibility of laser processing to the surface ofdebris. A good performance in laser emission under strong radiation of 10kGy/h will be just now demonstrated.
Present study includes the results of “Advanced study on remote and in-situ elemental analysis of molten fuel debris indamaged core by innovative optical spectroscopy” entrusted to Japan Atomic Energy Agency by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

60

J07

Application of ICP-MS to analysis of samples from 1F siteat Radioactive Material Analysis and Research Facility
Van-Khoai Do1,2, Takuma Horita1,2, Katsue Miura1,2, Maho Iwasaki1,2
1Japan Atomic Energy Agency (JAEA)
2International Research Institute of Nuclear Decommissioning (IRID)*

Radioactive Material Analysis and Research Facility has been recently established to support the analysis
of samples including fuel debris and radioactive wastes from the decommissioning of Fukushima Daiichi Nuclear
Power Station (1F). Conventional radiometry, which is laborious and time consuming, ICP-MS is considered as apotential alternative for long-lived nuclides in fuel debris because of its superior sensitivity. Moreover, the use of
ICP-MS is aimed to increase the analysis capacity, which make possible to repeat measurements in order to obtain
precise results and to reduce exposed dose of analysts because ICP-MS is much more rapid than radiometricmethods. However, isobaric interference from the same mass nuclides is likely to limit the application of ICP-MS
in our field. Employing latest advanced technologies, a triple quadrupole ICP-MS, Agilent 8900 (ICP-QQQ-MS),enables to suppress the interference. In this presentation, we are reporting some demonstrative separations ofdifficult-to-measure nuclides using the ICP-QQQ-MS system.
Figure 1 illustrates the schematic structure of the ICP-QQQ-MS.The ICP-QQQ-MS is equipped with two
quadrupole mass filters (Q1 and Q2) that are tandemly arranged, and an octopole collision/reaction system (ORS)
is set between the filters [1]. Mechanism ofReaction Gas
interference suppression is described asion
EMfollows: samples are first ionized in the Ar Ar+ ion current
Detector
plasma. Then, the resulting ions are led to the
Q1
ORS
Q2
first quadrupole filter (Q1), which allows to
MS/MS (QQQ)
pass through only ions with m/z of interest
Fig. 1: Illustration of ICP-QQQ-MS
including the same mass interferents. Thenthese selected ions are led to react with reaction gases to form different complex ions with different mass-shifts.
The second quadrupole (Q2) functions as the second mass filter that allows the ions of interest to be detected byan electron multiplier detector (EM detector).
For demonstration of selective measurement,interfering ion pairs were selected aiming at future applications,
including (Ni-59, Co-59), (Ni-63, Cu-63), (Se-79, Br-79), (Sr90, Zr-90, Y-90), (Zr-93, Nb-93, Mo-93), (Pd-107, Ag-107), (I129, Xe-129), (Cs-135, Ba-135), and (Sm-151, Eu-151).
Presently, non-radioactive solution standards are using toevaluate the separation. Four gases including He, H2, NH3, andO2 were tested as collision and reaction gases. Table 1
summarizes the selected reaction gases that are possible toseparate nuclides of interest. LOD (part per trillion-ppt or ng/L)
is limit of detection of the most abundant natural isotopes that
Table 1: Result of reaction gas selection
are used as tracers. The results indicated that it is possible toeliminate the interference by using a suitable reaction gas. ICP-QQQ-MS promises an alternative method for rapid
analysis of fuel debris as well as radioactive samples from 1F.

Reference[1] ICP-QQQ-MS Agilent 8900 brochure, available online at https://www.agilent.com/en/products/icp-ms/icp-mssystems/8900-triple-quadrupole-icp-ms* This presentation includes results obtained under the research program entrusted to IRID by METI.

61

J08

Estimation of radioactive nuclide distribution at NPS site
Developing model based on analytical data of on- and off-site

Kazuki Iijima1, Yoshikazu Koma1, Tadafumi Niizato1, Terumi Dohi1, Yoshihito Ohmura2,
Takayuki Sasaki3 and Kenji Nanba4
1

Japan Atomic Energy Agency, 2National Museum of Nature and Science, 3Kyoto University,
4
Fukushima University

Abstract
Distribution of radionuclides around the site was investigated, and the model of radioactive contamination
was developed to describe inside and outside of the Fukushima Daiichi Nuclear Power Station (NPS).
1. Introduction
To establish method of radioactive waste management, inventory of radioactivity contained in waste is
essential information. Distribution of contamination is also important to estimate inventory of the wastes including
those will be generated in the future decommissioning. Environmental contamination around the site is useful toobtain information to make a mathematical model of radioactivity distribution for the on-site. Contamination of
topsoil, lichen and tree was investigated, and a model was made by considering the data as well as reported onsite data.
2. Analysis of environmental samples and investigation on contamination model
Environmental samples were collected with considering on the traces of radioactive plumes and analyzed.
With utilizing the data a contamination model was proposed.
2-1. Topsoil
According to concentration depth profile, Cs was retained at thesurface, while Sr was slowly transported, suggesting correlation with
distribution coefficient. Model constituted from three processes well
reproduced the profiles.
2-2. Lichen
Lichens can accumulate radionuclides for a long period. Captured
small particles on the thalli were found to be different in chemicalcomposition. The ratio of 134Cs/137Cs in lichens and their particles werealso found to be different. These findings suggest multiple sources fromreactors and variable states in transport.
2-3. Tree137
Cs
For modeling, it was assumed that cedar retains initial
contamination on its bark, while Cs was also uniformly distributed
inside due to nutrient translocation. Any α emitters were not detected.
2-4. Modeling on-site contamination
By expressing a plume as Gaussian dispersion distribution andnuclide concentrations as relative ratio of transport, on-site distributionof radioactive was estimated as shown in Figure.
3. Conclusion
The derived radioactivity distribution will help analyticalinvestigation including planning. And knowledge of nuclide
composition of plumes and each reactor contamination will bring better
estimation of contamination inside the buildings.
90
238
Sr
Pu
Note: This work is the result from the work entrusted to the parties of the
Figure. Estimated radionuclide onauthors from MEXT.
site distribution. Unit in MBq/m2.

62

J09

Development of Corrosion Database under Radiation Environment
Tomonori Sato1, Kuniki Hata1, Yoshiyuki Kaji1, Hiroyuki Inoue2,
Kousei Taguchi3, Hajime Seito3, Eiji Tada4, Hiroshi Abe5, Eiji Akiyama5 andShunichi Suzuki6

1

Japan Atomic Energy Agency, 2Osaka Prefecture Univ., 3National Institutes for Quantum andRadiological Science and Technology, 4Tokyo Institute of Technology, 5Tohoku Univ., 6Univ. of Tokyo
Abstract
Since risk of corrosion degradation for plant materials in Fukushima-Daiichi Power Plant (1F) site havebeen increasing with time duration and/or environmental changes by decommissioning procedure, we will build
a corrosion and radiolysis database in irradiated condition including estimating data area for 1F corrosion.

1. Introduction
Risk of corrosion degradation for plant materials in 1F site have been increasing with time duration and/or
environmental changes by decommissioning procedure. Preventing methods for these corrosion risk aredeveloped based on corrosion mechanism. Preventing methods for these corrosion risks are developed based on
corrosion mechanism.

2. Research plan and results
2-1. Mechanism of corrosion in irradiated condition
To build a database for estimating the degradation of corrosion in irradiated condition, corrosion andradiolysis data will be accumulated as shown in Fig. 1. Concerning development of data set and database forradiolysis, arrangement/estimation of radiolysis existed data and preliminary radiolysis analysis were carried out.
Corrosion test environment in gamma-ray irradiation was constructed in Takasaki Advanced Radiation ResearchInstitute of National Institute for Quantum and Radiological Science and Technology (QST). In this facility,
gamma-ray irradiation can be carried out in the range from ten to thousands of dose rate. Regarding mechanismof corrosion in irradiated condition, preliminary corrosion tests for influence of gamma irradiation on bothde-passivation of carbon steel and corrosion in wet/dry
environment have been started.
2-2. Evaluation of corrosion risk in each equipment
Corrosion risk for individual equipment duringdecommissioning process is analyzed by using
investigation sheet. Concerning estimation method forpotential corrosion risk, basic concept was constructedwith overlooking overall decommissioning process of
1F by existed knowledge and literatures.

This project is conducted as a part of “Analysis ofCorrosion Mechanism in Specific Environment”
supported by the Ministry of Education, Culture,
Sports, Science and Technology (MEXT) of Japan.

Fig. 1 Data area for evaluation of 1F corrosion

63

J10

A Prediction of the Dose Rate Distribution in the PrimaryContainment Vessel of the Fukushima Daiichi Nuclear Power Station
Keisuke Okumura1 and Eka Sapta Riyana1
1Collaborative Laboratories for Advanced Decommissioning Science (CLADS),
Japan Atomic Energy Agency (JAEA)

Abstract
We have predicted the dose rate distribution inside PCV of Unit 1 and Unit 2 in 2021 by combination ofMonte Carlo calculation with local dose rates measurements obtained so far by PCV internal investigations.
1. Method
The dose rate distribution in PCV 𝐷𝐷(r⃗) is expressed as follows:𝐷𝐷(r⃗) = � 𝑆𝑆𝑖𝑖 𝑑𝑑𝑖𝑖 (𝑟𝑟⃗)𝑖𝑖where, i indicates the radiation source such as Cs contamination, fuel debris and activation materials, which are
estimated based on the accident progression analysis, burnup calculation, and activation calculation, respectively,
𝑑𝑑𝑖𝑖 (𝑟𝑟⃗) is a Green function obtained by the particle transport Monte Carlo calculation code PHITS [1] for the i-th

unit radiation source in PCV, and 𝑆𝑆𝑖𝑖 is source intensity of the i-th radiation source. Although the uncertainty of𝑆𝑆𝑖𝑖 is large, if local dose rate measurements were obtained by internal investigation, we can modify the source
intensities to fit the measured dose rate.

2. Predicted result
The above method was applied to the Unit 1 and Unit 2 of the Fukushima Daiich NPS. Table 1 shows theinternal investigations in which dose rates have been measured. The measured dose rates were considered tomodify the Cs contamination sources sensitive to measured dose rate. The results of modified dose rate
distributions in 2021 are shown in Fig.1.
Table 1 Internal investigations in Unit1 and Unit 2
Unit 1Name
1st entry

Unit 2
Date

Data

Oct.2012

8

Name

Date

Data

1st entry

Mar. 20128

B1 investigation

Apr. 201516 A2 investigation

Jan.-Feb. 20173

B2 investigation

Mar. 201720 A2' investigation

Jan. 20188

3. Conclusion
We have successfully predicted thedose rate distribution consistent withdose rate measurements so far. This
result is valuable to develop detection
technologies of fuel debris in PCV andto optimize fuel debris retrieval method.

(a) Unit 1(b) Unit 2

Figure 1. Model and predicted dose rate distribution in PCV

References[1] T. Sato, Y. Iwamoto, S. Hashimoto, et al., “Features of Particle and Heavy Ion Transport code System (PHITS) version
3.02, J. Nucl. Sci. Technol. 55, 684-690 (2018).

64

J11

Study on Wet Waste Storage - Irradiation Behavior of Carbonate
Slurry and Wet Zeolite Isao Yamagishi1 and Takafumi Motooka1
1Japan Atomic Energy Agency

Abstract
Carbonate slurry or wet zeolite as a simulated waste was irradiated by gamma-ray. Radiolytic gas
generation resulted in the slurry expansion and supernatant separation from the slurry.. The supernatant wasobserved during irradiation of the zeolite immersed fully in seawater but not in case of the partial immersion.
1. Introduction
Overflow of supernatant from storage containers of radioactive carbonate slurry generated from
ALPS(MRRSTM) was discovered at Fukushima Daiichi Nuclear Power Station in 2015. Hydrogen gas was also
detected in the containers. We have conducted irradiation experiments of the simulated slurry by Co-60
gannma-rays to elucidate the mechanism originated from radiolytic gas generation. Irradiation behavior of thezeolite, which is supplied for Cs adsorption vessel of SARRYTM was investigated using its mixture of seawater.
2. Irradiation experiments and results
The simulated carbonate slurry provided by Kurita Water Industries Ltd was poured in a quartz tube (ID 20mm) to 100 mm height, and irradiated with PMMA dosimeter at a dose rate of 8.5 kGy/h as shown in Figure 1[1]. Radiolytic gas generation and its bubble formation resulted in the slurry expansion and supernatantseparation from the slurry. The increase of dose led to the decrease of slurry expansion due to the partial bubble
release and the increase of supernatant volume.
The mixture of the zeolite (UOP IE96) and artificial seawater (ASW) was poured into a quartz tube (ID 40mm), and the ASW volume was adjusted to two levels; (1) top of zeolite bed (full immersion) and (2) a quarterbed (1/4 immersion). Two tubes were irradiated by Co-60 gamma-ray at 7.2 kGy/h for 27 h as shown in Figure 2[2]. In full case, the ASW level rose from 402 mm to 438 mm with the supernatant formation. However, thesupernatant was not formed in 1/4 case, which simulates a drained Cs adsorption vessel.
3. Conclusion
Radiolytic gas generation is a key issue for the safe storage of highly radioactive and wet wastes. We
observed its influence during experiments but need further investigation to elucidate the mechanism.
Full immersion 1/4 immersion

Before 8.5kGy/hIrradiation

after 6h

22h

44h

supernatantsupernatantbubbleafter 27h irradiation

Figure 1. Irradiation of carbonate slurry (SS: 95 g/L)

Figure 2. Irradiation of mixturesof zeolite and seawater

References[1] T.Motooka, R.Nagaishi and I.Yamagishi, Gas retention behavior of carbonate slurry under γ-ray irradiation”, QST
Takasaki Annual Report 2015 (2017).
[2] T.Motooka and I,Yamagishi, “Behavior of Radiolytic Gas in Zeolite Bed under Gamma-ray Irradiation”, Proc. of Fall
Meetings of Atomic Energy Society of Japan, 2A03 (2017).

65

J12

Visualization of radioactive substances by integrating radiation
measurement, remote technology, and 3-D optical measurement
Yuki Sato1, Yuta Terasaka1, Masaaki Kaburagi1, Yuta Tanifuji1 and Tatuso Torii1
1

Japan Atomic Energy Agency

Abstract
We visualized a hotspot from the sky over the waste storage space inside the Fukushima Daiichi Nuclear
Power Station (FDNPS) using a remote radiation imaging system consisting of a drone equipped with aCompton camera. By superimposing the images of the hotspots on the 3D building model of the measurementarea, we succeeded in drawing the map that can easily recognize the hotspots in the actual working environment.
1. Introduction
The measurement of a distribution of radioactive substances in the FDNPS is essential in order to executethe appropriate decommissioning tasks. Last year, we drew a 3D radiation distribution map by integrating theradiation image measured with the Compton camera into the point cloud data of inside the turbine building ofunit 3 acquired using a scanning laser range finder [1]; we fabricated a compact Compton camera based on thetechnology developed by the joint team from Waseda University and Hamamatsu Photonics K.K. [2]. In
addition, we are also developing the remote measurement technology of radioactive substances using a drone
equipped with the Compton camera. In this work, the measurement result using these devices are introduced.
2. Radiation imaging experiment inside the FDNPS
We visualized a hotspot from the sky over the waste
storage space in the FDNPS using a remote radiation
imaging system consisting of a drone equipped with aCompton camera [3]. Figure 1 is a 3D model of the waste
storage space created using a photogrammetry technic.
The hotspots are displayed in red on the 3D building
model. The areas where the survey meter showed ahigher dose rate than the surroundings roughly agree withthe red areas detected by the Compton camera. We
succeeded in drawing the map that indicates the positionsof the hotspots in the working environment.

Fig. 1. Radiation images are drawn on the 3D
building model of the waste storage space in the
FDNPS. The areas shown in red are the hotspots
detected by the Compton camera. Values are
ground survey results at 1 m height (mSv/h).

3. Conclusion
The images of the hotspots obtained by the Compton camera mounted on the drone are drawn on the 3D
building model of the working environment created using the photogrammetry. It is a helpful method forworkers to easily recognize the hotspot at the work environment inside the FDNPS.
References[1] Y. Sato et al., Radiation imaging using a compact Compton camera inside the Fukushima Daiichi Nuclear Power Stationbuilding, J. Nucl. Sci. Technol., (in press), doi.org/10.1080/00223131.2018.1473171.
[2] J. Kataoka et al., Handy Compton camera using 3D position-sensitive scintillators coupled with large-area monolithic
MPPC arrays, Nucl. Instr. and Meth. A, 732 (2013) 403-407.
[3] Y. Sato et al., Remote radiation imaging system using a compact gamma-ray imager mounted on a multicopter drone, J.
Nucl. Sci. Technol., 55 (2018) 90-96.

66

J13

Development of simulation function of remotely operated vehicle
for underwater reconnaissance tasks
Kenta Suzuki, Shunji Mukaida and Kuniaki Kawabata
Japan Atomic Energy Agency

Abstract
This paper describes the development of the ROV (Remotely Operated Vehicle) simulation function
implemented on our robot simulator system. The function provides virtual thruster device to generate thrusting
forces and torques and also, it calculates the effect of fluid dynamics, buoyancy, lifting force and so on. In thispaper, we introduce developed function and test result with ROV model utilized for underwater reconnaissancetasks.
1.

Introduction
Currently, we develop a robot simulator system for supporting effective decommissioning tasks execution

by utilizeing remotely operated robots [1]. We also continue to extend its functions by plug-in software. In someof survey tasks at Fukushima Daiichi Nuclear Power Station (FDNPS), ROVs were deployed and a ROV was
utilized for the task to investigate the situation inside PCV. Therefore, we decided to implement a new function
to simulate underwater behavior of robot to our system.
2. ROV simulation function implemented on robot simulator system
In order to simulate behaviors of ROV, the calculations for
applying the effect fluid dynamics and some visual properties must be
done and simulated device generates thrusting forces and torques is also
required. Therefore, we developed a plug-in software and virtual device
to implement such functions to the robot simulator.
For adjusting the effect of fluid dynamics, the buoyancy, lifting
force, fluid resistance force, and giving the effect of additional mass on
the ROV’s body can be set on developed plug-in by the user. Simulated a
Figure 1. Example view of ROV simulation
thruster device can be added on the body part where the user wishes
at. By these functions, the behavior of ROV in underwater environment are realized. Fig.1 shows an example
view of ROV simulation. In this test run, we use a simple plant-like model that referring to TOSHIBA’s ROV
that was deployed into the PCV of FDNPS [2].
3. Conclusion and future works
Developed ROV simulation function provides thruster device to add thrusting forces and torques and some
fluid dynamics effects to ROV’s behavior in underwater environment. By test run, we could confirm possible tosimulate the behavior of remotely operated robot for underwater survey tasks.
In our future work, we would to develop the other plug-in functions to extend our robot simulator systemfor comitting robot development and the operator proficiency training.
References[1] Kenta Suzuki, Kuniaki Kawabata, Mitsuru Isowa, Tatsuo Torii, “Development of a Robot Simulator for supporting todevelop Nuclear Emergency Response Robots”, The Robotics and Mechatronics Conference 2016, 2016[2] Handouts at press conference 2017 | Archives “Progress of Unit 3 PCV internal investigation (Preliminary report of July
19 investigation)”, Tokyo Electric Power Company, http://www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/index-e.html,(June, 2018)

67

J14

Information Gathering of Working Environment Based on
3D Reconstruction from Plural Images for Decommissioning
Toshihide Hanari1 and Kuniaki Kawabata1
1Japan Atomic Energy Agency

Abstract
This paper describes our research on an environmental information gathering method of working
environment based on 3D reconstruction from plural images for decommissioning work. In this paper, we report
to apply 3D reconstruction method and post-processing for improving visibility of the environment model to theimages from an underwater investigation task inside Primary Containment Vessel (PCV).
1. Introduction
For decommissioning work of Fukushima Daiichi Nuclear Power Station (FDNPS) of Tokyo ElectricPower Company Holdings, Inc. (TEPCO), task execution by remotely operated machine is one of the key issues.
For such remote operation, structural data of the working environment is useful for sharing information andplanning next or future approach. Currently, we consider to utilize a 3D modeling method from images thatcollected by past operation. Structure from Motion (SfM) is a method for estimating 3D structure of objects andcamera’s trajectory, simultaneously. In previous work, we generated 3D point cloud model of some targets byusing SfM and evaluated the generated model[1]. From the results, we consider that the improvement of visibility
is required. In this paper, we describe to improve visibility of the obtained 3D model by SfM.
2. 3D reconstruction experiment and result
For improving visibility of the generated model, we applied dense reconstruction, mesh generation andtexturing as post-processing. These processes generate texture model from 3D point cloud data by SfM. In thiscase, we used OpenMVS software[2] for post-processing. In this experiment, we used the images captured from a
video by the investigation task using remotely operated vehicle inside PCV of Unit 3 of FDNPS in 2017[3] asshown Figure 1 (Source: TEPCO). The video was recorded around the bottom of the Control Rod Drive (CRD)
housing.
Figure 2 shows 3D reconstruction result of the CRD housing and supporting clamps using SfM and theabove-mentioned post-processing. As shown in Figure 2, the reconstruction was almost done and a part of the
supporting clamp was not generated. We could recognize 3D structure of the CRD housing and supporting
clamps from the result.
4. Conclusion
In this paper, we described the result of 3D structure reconstruction based on SfM and post-processing
from the images recorded in the PCV of FDNPS. As the result, we could confirm that 3D structural
reconstruction was partly done in this experiment.
In the future work, we will attempt to improve the reconstruction accuracy.

Figure 1. Outline view of investigation in PCV[3] (Source:TEPCO)

Figure 2. Result of 3D reconstruction

References[1] T. Hanari, K. Kawabata: Information Gathering of Working Environment Based on 3D Reconstruction Using Structurefrom Motion for Decommissioning, UR2018, Th1O-13, 2018/6[2] OpenMVS: open multi-view stereo reconstruction library, https://github.com/cdcseacave/openMVS, 2015 (Jul. 2018)[3] Tokyo Electric Power Company Holdings, Inc.: Photos and Videos Library, Progress of Unit 3 PCV internal investigation(Preliminary report of July 21 investigation), http://photo.tepco.co.jp/en/date/2017/201707-e/170721-01e.html

68

J15

Sensor Database for Research of SLAM in Reactor Buildings
Taichi Yamada1 and Kuniaki Kawabata1
1Japan Atomic Energy Agency

Abstract
In reactor buildings of Fukushima Daiichi (hereinafter referred as 1F), there are still uninvestigated areas
due to the accident on March 11th, 2011. Hence the reconnaissance by using remotely operated robot in reactor
buildings is important for planning the decommissioning. Therefore, Simultaneous Localization And Mapping(SLAM) is a key technology in robotics. This research aims to construct the sensor database for the SLAM
research base.

1. Introduction
There are still uninvestigated areas in reactor buildings of 1F. For such investigation, SLAM is one of the
key technology developed in robotics. SLAM technology commits to provide the information on where the datasuch as image and dose rate was token. Thus, in order to enhance SLAM research, we construct the sensor
database for SLAM research like face image database for face recognition research base. This paper discusses
the required specification of the sensor database for such purpose and applying decommissioning works.
2. Sensor database for SLAM research
SLAM computes the location of the system, from only sensor data such as Light Detection and Ranging(LiDAR), camera and so on. If we have sensor data about the place for investigation, it is not necessary that the
SLAM works online. Therefore, the sensor database makes possible to test SLAM method easily and repeatedly.
Requirements for the database as a research base for 1F decommissioning and these resolutions are listed below.
1. Logging sensor data of the database are performed in a

Mockup field configuration

realistic workspace for assuming the workspace in 1F. In

Motion capture
Robot Computer

this reason we use mockup fields in Naraha Center for
Remote Control Technology Development [1].

Record of
Mockup Field

Log
Sensor data
Visual marker for
logging robot trajectory

2. Reconstructability of mockup fields for logging isrequired to update the mockup fieldwith 1F

Mockup field

Associateby timestamp

Time adjust
Visual marker for recording
mockup field configuration

investigation results. Because of this, we record amockup field configuration with visual markers.
3. Ground-truth of robot trajectory is necessary forquantification of test results to compare between SLAM
algorithms. Since, we regard the robot position andLog

Sensor
Database

Robot pose by
motion capture

Pose ofrobot

Computer for logging
motion capture

Computer for
motion capture

Figure 1 Logging System Summary

heading direction measured by motion capture as the ground-truth. The quantified result is provided bycomparing the robot trajectory estimated by SLAM with the ground-truth.
3. Conclusion
This paper describes the sensor database for testing SLAM algorithms to apply them to the 1F
decommissioning task. In the future works, we will examine the specification of the robot and the mockup field
and to log more sensor data with various type of mockups to construct a sensor database.
References[1] Naraha Center for Remote Control Technology Development, URL: https://naraha.jaea.go.jp/en/index.html

69

J16

3-D sensing technologies of radioactive substances andenvironmental structure inside the Fukushima Daiichi Nuclear
Power Station
Yuta Tanifuji, Yuki Sato, Yuta Terasaka, Hiroshi Usami, Masaaki Kaburagi,
Kuniaki Kawabata, and Tatsuo Torii
Japan Atomic Energy Agency

Abstract
For promoting the decommissioning works, collecting 3D structure information of the working
environment is important to understand detail status like distribution of the obstacles and radioactive substances.
In this report, we introduce experimental setup for collecting 3D point cloud data and measurement result in theexperiment in the turbine building of Unit3 of Fukushima Daiichi Nuclear Power station (FDNPS).
1. Experiment inside the turbine building of the turbine building of Unit 3
Point cloud produced by LiDAR (Light Detection and Ranging) sensor is a typical data to express andvisualize the structure of the object. Therefore, we utilize a LiDAR sensor to collect 3D point cloud data for theinvestigation in the FDNPS. We utilized a portable LiDAR (YVT-X002 made by HOKUYO AUTOMATIC CO.,LTD.) for measuring the structure of the environment. This LiDAR can measure 10360 three-dimensional points
by one scan (5frames/sec.). The LiDAR is connected to PC and ROS is utilized for collecting point cloud data.
In order to avoid occlusion problem, we attempt to compose 3D structure data by merging the measurement data
at plural locations.
Fig.1 shows integrated result with 3D point cloud data obtained by the LiDAR and radiation data by thecompact Compton camera developed by Waseda University and Hamamatsu Photonics K.K. [1]. In this figure,point cloud data appears as gray points and we can recognize 2.2m width of the narrow corridor, the existence ofa step at 8m from the entrance position and no obstacle in the air. Also, measured data by the LiDAR indicatesthat highly contaminated spot is located at approximately 10m from theentrance of the corridor. At that place, there were the hoses and it isconsidered that concentrated contamination existed around for these hoses.
2. Conclusion
In this report, we describes our attempt on 3D structure measurement in
FDNPS by using a portable 3D LiDAR. We also introduce measurement resultof 3D point cloud at the inside of the turbine building of Unit 3 of the FDNPS
and obtained 3D point cloud data was integrated with the measured radiation
data to generate 3D distribution map of radioactive substances. Such 3D
visualization gives good effort to recognize the state of the environment.
In our future works, we would aim to implement sensory data collection
and integration system to remote control system.

Fig.1 Radiation distribution map
superimposed 3D point cloud data [2]References[1] J. Kataoka et al., Handy Compton camera using 3D position-sensitive scintillators coupled with large-area monolithic
MPPC arrays, Nucl. Instr. and Meth. A, 732 (2013) 403-407.
[2] Y. Sato et al., Radiation imaging using a compact Compton camera inside the Fukushima Daiichi Nuclear Power Stationbuilding, J. Nucl. Sci. Technol., (in press).

70

ON01

NEA’s role in addressing International Challenges in Radioactive
Waste Management
Ichiro Otsuka and Gloria Kwong
OECD Nuclear Energy Agency

The nuclear community has long recognized that long-term management of nuclear waste, inparticular the disposition of high-level radioactive waste and spent nuclear fuel, is a socio-technical issue and a
Decide-Announce-Defend (DAD) approach cannot be successful. The NEA acknowledged in the 1990s that “aninform societal judgement is necessary” while technical expertise and expert confidence in the safety of geologic
disposal are insufficient, on their own, to justify to a wider audience geologic disposal. The decisions, whether,when and how to implement it will need a thorough public examination and involvement of all relevant
stakeholders. Noting the increased stakeholder involvement in the consensus decision-making process, the NEA
Radioactive Waste Management Committee (RWMC) is assisting member countries to develop safe, sustainable
and broadly acceptable strategies for the long-term management of all types of radioactive waste. In this role, the
RWMC also provides member governments and other relevant stakeholders with authoritative, reliable
information on the political, strategic and regulatory aspects of managing radioactive waste anddecommissioning. In 2016, the RWMC issued a collective statement emphasizing the importance of
implementing a holistic and sustainable management strategy for radioactive waste and spent fuel at an earlystage. In particular, this RWMC Statement presents a methodology for transforming the vision of the Committee
into tangible work activities. Following the guidance of this Statement, future activities of the RWMC, i.e.2019
and beyond, will focus on three pillars, namely (i) environmental and operational safety; (ii) societal aspects and(iii) economics. It is envisaged that a work program organized in this manner will benefit from a strongly unified
vision that fosters deeper global collaboration while maintaining a broad outlook for existing and future work
project.

The Committee, in implementing its new programme of work, maintains its fundamental purposes:fosters a shared and broad-based understanding of the state of the art and emerging issues;

facilitates the elaboration of waste management strategies that respect societal requirements;

helps to provide common bases to the national regulatory frameworks;

enables the management of radioactive waste and materials to benefit from the progress of scientificand technical knowledge, e.g., through joint projects and specialist meetings;



contributes to knowledge consolidation and transfer, e.g., through the publication of technical reports,
consensus statements and short flyers; andhelps to advance best practice, e.g., by supporting international peer reviews.

For more information on this NEA committee, visit http://www.oecd-nea.org/rwm/

71

NEA Committee on Decommissioning of Nuclear Installations andON02 TheLegacy
Management (CDLM): a new platform for international
cooperation
Ichiro Otsuka and Gloria Kwong
OECD Nuclear Energy Agency
As commercial nuclear power continues into its sixth decade, a growing number of nuclear reactors
are at or reaching the decommissioning phase. Efficient decommissioning of a nuclear facility can be a complex
process which requires careful planning and skillful selection of decommissioning strategies and dismantlingtechniques and processes. In addition to the increasing demand in nuclear decommissioning, many NEA
countries are also facing challenges in managing their legacy waste and sites as a result of historic nuclear
activities. To assist member countries to develop safe and cost-effective practices in managing these challenges,
the NEA is creating a new technical committee on Decommissioning of Nuclear Installations and Legacy
Management (CDLM).
The key objectives of the CDLM include the following:i)

provide a forum to facilitate the exchange of experience and information on policies
and practices in the areas of decommissioning and legacy management between NEA
member countries, and to advance the state of the art on technical, environmental,
policy, financial and societal aspects in these areas;ii)

advance and develop international understanding and guidance, in support of national
authorities, to address questions of common concern regarding concepts, policies,
regulatory frameworks, and strategies for decommissioning and legacy management.;iii)

share and further develop concepts, policies and approaches for reaching accepted andsustainable decisions on decommissioning and legacy site end-state aspects aiming at
reducing the burden on current and future generations from facility decommissioning
and remediation; in particular to keep under review the developments in the
optimisation of decommissioning strategies bringing in balance technical, economic,
safety and ethical aspects;

iv)

enhance worker and public safety, and environmental protection by developing
practical radiation protection guidance to manage nuclear legacy sites anddecommissioning;v)

further develop approaches for the estimation of decommissioning project costs byimproving understanding of the risks associated with financial consequences in regards
to cost estimating and financing; and decommissioning planning for uncertainties.

72

ON03

NEA’s Support to the Fukushima Daiichi Decommissioning andAddressing Public Concerns
Ichiro Otsuka, Edward Lazo and Kentaro Funaki
OECD Nuclear Energy Agency

Seven years after the Fukushima Daiichi nuclear power plant accident, the Japanese government andTokyo Electric Power Holdings, Inc. (TEPCO) are shifting their focus to strategic planning for long-term
challenges related to the decommissioning of the damaged reactors. The international community has been
helping to address the unprecedented challenges of managing the accident facilities. The NEA is playing a key
supporting and coordinating role in the international community, in particular in the area of radioactive waste
management and the evaluation of the conditions and location of fuel debris.
The NEA Working Party on Decommissioning and Dismantling (WPDD), under the aegis of the
NEA Radioactive Waste Management Committee (RWMC), provides a good knowledge base for the mid- tolong-term planning of decommissioning the Fukushima Daiichi NPP. In 2017, TEPCO joined the NEA
Co-operative Programme on Decommissioning (CPD), a joint undertaking gathering and sharing experience
gained from recent decommissioning projects among NEA member countries, to learn and exchange practicaldecommissioning experience in order to achieve safe, economical and environmentally sound decommissioning.
The NEA Committee on the Safety of Nuclear Installations (CSNI) has been assisting member
countries in developing the required scientific and technical knowledge base for maintaining the safety of
nuclear reactors and fuel cycle facilities. The NEA Joint Project Benchmark Study of the Accident at the
Fukushima Daiichi Nuclear Power Station (BSAF) assists member countries to improve severe accident codesand knowledge of the reactor core status so as to prepare for future decommissioning. The CSNI Senior Expert
Group on Safety Research Opportunities Post Fukushima (SAREF), using information gained from post-accident
examinations of Fukushima Daiichi, identified safety research knowledge gaps to advance safety. The CSNI
Working Group on Analysis and Management of Accidents (WGAMA) has accumulated
knowledge and expertise relevant to decommissioning strategy planning for the Fukushima Daiichi NPP.
With respect to radioactive waste management, the NEA Expert Group on Fukushima Waste
Management and Decommissioning R&D (EGFWMD) s provided strategic waste management advice to Japan
in managing post-accident waste of

Fukushima Daiichi. In 2018, the RWMC is also establishing an Expert

Group on Characterisation Methodology on Unconventional and Legacy Waste (EGCUL) whose aim is to focus
on the development of an integrated methodology for characterizing unconventional radioactive waste.
The Committee on Radiological Protection and Public Health (CRPPH) continues its risk
communication through dialogues with the International Commission of Radiological Protection (ICRP) and theNEA International Workshop on Post-Accident Food Safety Science Workshop.
Recent activities of these groups are further explained in separate posters in this conference.

73

ON04

Strategic Approach to Managing Fukushima Daiichi Waste
Ichiro Otsuka and Gloria Kwong
OECD Nuclear Energy Agency

In response to the urgent need to safely manage radioactive waste as a result of the Fukushima
accident, the NEA Radioactive Waste Management Committee (RWMC) established an expert group to advise
the Japanese Government on strategic management of large quantities of post-accident waste that has complex or
unknown properties. The NEA Expert Group on Fukushima Waste Management and Decommissioning R&D(EGFWMD) undertook this three-year initiative in 2014 to assess the technical challenges faced by the waste
managers, regulators and other stakeholders. In this project, the EGFWMD examined the physical and chemical
nature as well as the radiological characteristics of the waste. Specific issues related to feasible wasteconditioning and volume reduction to facilitate interim storage and future disposal, decontamination of the
facilities were also assessed. In December 2016, the expert group concluded its work and published a project
report that provides technical opinions and recommendations on efficient post-accident waste management
strategy and R&D at the Fukushima Daiichi site. The report concluded that characterisation and categorisation of
post-accident radioactive waste are among the most difficult challenges for waste management at the Fukushima
Daiichi NPP in the near term. A strategic approach which includes a sampling and characterization plan based on
statistical approaches, calculation methods, and a review and evaluation process for the data obtained should be
developed to manage the complex characterisation process of the post-accident waste. The group also
acknowledged that international knowledge and experience in managing legacy waste, accident waste and otherpertinent examples would be valuable in address challenges in managing the Fukushima Daiichi waste.
In continuing its technical support to Japan, the RWMC is creating a new Expert Group on
Characterisation Methodology of Unconventional and Legacy Waste (EGCUL) whose aim is to share knowledge
and experience in characterisinga large amount of unknown waste. A key deliverable of the EGCUL is todevelop an integrated methodology for the characterization of unconventional and legacy waste of Japan. The
EGCUL is anticipated to include specialists from Japanese organisations such as the Japan Atomic Energy
Agency (JAEA), the Nuclear Regulation Authority (NRA), TEPCO and the NDF, alongside experts from aroundthe world with experience in radioactive waste characterisation. More detailed outcomes of the EGFWMD and
the work-plan of the EGCUL are presented in this poster session.

74

)

IN CAN MELTER for the conditioning of high and medium-level and longlived waste from D&D operations
Christophe GIROLD 1, Christine GEORGES 1, Stéphane CATHERIN2,
Thierry PREVOST 3, Guillaume LECOMTE 4

Abstract

1 CEA, Nuclear Energy Division, 2 ANDRA, 3 AREVA NC, 4ECM Technologies“IN CAN MELTER” process is being developed by CEA for its own waste coming from D&D
operations, when “glass-like” encapsulation is required for further disposal but when cost
reduction is an aim. Low temperature solidification with few additives would be of great interestis the encapsulation of mixed effluent waste such as zeolite, co precipitation sludge or powderof fuel debris.

• A robust, simple and versatile in situ vitrification process adapted to D&D waste• Adapted to high or medium level waste,• The container is used as a crucible ("in can"), renewed with each batch: no need todrain the glass and corrosion is limited by the batch limited time,• Waste package complying with existing routes and/or on-site storage facilities (CSD),• Heated by a simple and robust resistance heated furnace, inexpensive, easy tomaintain, easy to implement and offering a great homogeneity to the melted mixture.
This allows a metallic fraction in the canister and an excellent control of temperature.
• Flexible enough to accommodate uncertainties on waste composition and however
produce a confinement with quality description, with source term calculation
methodology for long term storage.
• Small amount of secondary waste (no components with molten glass remains afteruse ) and minimum investment and operating cost.

• Tested for High Level Cs deposits from Marcoule UP1 shop under decommissioning• Application to effluent treatment waste (zeolites, sludges, …) on goingLab scale glass formulation adapted to the in can process with a 70 -80 %waste loading of zeolites, carbonated and ferric sludge, sand, silicotitanate.

• In the frame of a French founded program (DEM&MELT project), a new prototype isunderbuilding to develop a vitrification tool adapted to D&D•

All the process being include in a small footprint frame tobe easily forward-deployable on an existing D&D site•

Capacity : With a 300 kg (x3 / in can) CSD canister•

Producing a small amount of secondary waste•

Designed to cost as aimed to be dismantled after
treatment operation

75

F02 Development of Laser Cutting and Dust & Fumes Collection
technologies for Fuel Debris Retrieval
D. Roulet1, C. Georges2, C. Chagnot2, C. Journeau2, E. Porcheron3
1

ONET Technologies, France, 2 CEA, French Alternative Energies and Atomic, 3 IRSN

Abstract
Among the numerous challenges towards the decommissioning of Fukushima Dai-ichi reactors, removing fueldebris is for sure on of the tougher ones. A French consortium “ONET Technologies, CEA and IRSN” isdeveloping laser cutting technologies for fuel debris with the implementation experience in UP1 French
reprocessing plant. In addition to laser cutting development in air and underwater, dust and fumes collection
technologies are being developed to capture the aerosols released during the debris laser cutting and theircollection.

1. Cutting Technologies
The variety of fuel debris location, material and shape will needvarious cutting techniques to be developed. Laser cutting
technologies are capable to overcome many configurations withproven performance, and great benefit with regards to remoteapplication. Research for application to Fukushima environmentis on-going with promising results.
Several technologies are being developed (in air, underwater,
emerging or non-emerging) with good results and performance fo
Simulation of fuel debris non-emerging cutting by laser

future application to Fukushima Fuel Debris.
2. Knowledge of dust and fumes generated during cutting
Laser cutting being a thermal cutting method, aerosols, gases and particles are emitted during the process andthe characterization of such secondary outlets is of paramount importance. The conducted trials and associatednumerical simulation enable estimation of released radioactive elements, aerosols size and gases, which are an
important input data for collection and treatment systems definition and estimation of release to environment.
3. Collection Technologies
In parallel to cutting technologies, collection systems are being
developed to capture the dust and fumes generated during the cutting.
Such technologies are to be deployed remotely, and in case ofunderwater cutting spray systems for capturing aerosols are underdevelopment.
Conclusions
Laser cutting technologies associated with collection technologies are key processes for fuel debris retrieval, but
could be also used for other systems cutting as part of Fukushima-Daiichi decommissioning as CRD or RPVinternal structures.

76

F03

Potentialities offered by New Binders in conditioning of
problematic low- and intermediate-level radioactive waste
David Lambertin, Christine Georges, Fabien Frizon, Céline Cau dit Coumes,
CEA, Nuclear Energy Division

Abstract
Some waste produced by decommissioning operations may chemically react with cement phasesor mixing water, affecting hydration and eventually resulting in the deterioration of the waste
form in storage or disposal, for instance by swelling and cracking. Three kinds of inorganic binders
offer better chemical compatibility, thus avoiding pre-treatment or increased volumes of waste:calcium aluminate and sulfo-aluminate cements, magnesium and calcium phosphate cements,
alkali-activated binders and more specifically geopolymers.

Calcium Sulphoaluminate Cements (CSAC)
CSAC clinker are produced by firing mixtures of limestone, gypsum and bauxite.
Foreseen applications of CSAC for waste conditioning ranges from Cementation of borate or heavy
metals such as lead, cadmium, zinc, and trivalent chromium or to immobilization of Aluminum.
Due to the different cement chemistries, the rate of hydration of CSAC may be less affected by strong
retarders of OPC such as heavy metals or borate ions and its mineralogical structure is favorable forwaste immobilization, offering possibilities of ion substitutions.

2

Phosphate cements are the main representatives of acid-basecements. Waste immobilization results from two processes:
precipitation of many contaminants, particularly actinides, as
phosphates with very low solubiphysical encapsulation in a dense
phosphate matrix. Good results have been reported for several
types of wastes, e.g. low-level debris wastes contaminated by
137Cs, highly saline effluents or metallic aluminum by decreasinghydrogen corrosion evolution (Figure 1).

H2 evolution (L/m of Al)

Magnesium Phosphate Cements (MKP)

Time (days)

Figure 1: Impact of addition of LiNO3 (2 wt.%)

Geopolymers: alkali-activated binders
Alkali-activated binders are made by mixing solid
aluminosilicates, such as fly ash, metakaolin, various clays
usually activated by heat or blast furnace slag, with anactivating solution comprising high concentrations ofalkali hydroxide and / or polysilicate. The reaction
product, formed according to a dissolution/polycondensation process, exhibits an X-ray amorphous
network structure.
It is possible to vary the properties of the produced alkaliactivated binders over a wide range, and to tailor them to Figure 2: Encapsulation of Mg Figure 3 : Oil/ geopolymer composite
cladding in geopolymer
specific requirements like high compressive
strength, low permeability, low shrinkage.
Geopolymers offer great potential to stabilize/solidify hazardous waste like heavy metals, alkali-earth,alkali ions and magnesium alloy encapsulation (Figure 2).
Geopolymers present a new way for direct organic liquid waste encapsulation by stabilizing emulsionof oil in an alkali silicate solution and being directly solidified by addition of metakaolin (Figure 3).
Understanding the chemistry of cement – waste interactions, and their consequences on thephysical properties of the solidified waste forms, including their long-term evolution, has to befurther considered for their acceptance.
77

F04

Foams and gels for decontamination of structure and concrete walls
Alban GOSSARD 1, Sylvain FAURE1, Célia LEPEYTRE1, Agnès GRANDJEAN1, Christine
GEORGES1, Jérôme ROUSSELLE2, Lucien GALLO2, Thierry VARET3, Fabrice Moggia 3
1
CEA, French organization for nuclear and alternative energies, Nuclear Energy Division,2
FEVDI, 3 ORANO

Abstract
The specificity of gels processes developed by CEA and ORANO for the decontamination of structure are that they are
sprayable and vacuumable. The main advantage of these gels is that the final secondary waste is a non-powdery solid wastethat can be recovered by brushing or vacuuming. This self-drying technology is well adapted to treat large surface as walls or
floor but in case of complex geometry or of vessel, CEA and ORANO recommend the use of gellified foam. These bothtechnologies give unique tools for treatment of structure with the objective to minimize to production of secondary waste.
CEA develops new self-drying gel since a couple of years designed to remove bitumen stain contamination, or specific to thedecontamination of concrete walls.
1. Introduction
The decommissioning of nuclear facilities for end-life management is a major issue in present and future years. However,their prior decontamination is a necessity to avoid the generation of high volume of contaminated waste. These operations
will conduct to the de-categorization of structures in order to valorize them in traditional deconstruction sectors or torehabilitate them for a new utilization. Among the available decontamination techniques, we must chose technologies thatminimize the secondary waste while being highly efficient.
2. Specification
In this context, self-drying vacuumable gels and gellified foams technologies
are mature processes developed a few years ago.
The gels are formulated by mineral colloidal particles (alumina, silica ortitanium oxide) dispersed in an aqueous solution containing the reactants(acid, base or oxidant). This reactant is chosen to remove the contamination
by attacking the substrate and then transferring the radio-contaminant fromthe surface to the gel flakes after drying. The specific rheological propertiesof these gels make them sprayable on the contaminated surface and their
thixotropic behavior allows them to recover their viscosity immediately
after spraying and adhere to the surface. After a few hours of drying, the gel
has turned onto a non-powdery solid which can be easily brushed or vacuumed. The main advantage of this process isthat no liquid effluents are produced. These gels are commercialized by the FEVDI company with a large range of
specific gels for different applications depending on the nature of the surface to be treated and the nature of thecontamination. FEVDI can also purchase specific sprayer for this application.
A gellified foam has a long lifetime and then allows the decontamination of complex shape (vessel containing tubes,
complex geometries…) by maintaining during a few hours the contact between the contaminated surface and the aqueous
reactant containing into the foam. These kinds of foam are formulated by mixing into an aqueous phase a very small
amount of surfactant, a viscous polymer gel (the gelling reactant) and the reactant (acid). This process produces lower
amount liquid effluents than batch processes with a volume reduction by a factor of ten.
3. New developments
To spread off these technologies beyond their traditional use, we develop two kinds of new formulations:Hybrid gel well designed for the treatment of organic layers and specifically the removal of bitumen smudges [1]. The
replacement of aqueous solution classically used in vacuumable gel by an organic bio-solvent leads to the definition of
a self-drying gel able to dissolve fast and efficiently contaminated organic layer and specifically bitumen;
Wet pastes specifically designed for the decontamination in depth of porous surface, and then can be applied for
concrete walls decontamination. This is a two steps treatment. First water contained into the paste diffuses to theporosity of the contaminated substrate (concrete) to solubilize contaminants. Then during the drying phase, water andsolubilized contaminants flows back from the solid to the drying paste allowing the extraction of the contaminants byadvection.
4. Conclusion
Gellified foam and self-drying gel processes are mature processes for the decontamination of solid waste. They can be used
for the treatment of structure, or complex shape as vessels, tube…These technologies still go on by the improvement of theirimplementation, and by the development of new kind of formulations.
References[1] Gossard A., Frances, F. “Vacuumable gel and method for eliminating contamination contained in an organic layer on

the surface of a solid substrate” WO2018024990A1[2] Faure S., Fournel B, Fuentes P. “Composition, foam and method for surface decontamination” WO2004008463

78

F05

Fission Products, Uranium and Plutonium measurements ingases and corium after a PWR severe accident
L.Boucher1, P.G. Allinei2, H.Hupont2, F.Jallu2, C.Roure2,

1 CEA, Nuclear Energy Division, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance, France
2 CEA, Nuclear Energy Division, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance, France

Following the Fukushima accident, a lot of recommendations were drawn by internationalorganizations (IAEA, OECD, NUGENIA,etc.) in order to improve the safety in such accidental
conditions and limit consequences. One of these recommendations was to improve therobustness of the instrumentation, as well as to better determine the Source Term involved ina nuclear accident.
In this context, CEA designed two systems: one to assess the reactor building gas releases incase of severe accident and one to measure Uranium and Plutonium contained in spent fuelhulls.
The first system, designed in collaboration
with Institut Max von Laue-Paul Langevin(ILL) consist in a gamma spectrometry
system devoted to the measurement of
Fission Products gases. It implements ahigh sensitivity and high energy resolution
Hyper-Pure Germanium detector (HPGe).
It is robust and able to operate under thesevere environment encountered after
accidental situations.
The gamma spectrometry measurement
system allows the measurement of thenumerous gamma ray lines emitted by theshort live fission products and then to detect
a situation of re-criticality
The second system, called “PACCMAN” is a transportable systemallowing an active neutron measurement of U and Pu fissile isotopes
in a 25 liter waste drum. The measuring setup is designed to be
docked to a hot cell (PADIRAC system). It is made on the basis of a
RD10-type castle around which measurement instrumentation is
implemented. It uses a neutron generator and 3He counters to reacha detection limit mass of about 100 mg 239Pu equivalent in 15
minutes. Under several conditions, this system could be efficiently
used to measure a fissile content in small quantities of corium toassess the criticality risk.
PACCMAN System

The Nuclear Measurement Laboratory is specialized in non-destructive methods, such as
gamma-ray spectrometry, passive neutron measurement, active neutron interrogation, neutron
activation analysis, high energy X-ray imaging systems for radiography and tomography
applications. It concerns expertise, R&D, experimental assays, modelling, perfecting inlaboratory and in-situ measurements. The applications are for example nuclear fuel,radioactive waste, dismantling and decommissioning, activity and fissile masses
characterization, NRBCE security

79

F06

Sorbmatech sorbent for Cesium removal fromcontaminated liquid effluents
Yves BARRE 1, Audrey HERTZ, Jérémy MASCARADE1, Agnès GRANDJEAN1, Christine
GEORGES1, Eric LOURADOUR2, Didier DHALLER2, Thierry PREVOST 3
1
CEA Nuclear Energy Division, 2 CTI,3 ORANO

Abstract
CEA, ORANO and CTI collaborate since a couple of years for the development of a new family of inorganic
sorbent, named Sorbmatech®Cs, highly efficient for the decontamination of liquid effluents and suitable as final waste
form. A new sorbent developed in this context was able to extract 137Cs selectively (Kd equal to 105 to 106 ml.g-1
depending on the salinity of the effluent), with a very fast kinetics (equilibrium reached in a few minutes), either in stirred
reactors or in column processes.
1. Introduction
Permanent needs to reduce radioactive species release into the environment requires continuous technology improvement
in liquid effluent treatment process, minimizing secondary solid waste production. Not only accidental situations (such as
in Fukushima Daichii), but also nuclear energy production industry, nuclear fuel recycling plants and end-of-life nuclear
facilities decommissioning, are at the origin of contaminated aqueous liquid waste production. Among all radionuclides
that need to be treated, 137Cs is considered as the most abundant and hazardous element due to its presence in many types
of waste and its relatively long half-life (30 years).
2. Specification
In this context, we have develop an efficient sorbent based on porous silica support (pores size a few to a few ten nm)
loaded with hexacyanoferrate nanoparticles (HCF). The general formula of HCF is KxMy[Fe(CN)6]z where M, a bivalenttransition metal ion, can be Cu, Ni or Co. The selectivity of these various HCF varies with the pH of the effluent versus
the nature of the transition metal, so we can simply adjust the efficiency of the sorbent depending on the HCF inserted into
the porous silica support. This fabrication approach can be applied for various support, from porous grain to filtration
membranes.
To date the partnership between CEA-CTI-ORANO leads to the development of a simple and low cost synthetic route
allowing the fabrication of a large-scale manufacturing of this product. Two products have been developed so far: (1) a200-500µm grain size (S250g) well adapted for column processes (low pressure-drop) and (2) a <100µm grain size
(S250p) well adapted for stirred reactor uses. The evaluation of the efficiency of these pre-industrial products, from preindustrial mockup (1L column, 3L reactor) to lab-scale with 137Cs doped effluents show very high selectivity even insaline solution (KdCs from 105 to 106 mL/g). A wide range of effluents can then be treated: high, medium and low
activity, pH from 2 to 10, high salinity (saline, mineral or sea water). But the most interesting property of this product is
its really high sorption kinetics even with large grain size suitable (200-500µm) for column process. This allows the useof high flow rates of treatment (up to 10m/h) with low pressure drop and deep breakthrough curve. Smaller granulometry
silica grain size (<100µm grain size) enables both good dispersion and easy sedimentation in stirred reactor.
3. Developments
100Kg of SORBMATECH®Cs for column treatment (200-500µm grain size) was produced
and the pre-industrial manufacturing of an equivalent product (Grains <100μm), usable instirredreactor,
is
also
being
evaluated
up
to
10Kg
production.
The outlook for SORBMATECH® sorbent series is not only the widening of the pH
range of the potentially treated effluents (by modifying the exchanger for increasing
performances in acidic medium, and by modifying the support for holding in basic medium),but also the widening of radionuclide range that can be treated, with the first line of sight of
90Sr.
Offering customers a complete solution, for effluent treatment using dimensioning
model for implementing column process at industrial scale, and also for long-term
stabilization of generated waste, directly suitable with vitrification process for example, areunder progress.
4. Conclusion
The insertion of nanoparticles HCF into porous silica support leads to a high selective Cs-sorbent with fast kinetics ofadsorption suitable for industrial effluent treatment, whatever the technology, stirred reactor or fix bed. These products arein a pre-industrial manufacturing step and will be complete to treat other RN such as Sr.
References[1] Caroline Michel, Yves Barré, Mehdi Ben Guiza, Caroline de Dieuleveult, Laurent de Windt, Agnès Grandjean, Chemical Engineering
Journal, 339 (2018) 288-295.
[2] Caroline Michel, Yves Barré, Laurent De Windt, Caroline de Dieuleveult, Emmanuelle Brackx, Agnès Grandjean, Journal ofEnvironmental Chemical Engineering, 5 (2017), 810-817.
[3] Agnès Grandjean, Yves Barré, « procédé de préparation d’un matériau solide nanocomposite à base d’hexa- et octacyanométallatesde métaux alcalins », patent WO2016038206 A1
80

F07

Virtual Reality: a way to prepare andoptimize operations in decommissioning
projects
Caroline CHABAL, Yves SOULABAILLE, Christine GEORGESCEA, DEN, [SDTC, LSTD], 30207 Bagnols-sur-Cèze, France

CEA must manage the end of life of its nuclear fuel cycle facilities built in the sixties.
These high priority actions led to the creation of a dismantling R&D division which provides
innovative tools, including in-situ radiological characterization, remote handling and cutting, andintervention scenario simulation.
Simulation is an ideal means of visualizing and therefore better knowing highly radioactive
environments where humans cannot enter, of testing different technical alternatives, and of training
workers prior to interventions.
A VR simulation can be defined as an interactive and immersive simulation that enables the user tointeract with a computer-simulated environment. VR environments, mostly based on visual immersion
displayed through stereoscopic devices, can also include additional sensory information, such as soundor touch.
New PRESAGE immersive room in Marcoule, based on audio, tactile and visual immersion, provides
a useful support to verify pre-defined scenarios and to design alternative solutions if necessary.

iDROP, A VR SOLUTION FOR NUCLEAR OPERATIONSiDROP is a new software developed by the CEA,
and offers a global approach taking into accountdose rate, remote handling and human operations.
iDROP offers 5 modules:- The physics module calculates forces andcollisions and avoids object penetration.
- The robotics module includes the mainindustrial robots (about 40) which can operate
in nuclear facilities.
- The virtual human module enables thesimulation of manual operations taking into
account the human degrees of freedom andfeatures.
- The dose rate module integrates
MERCURE calculation code and estimatesdose rate received at any point taking into
account the radiological data.
- The immersive module enables the simulation to
be connected to immersive VR and AR devices.

81

Chain for Decommissioning in France
F08 Supply
Christine GEORGES*, Anne Sophie DEFAY**, Serge PEREZ *, Serge BOUFFARD***,
Claudia ROCHE****(*): CEA, DEN, France, (**): AIFEN; (***): Nucléopolis; (****): Nuclear Valley

Decommissioning benefits from France’s nuclear supply chain that was constantly in action since thesixties, thanks to its uninterrupted nuclear programs. Nuclear industry is France’s third-largest industry,
with over 2,500 French companies, nearly 220,000 employees, a yearly turnover of over €46 billions
and added value reaching €15 billion, R&D investment ranking 4th amongst innovative industries.
CSFN (French nuclear governmental Steering Committee) coordinates the actions of the French
nuclear stakeholders in the unifying topics: innovation, training, exports, etc.
In 2013, CSFN created AIFEN to support the French nuclear team abroad and specifically its Small andMedium-sized Entreprises (SMEs). AIFEN designed the Nuclear Expertise from France logo, which is a
tailored brand illustrating the French nuclear team and organizes, every two years, the World Nuclear
Exhibition (WNE), meeting point of the global nuclear community, gathering around 10 000 visitors,
from more than 70 countries.

CSFN aims to strengthen ties and partnerships between the different bodies, including Nuclear Safety
Authority, public authorities and related governmental Departments, Certification and qualification
bodies, Nuclear operators and research institutions (EDF, CEA, AREVA, and ANDRA), Suppliers of nuclear
technologies, AIFEN and regional groups (NUCLEAR VALLEY, PVSI, Nucleopolis, etc.), Trade groups andlabour unions (GIIN).
Set up in 1959, GIIN - French Nuclear Suppliers Association is an independentand non-governmental organization acting to support the French supply
chain. GIIN represents over 450 SMEs and mid-caps companies with expertisein new build, plant operation, fuel cycle, dismantling & decommissioning, andresearch reactors.
Nuclear Valley, founded in 2004, national cluster dedicated to civil nuclear in
France, with more than 230 companies, half SMEs and Mid-Caps, operatingin civil nuclear. It provides support innovation and training throughcollaborative innovation projects in the field of metallurgy, big components,
non-destructive testing, maintenance and decommissioning and civil work.
PVSI, created in 2014 around Marcoule CEA site, combines 24 companies
from industry, research and education, aiming at forging a French sector of
excellence in the field of Decommissioning, through increased ETIs to large
account resources, training and professional retraining of the operators.
Nucleopolis, nuclear cluster of Normandie Energies, created in 2010,
federates about 100 companies: major industrial groups, high-performance
SMEs, renowned research and training institutions (CEA, GANIL, Universities,etc.) and care centers.

82

*%

UK | Japan
Natural Partners in Nuclear DecommissioningⱥᅜࡣཎᏊຊᗫṆᥐ⨨ࡢ᭱㐺࡞ࣃ࣮ࢺࢼ࣮

1. The NDA: Supporting Innovation 㠉᪂ࢆᨭ࠼ࡿ
The NDA ensures the safe and efficientclean-up of the UK’s nuclear legacy from the
UK’s pioneering post-war programme.
Covering 17 sites, the NDA’s mission includes
taking all the structures apart, dealing withthe waste and ultimately restoring sites.
One of NDA’s responsibilities is to ensure theadequate amount of R&D is carried out todeliverthefulldecommissioning
programme.
Many ‘never-done-before’
projects require significant innovation andnovel engineering approaches.
R&D is required to ensure technicalchallenges are understood, to encourage
development of innovative ideas and enable
the demonstration of new technologies
followedbysuccessful
on-siteimplementation.
Funding
NDA fund R&D through two main routes:• Directly through projects commissioned by the
NDA to support strategy (DRP)• Via Site Licence Companies on specific site
projects where spend is part of main annual
budgetSupport is provided on a working level by the
Nuclear Waste and Decommissioning
Research Forum (NWDRF).
NWDRF
promotes collaboration across the NDA
estate and the UK whilst ensuring innovation
is delivered efficiently, cost effectively,
quickly and with less duplication.
Case Studies: Innovative Remote
Technologies
In 2017, NDA and Innovate UK launched acompetition named ”Integrated Innovationin Nuclear Decommissioning” inviting
proposals for projects that could deliver
solutions to decommissioning challenges at
Sellafield. Submissions were invited from allindustrial sectors.
The response from suppliers and academic
institutions was so promising that additional
funding was made available by BEIS.
Suggestions came from consortia made up ofSMEs, large corporations and universities,
belonging to industries as diverse as seafishing and medical imaging.
Five promising ideas have been shortlisted
for the final stages. Each will receive up to£1.5m to build prototype demonstrators fortesting in a mock up radioactive
environment.

The DRP is the main mechanism by which
the NDA directly funds R&D. R&D proposals
are developed with three key drivers in
mind: Inform Strategy, Deliver Innovationand Maintain and Develop Skill.

Robot Laser Cutting
Laser cutting minimises spread of contamination
during operations. A commercially available
robot was fitted with laser technology andtested in a radioactive environment. To allow
laser cutting to be deployed remotely in a safeand efficient way, trials and development work
included metal cutting and testing of control
systems and ventilation.
The potential to save money and reduce risk has
been seen and the learning is now being used
for alpha cutting.

Robotics: Nu-Decom

Plant Operations using remotely-operated tools:
Utilisation of the BROKK 150C with a Pecker
Attachment for Decommissioning activities.
Elephants to Ants: Innovation in Integration

Elephants to Ants is creating a real-time sensed,
hardware-in-the-loop VR Operating System fordecommissioning, to enable engineers to control
teams off-the-shelf robots with confidence, speed andprecision.
Integrated Innovation for Nuclear Decommissioning

A nuclear-centred integration approach, which
innovates a new modular control & automation
strategy that can be proven & validated within the
nuclear environment.
Sellafield In-Cell Decommissioning System (SIDS)

Governance
Advice and guidance for co-ordinating
nuclear decommissioning R&D is provided by
the NDA’s Research Board. Independently
chaired, membership comprises senior
representatives from UK government, UK
industry, NDA Estate, regulators and fromoverseas.

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UK | Japan
Natural Partners in Nuclear Decommissioning
英国は原子力廃止措置の最適なパートナー

2. Remote Operations: Under the Water 水中技術
This poster focuses on the use of remote handling technology in the decommissioning and demolition of nuclear
facilities in the UK. Remote handling solutions are providing effective, more efficient solutions to challenges faced
across the NDA estate. Underwater-environment technologies range from novel navigation and samplingtechniques to commercially available Remote Operated Vehicle (ROV) platforms. These technologies enable accessto areas that would be otherwise impossible to work within without huge investments of time and money required
to drain down and shield to allow traditional decommissioning techniques to be used safely.
A.

INTEGRATE | 統合

Underwater surveys at Sellafield - FORTISFORTIS sonar technology has been deployed in legacy ponds at Sellafield, where it would be too
difficult or dangerous for human entry. The system uses point cloud technology to generate 3D CAD
models that can be interrogated to allow better understanding of the geometries within a facility
and has enabled detailed planning of future decommissioning works.

B.

MODIFY | 修正

ROVs have proved to be efficient in taking on a wide range of activities, from pond liquor sampling andcharacterisation to underwater inspections, surveys and maintenance tasks.
ROV platforms at Sellafield’s Legacy facilities - Rovtech
The Nano platform is a modular system designed to allow fast disassembly and cleaning while wearing
multi-layer protective clothing. Variable buoyancy systems are designed to operate in areas that have large
quantities of silt and sludge that traditional depth control systems can disturb, spreading contaminationand obscuring vision. They were deployed at
In-pond operations at Sellafield – James Fisher Nuclear (JFN)JFN have built up a vast experience base of operating ROV systems within the UK’s nuclear
facilities.
Larger ‘Work Class’ ROVs regularly carry out successful size reduction and remediation
tasks at Sellafield’s FGMSP. They have successfully employed a series of tools developed
by JFN, ranging from manipulator arms, saws, cutters through to scoops and dredgers.

C.

DESIGN & DEVELOP | 設計と開発

Underwater sampling at Sellafield - NNL
Sampling sludge and sediments left over from nuclear
waste while under water is a difficult task due to thefact that disruption of the environment in and aroundthe sludge can change its properties providing false
data. NNL has developed a series of custom toolingthat minimises this distubance and prevent the uptakeof additional liquor.
Underwater sampling at Bohunice NPP - Wood
The submersible Crawler Systems utilise simple but robust hydraulic and pneumatic drive
systems to minimise the need for hardened electronics when operating in high dose rateenvironments therefore greatly increasing the systems’ operating lifetime. These systems havebeen used in decommissioning activities across Europe.

Conclusion
The ability to access, monitor, investigate and operate equipment and plant under water within a nuclear facility
can be critical for successful and safe decommissioning. Remote operation technologies have been used in the UK
to save money and reduce the risk of working in these environments while also allowing physical operations to be
carried out in areas which have not been easily accessible. The UK continues to benefit from these technologies
and is proud to share their experience with Japan.

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UK | Japan
Natural Partners in Nuclear Decommissioning
英国は原子力廃止措置の最適なパートナー

GB03

3. Remote Operations: On the Ground 地上でThis poster focuses on the use of remote handling technology in the decommissioning and demolition of nuclear
facilities in the UK. Remote handling solutions are providing effective, more efficient solutions to challenges faced
across the NDA estate. Remote technology provides a safer, controlled barrier between the workforce and manyhazards present on a decommissioning site including radiation, contamination and health and safety issues. The
technology can be integrated across a wider waste retrieval system to ensure that safe, and appropriate tools arebeing used and reduces the lifetime of these retrievals projects.
A.

NNL Kuka Robots for
Sorting & Segregating

INTEGRATE | 統合

Commercially available robotic technologies arereviewed for use and tested under a range offunctions including teleoperation and automatic
modes for standard tasks. This promotes safety
benefits and supporting evidence to underpin thesafety case for the operation of the equipment.
B.

MODIFY | 修正

Projects with commerciallyavailable robotic technologies
have adapted specially designed
components to be fit for purpose
withinvarioushostile,unpredicted environments.
This provides highly optimisedand engineered solutions thatmeet all the client requirements,
de-risk the design from afunctional aspect and reduce thefull life-cycle decommissioning
costs.
C.

Dounreay shaft intervention
platform – Veolia Nuclear
SolutionsThis project is unlike anything
undertaken in the UK before,possibly the world.

DESIGN & DEVELOP | 設計と開発

Berkeley Chute Silo –AquilaIn-silo size reduction,
retrieval and transfer
of waste to shieldedcontainers for export toon-site storage facility.

Brokk ROV at
Trawsfynydd NPP

Harwell & Hunterston HiRAM &GEM assay - Nuvia
High Resolution Assay Monitor(HiRAM) is a trailer-mounted
high-res gamma spectroscopy
semi-automatic assay system.
Gamma Excavation Monitor(GEM) system is a remotely
operated gross gamma system.
Both used successfully to assay
contaminated land and rubble.

Hot Spot locators at Fukushima Daiichi –Innovative Physics Ltd (IPL)MHI Fuel Debris Retrieval - Veolia Nuclear
SolutionsRadiation imaging systems can map
contamination in hazardous environments.
Integrating a pixelated sensor andphotomultiplier tube with an opticalcamera, these systems have been proven tobe effective for decontamination anddecommissioning purposes, for example
aiding the post event clean-up operations
around the Fukushima Daiichi Power Plant.

This bespoke remotely-controlled robotic arm system is
desgined to investigate fuel debris in Fukushima Daiichi
Unit 2 Reactor. The solution combines knowledge
gained through work in the nuclear fusion
section and demanding environments
at Dounreay and Sellafield and wasdeveloped with UK supply chain
including NIS and Createc for enclosure,Conclusion
detector and manufacture.
With experience from a range of sites and projects, the UK industry has gathered experience in remote handing,sort and segregation by looking at each individual situation and creating solutions with either commercial off the
shelf equipment, modified industry robotics or bespoke systems. UK industry is pleased to be working inFukushima and transferring the experience and knowledge gained from these projects.

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GB04

UK | Japan
Natural Partners in Nuclear Decommissioning
英国は原子力廃止措置の最適なパートナー

4. Remote Operations: In the Air 気中技術
This poster focuses on the use of remote handling technology in the decommissioning and demolition of nuclear
facilities in the UK. Remote handling solutions are providing effective, more efficient solutions to challenges faced
across the NDA estate. Access to high places to inspect and to do work is a recurring challenge in nuclear
decommissioning. Traditionally, this could be done using scaffolding or mobile work platforms, but these haveseveral limitations: they expose workers to hazards (both radiological and working at height); they can take up a lotof space, where space is precious, and they can take time to set up and operate making them costly. The UK has
developed remotely-operated technology that can address some of these limitations.
A.

MODIFY | 修正

UAV at Sellafield – James Fisher Nuclear (JFN)JFN has adapted and deployed commercially available internal and external UAVs and sensors
to collect critical data quickly, more cost effectively and with much greater safety than by
conventional means. UAVs have delivered a range of instruments and tools including highresolution digital and thermal cameras, radiation probes, LASER scanners and air samplers into
restricted access or high hazard areas. This flexibility and versatility has enabled easy access tovital data to give decision-makers a more complete picture. All data has high level encryption forcomplete security in any eventuality including complete loss of the UAV.

B.

DESIGN & DEVELOP | 設計と開発
RISER UAV at Fukushima Daiichi – Createc

C.

COTS UAVs use GPS for control, which means they can only be used outside where they
have a view of satellites. To extend the use of UAVs to internal structures, Createc
developed a lidar navigation system for UAVs. The system was used to make ‘RISER’ incollaboration with UAV development specialist Bluebear Systems. RISER has been
operated at Fukushima Daiichi since March 2017, and has investigated many hard-toaccess areas, including inside damaged turbine halls and reactor buildings. RISER is
integrated with Createc’s N-Visage 3D radiation imaging technology to provide maps ofradioactive contamination which help with planning decommissioning.

Self-Climbing Platform (SCP) at Sellafield – Nuvia
Sometimes it is also necessary to carry out decommissioning works at height without good access.
A current example is the SCP deployed to decommission one of the highest stacks at Sellafield. This
stack is 60m high and projects from the top of a 60m tall building. Access around the stack is very
limited due to other high hazard buildings. The SCP climbs up the stack and then dismantles thestack from the top down. The 3-tier SCP was installed in 2017, climbed to the top of the stack andis currently in demolition operations. It has total of 52 descend cycles ~1m each with demolition
activities taking place at each descend cycle to remove the liner, concrete wall and rebar. The SCP
is currently at the 13th descend cycle with 20% of the stack already demolished.

Conclusion
The use of UAVs is beginning to become a standard technique for inspection of building at nuclear sites in the UK.
The use of drones was pioneered at Sellafield, where Sellafield recognised the potential to save money, reducerisk and actively engage with its suppliers to set up standards for operating UAVs at its site. Since then, other sites,
including operating sites, have followed a similar model.
Innovative solutions continue to be developed in the UK to tackle the issues around access, space and proximity
to radiological and conventional safety hazards to make decommissioning and dismantling of nuclear facilities
safer, quicker and more cost effective. UK industry is pleased to be working in Fukushima and transferring theexperience and knowledge gained from these projects.

86

R01

Polyfunctional Adsorbent LHT-9: “All-in-One” Platform for theDevelopment of Multi-Nuclide Removal Systems
Sergey Britvin1, Boris Burakov2, Valery Mararitsa1, Yury Demidov1, Yury Petrov2,
Bella Zubekhina2 and Victoria Gribova1
1

Socium Ltd., 2V.G. Khlopin Radium Institute

Abstract
LHT-9 is a multi-nuclide sorbent applicable for both HLW and LLW decontamination. Granulated LHT-9
is suitable for use in flow columns, submersible filters and even exhaust gas cleaning systems. Besides, thesorbent can be readily modified into derivatives, like selective iodine scavengers. Altogether, LHT-9 can be
regarded as a perspective platform for “all-in-one” implementation at the ALPS-type decontamination facilities.

1. Introduction
The distinctive feature of nuclear waste is multi-nuclide composition. Consequently, its decontamination
requires multistep sorption by a series of selective sorbents. The development of single “all-in-one” sorbent isstill a technological challenge. We herein present LHT-9, a promising candidate for multi-nuclide removal.

2. Layered Hydrazinium Titanate
LHT-9 is a layered titanate containing chemically bound hydrazine [1,2]. It is capable of simultaneous
sorption of more than 50 chemical elements, including 62 radinuclides known at the Fukushima site. It
possesses specific surface area up to 500 m2/g and available in granulated form. LHT-9 can be used fordecontamination of both HLW and LLW. The difficult-to-clean volatiles, like elemental iodine and methyl
iodide can be readily removed from the exhaust gases. Besides, the modified LHT-9 sorbents are effective
halogen-selective scavengers of both iodide (I−) and iodate (IO3−) from the liquid waste.

Figure 1. The flowchart illustrating implementation of LHT-9 for nuclear waste disposal
3. Conclusion
Based on the properties of LHT-9, this sorbent can be successfully implemented into the multi-nuclide
removal systems at the Fukushima site. It can be also used for decontamination of ordinary technological waste
produced during normal operation of atomic power stations.

References[1] Britvin, S.N. et al. (2011): J. Am. Chem. Soc. 133, 24, 9516-9525; DOI: 10.1021/ja202053q[2] PCT WO2011116788A1 / RU2564339C2[3] Zubekhina, B.Yu. et al. (2018): MRS Advances; DOI: 10.1557/adv.2018.208

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R02

Development of prediction model of the F1’s fuel debris accelerated
aging
Main approaches for estimation of fuel debris properties changing.
Oleg Bagryanov, Senior Project manager, TENEX*, Sergei Poglyad, Deputy
head of radiochemical technologies department PhD, RIAR*, Boris Burakov, Headof laboratory DSc KRI*

Abstract
While sampling of F1 fuel debris (FD) is still to be done in the future, projects for developing systems,
technologies and equipment of FD retrieval, storage and transportation are ongoing and require input data on the
FD properties. In the absence of real sampling such data may be obtained through research of overseas accidents
and modeling of samples simulating FD of F1. Rosatom is carrying out project to study ageing properties of F1
FD by researching Chernobyl samples and modeling simulated samples imitating FD of F1. This will allow toestimate the mechanisms of fuel debris self-destruction, to determine the factor most affecting changes inproperties, to develop the prediction model of F1’s fuel debris accelerated aging (based on the results ofresearches of Chernobyl’s fuel-containing materials).
1. Introduction
Rosatom develops a prediction model that will allow to estimate the changes of the main FD properties intime. Such prediction model will provide researchers (working on the developing of fuel debris retrieval,
packaging, storage and stansportation) with a possibilities to verify their solutions.
2. Main steps in development of prediction model
2-1. Investigation on the characteristics of fuel debris in overseas accident reactor and their influences on
the system to retrieve fuel debris
The main purpose of this stage is to summarize prior research of Chernobyl fuel containing materials(FCM), make the assumptions on major factors contributing to degradation due to ageing, and develop research
programme based on those investigations.
2-2. Acquisition of data on the changes by aging of fuel debris and its migration into water
Based on the initial Chernobyl research and programme of research model active samples will be preparedand tests with simulated fuel debris will be performed to investigate degradation processes due to ageing.
2-3. Comprehensive prediction of changes by aging
Based on the results of Stage 2, comprehensive prediction of changes byageing should be made to estimate changes after 10, 20, 30 and 50 years since
generation of fuel debris.
3. Conclusion
The tests with simulated active samples is the one of the possible scheme toapprove (by regulatory body) solutions developed for the FD management.

*Companies of Rosatom group
88

R03

Mobile robotic manipulator of anthropomorphic design

Sergei Floria1, Dmitry Kucherov1
1

RosRAO, FSUE

Abstract
RosRAO has developed and conducted an experimental operation of mobile robotic manipulator (MRM)of anthropomorphic design to conduct work within radiation fields.
The manipulator is built on hydraulic circuit with replaceable electromechanical grippers for precise
operations. The robot copies the movement of operator through a special copying suit and allows to conduct
works at a distance up to 100 meters.
1. Introduction
The MRM ensures manipulations with stuff under 15 kilos with the help of one hand, there are possible
operations with any common industrial tools. Such a configuration allows careful segregation of non-standard or
fragile radioactive wastes as well as applying common manual tools to dismantle equipment in high radiation
fields. Robot’s hands are manufactured with additive technologies.

Pic.1 MRM mounted on the industrial manipulator

Pic.3 SRW storage

Pic.2 Control suitPic.4 Screenshot: operator’s training program

89

SK01

A Tritium Measurement Technology of Underwater around
Decommissioning Site
Ki Joon Kang, Jun Woo Bae, Hee Reyoung Kim

School of Mechanical, Aerospace and Nuclear Engineering, Ulsan National Institute of Science andTechnology, Ulsan 44919, Republic of Korea
Abstract
Tritium is a pure-beta emitter. When tritium is absorbed into the body, it displaces protium of water molecule toform Organically Bound Tritium (OBT) and causes ionization damage of Deoxyribonucleic Acid (DNA). Whenthe nuclear power plants were dismantled, it is necessary to develop a device for measuring the tritium
concentration in underwater around the decommissioning site.
1. Introduction
In order to measure the radioactivity concentration of tritium in water, tritium used in this study is concentrated
tritium gas sample. Tritium calibrator is currently in development. When tritium calibrator is developed in the
future, to compare the experimental results with commercial tritium monitor calibrator and to verify the validityof the experimental results by the developed system, the radioactivity concentration was measured by commercial
tritium monitor calibrator. In order to confirm the validity of the measurement results by commercial tritium
monitor calibrator, the expected theoretical results of the radioactivity concentration got by measuring of thecalibrator volume, pressure, tritium gas decay factor, and tritium gas initial concentration should be compared withthe actual experimental results.
2. Method and Result2-1. The Theoretical Calculation
The theoretical calculation was carried out by the equation derived by measuring the volume and the pressure ofcalibrator [1]. Figure 1 shows activity concentration of tritium
according to pressure in metering volume. the range of pressure inthis study was varied from 11 psi to 20 psi. The theoretical
calculation results are represented as black closed square of figure
1. The slope of the theoretical calculations was estimated by 0.567.
2-2. The Experimental Measurement
The experimentally measured results are represented as red open
square on the figure 1. The slope of the experimentally measured
results was estimated by 0.565 using linear fitting.
3. Conclusion
Both the theoretical calculation and the experimentally measured
Figure 1. radioactivity concentration

results werelinearly increased. The slope of each results was highly similar

according to pressure.

with relative error of 0.35 %. These results can be a reference for the performance of device to be developed.
References[1] OVERHOFF TECHNOLOGY CORPORATION, CALIBRATION PROCEDURE TRITIUM GAS CALIBRATORMODEL 10017

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SK02

Detection Technique for Real-time Monitoring of Tritium in
Underwater
Jun Woo Bae, Ki Joon Kang and Hee Reyoung Kim

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50,
UNIST-gil, Ulju-gun, Eonyang-eup, Ulsan, 44919, Republic of Korea
Abstract
A detection technique for real-time monitoring of tritium in water was proposed. The method was based on
the electrolysis. Experimental set-up was carried out to measure tritiated water samples continuously using
electrolysis. The detection efficiency for hydrogen with tritium was estimated as 27%.

1. Introduction
The decommissioning of the Fukushima Daiichi Nuclear Power Plant is a pending issue not just for Japanbut also for the international community. During the D&D process radioactivity monitoring is essential toprevent secondary contamination and dispersion of contaminated material [2]. Monitoring of tritium in thelaboratory is time-consuming work. In this study, a detection technique for real-time monitoring of tritiatedwater was proposed based on the electrolysis.

2. Methods and Results
Figure 1 shows overall features of real-timetritium monitoring system. When the sample(tritiated water) is poured in the sample container,
the pump pushes the sample to the electrolysis cell.
The sample is electrolyzed and hydrogen gas withtritium is released. Overflowed water is trapped andonly hydrogen gas goes through the plastic
scintillator based detection chamber. The detection
efficiency of the tritiated hydrogen gas was

Figure 1. Real-time tritium monitoring systemestimated as 27%. The detection limit of the system was 17.8 kBq/L for 15,000 s of measurement time.

3. Conclusion
A detection technique for real-time tritium monitoring was proposed. The method was based on theelectrolysis and experimentally characterized. The technique was applicable for high-radioactive tritiated water
which can be released during decommissioning process.

References[1] Mori, Koji, et al. "Continuous improvement of fitness-for-duty management programs for workers engaging in stabilizing
and decommissioning work at the Fukushima Daiichi Nuclear Power Plant." Journal of occupational health, 60.2:196-201(2018).
[2] Kang, Houjun, Dong Zhang, and Zhengkun Shi. "Measurement of tissue free water tritium in plant." Atomic Energy
Science and Technology, 43.3:221-223 (2009).

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SK03

The Efficiency Calculation of Plastic Scintillator for in-situ Beta
Measurement System in Groundwater
Woo Nyun Choi, Ukjae Lee, Jun Woo Bae and Hee Reyoung Kim

School of Mechanical Aerospace and Nuclear Engineering, UNIST, 44919 Ulsan, Republic of Korea
Abstract
Develop in-situ groundwater beta measurement system ofwhich detection part comes into direct contact with the matter. The
performance of plastic scintillator for in-situ measurement system
was analyzed using simulation and experimental methods.
Fig. 1 Concept of in-situ system

1. Introduction

Decommissioning site can be contaminated with various radionuclides, this site characterization is needed.
Conventional analysis method of beta-emitting radionuclides is usually costly and result in long delay between
sampling and obtaining results [1]. Therefore, in-situ monitoring technology is requested like Fig. 1.

2. Method and Result2-1. Plastic scintillator and beta source
Plastic scintillator is used as it is non-hygroscopic and can be fabricated into various shapes with large
sensitive area. 3H, 14C, 32P, 90Sr/90Y were pure beta emitting radionuclides used in the simulation and experiment.
2-2. MCNP simulation and experiment
The efficiency of the plate was determined instead by the energy deposition of beta particles using MCNP6F8 (e.p) tally. For experiment radioactive aqueous samples were prepared and counted for 600 seconds. The
dimension of vial was similar to simulation one. The
thickness of plastic scintillator was chosen as 1 mm.

Table 1. Experiment results compared to simulation

2-3. Result
The counts of 3H and 14C are similar to that of the

Sourceblank sample. But 32P and 90Sr/90Y showed good agreement
32

P

between simulation and experiment. The relative difference
values of efficiency were 1.8 % for 32P and 0.4 % for90

Sr/90Y as shown in Table 1.

90

Sr/90Y

Experiment

Simulation

Relative

efficiency

efficiency

difference

5.55 ±

5.650 ±

0.08 %

0.002 %

4.60 ±

4.620 ±

0.01 %

0.001 %

1.8 %

0.4 %

3. Conclusion
The efficiency of plastic scintillator was calculated based on MCNP simulation and experiment. 1 mm ofplastic scintillator could be used in in-situ system. Thin plastic layer (0.5 mm) would degrade accuracy due tonoise and background.

References[1] In-situ radiation detection demonstration, Stephen W. Duce, Amir H. Mohagheighi, Mark L. Miller, Robert R. Reese andDavid R. Miller. WM’00 Conference, February 27 – March 2, 2000, Tucson, AZ.

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SK04

Radiation Dose Rate Distribution Monitoring System with Portable
Gamma Detector for Decommissioning Site
Rapid and broad scale of site monitoring
UkJae Lee1, Jun Woo Bae1, Hee Reyoung Kim1

1)

School of Mechanical, Aerospace and Nuclear Engineering, Ulsan National Institute of Science andTechnology, Ulsan 44919, Republic of Korea

Abstract
In order to secure the radiological safety of the decommissioning site, a system for monitoring the radiationdose rate which can apply to a wide area was established.

1. Introduction
It is essential to secure the radiological safety of radiation workers and surrounding residents within thedecommissioning site. Based on this need, the system which can be applicable to a wide range of areas, andrapid site monitoring technology is suggested.

2. Method and Result2-1. System components
The system is based on portable 3 in by 3 in NaI(Tl) scintillator detector which has energy range of 80 keV~ 3 MeV and 7.5 % of detection resolution. The contour-mapping program, implemented with MATLAB
software, enables the 2D or 3D display of radiation level distributions with one click of a mouse immediately
after the corresponding area is scanned by the portable gamma radiation detector.
2-2. Monitoring methods
Scanning around site was performed in a vehicle with the installed system. The information on latitude,
longitude, and radiation dose rate were obtained through actual measurements and used as data sets for theradiation contour mapping. The surrounding area of the monitoring point was scanned by a vehicle and its speed
was approximately 30 km/h. The distance between the ground and the detector was 1 m.
2-3. Results
Figure 1 shows contour maps of radiation distributions. It
verifies that the radiation distributions were shown in the 2D and
3D contour maps with geographical information distributions.

3. ConclusionRealmeasurement-based

quick

radiation

distribution

monitoring system was established. The system could be directly
applied to in-situ finding of hot spots in the vast area subject todecommissioning of nuclear power plants.

References[1] Kim, H. R., et al. (2008). "The Establishment of an In-Situ Real Time
Radiation Contour Mapping Technique." Journal of Nuclear Science andTechnology 45(sup5): 662-665.

Figure 1 The Result of Monitoring System
for Radiation Dose Rate Distribution

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SK05

Criteria Analysis of Autonomous Aerial and Marine Radiation
Monitoring System for Safe NPP Decommissioning
Chanki Lee, Hee Reyoung Kim

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology
Abstract
A new aerial and marine radioactivity monitoring system is developed based on application of autonomousvehicles. Specifically, by assuming several radiological accident scenarios during nuclear decommissioning,
requirements for MDA (Minimum Detectable Activity) and deployment and motion pattern are determined.

1. Introduction
Decommissioning of nuclear power plants requires a lot of workforce and cost. Autonomous vehicle can besuitable candidates for radioactivity monitoring system, reducing additional workforce and cost. There are types
of autonomous vehicles for both aerial and marine environmental survey, such as fixed wing UAV (Unmanned
Aerial Vehicle), rotary wing UAV, autonomous underwater glider, and autonomous float. In order to construct
efficient system by using above vehicles, requirements including MDA quantities, deployment/motion patternshould be determined appropriately.

2. Requirements for autonomous radioactivity monitoring system
2-1. Accident scenarios during nuclear decommissioning
IAEA suggested 6 example list of incidents and accidents during decommissioning of nuclear facilities(e.g., Spill of decontamination fluid, failure of ventilation system) [1]. Such scenarios can affect not only
workers, but also neighboring residents. Thus, long-range aerial and marine monitoring activities should be
conducted to protect nearby people.
2-2. MDA requirements
Aerial MDA quantities were set up based on AMS (Aerial Measuring System) of U.S. DOE [2], which uses
helicopter and airplane for airborne radiation measurement. Marine MDA quantities were determined by 10
years radioactivity survey data of KINS (Korea Institute of Nuclear Safety) over the Far Eastern Seas [3].
2-3. Deployment and motion pattern for autonomous tracking
Autonomous vehicles would be deployed in and around nuclear power plants to monitor aerial and marine
radioactivity year-round. When radiological accident occurs during nuclear decommissioning, motion forfinding higher radioactivity level, serpentine motion pattern, and concentric motion pattern will be beneficial toassess radiological dispersion efficiently.

References[1] IAEA, 2013, Safety Assessment for Decommissioning, Safety Report Series No. 77, IAEA.
[2] U.S. DOE, 2000, Overview of FRMAC Operations, U.S. DOE.
[3] Yoon et al., 2015, Marine Environmental Radioactivity Survey, KINS.

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US01

AVANTech’s Advanced Polymer Solidification for Radioactive Waste Stabilization:
Case Studies
Tracy Barker, Gary Benda, Jim Braun, Raja Beereddy and Dennis Brunsell
AVANTech, Inc., 2050 American Italian Way, Columbia, SC 29209

This poster describes recent developments in polymer technology, as well as summarizes several applications
and the history of process improvements that simplify the process and improve the efficiency for stabilizing
low and intermediate level radioactive and mixed radioactive/hazardous wastes.
Polymer technology has been used successfully since the early 1990s to solidify and stabilize radioactive andmixed hazardous/radioactive waste at United States (US) Department of Energy (DOE), US Department of
Defense (DOD), United Kingdom government sites, and at commercial nuclear power plants. It is aneffective, efficient, and economical means of preparing waste for storage and disposal.
In 1992, Dow Chemical effected a transfer of the polymer solidification technology to Diversified
Technologies Services, Inc. (DTS) of Knoxville, TN, (a firm acquired in 2014 by AVANTech, Inc. ofColumbia, SC). DTS undertook to expand the Dow polymer solidification initiative to applications beyond
liquids and resins from chemical cleaning solutions to in-situ solidification of granular waste such as ion
exchange resin and activated carbon.
Based on extensive waste form testing, the Advanced Polymer Solidification (APS™) was shown to be a
viable method of generating stable US NRC Class B & C resins and filters for on-site storage pending the
opening of a US NRC Class B & C disposal facility. A key advantage of APS™ is that it renders a productthat is compliant with NRC 10 CFR 61.56(b) Branch Technical Position (BTP).
The AVANTech APS™ processes use in-container mixing, high-speed mixing, and in-container In-Situ
solidification to address a wide variety of radioactive and mixed waste materials, including liquids, sludges,
bead resin, granular media, and powders. The resulting polymer-solidified product forms a rock-hardmonolith and meets or exceeds the US NRC 10 CFR 61.56(b) BTP requirements for a Class B & C waste
form whose integrity is assured for long-term storage, transportation, and land burial environments.
Recently, polymer solidification has been combined with other waste pretreatment technologies, such as
evaporator concentrate drying, to further enhance volume reduction and packaging efficiency.
This poster reviews applications of the technology for use at Fukushima and identifies waste streams wherepolymers have advantages over cement, vitrification, and other stabilization processes.

APS™ Mobile Metered Injection System Primary Resin Solidification at the US Diablo
Canyon Nuclear Power Plant

95

US02

Monitoring Internal Conditions of Radioactive Waste Storage Containers
Gregory Allan, Pentek

The cleanup process at the Fukushima Daiichi site presents many technical challenges. One of the more
important tasks involves the packaging, short-term storage, transport, and long-term storage ofradioactive waste.
The waste will consist of traditional and non-traditional forms. An example of a traditional form-factor
are the spent fuel assemblies loaded in fuel storage containers. Non-traditional form factors include thecontaminated steel, concrete and other building materials being removed by robotic systems and otherdecontamination and demolition (D&D) activities. In fact, Pentek supplied an WIFI controlled robotic
system, known as the Moose equipment, to TEPCO and Kajima to remove rubble from the refuelingfloor in U3 at Fukushima. Custom, fork-truck compatible waste containers were used for the storage oflarge rubber (< 10 cm in diameter) and standard 200-liter drums were used for small debris and dust.
In 2016-2017, Pentek, and its sister company CM Technologies, completed a research program for the
US Department of Energy to develop a technology that could pass data and power through the thicksteel containers used for spent fuel storage in the USA without any wires or penetration holes in thecontainer.
This presentation explains this technology and how it could be applied in radioactive waste packaging at
Fukushima.

96

US03

UK National Waste Program: A Successful Transformation inRadioactive Waste Management
Tim Carraway, Dennis Thompson, and Max Ehrhardt
AECOM

Abstract
This Poster illustrates AECOM’s experience in developing and executing a National Waste Management
Program for a country with very limited radioactive waste disposal capacity. Similar to Japan, the United
Kingdom’s radioactive waste management approach is to minimize the volume of waste that requires disposal attheir only radioactive disposal sites – Low Level Waste Repository (LLWR) and Dounreay’s Low Level Waste
Vaults.
1. Implementing a New Waste Management Policy
To account for large-scale decommissioning and environmental remediation projects, the UK Government
announced the LLW Policy in 2007. Since 2011, AECOM has developed and implemented the UK National
Waste Program that accomplishes the goals of the Policy. Before AECOM developed the UK National Waste
Program, it was projected that the LLWR site’s disposal capacity would be fully exhausted by the year 2023. As
a result of AECOM’s successful implementation and execution of the UK’s National Waste Management
Program, the LLWR site’s useful capacity and life has been extended by one hundred years.
2. AECOM Approach to UK Low-Level Waste Management
The success of the UK’s LLW Policy has contributed to the extensive focus on detailed characterization,
application of the appropriate waste hierarchy, and application of processing technologies to minimize or avoid
waste disposal. To preserve the limited disposal capacity in the UK, a waste-disposal-avoidance hierarchy isdiligently applied to Decommissioning Projects as well as other waste generators including Defense Projects,
Medical Facilities, Educational Facilities and other industrial generators of radioactive waste. A waste-informed
decommissioning process is applied such that limited disposal waste volumes are incorporated into thedecommissioning strategy to provide the best overall solution for the country.
The National Waste Program includes individual services
contracts with all UK waste generators. AECOM’s waste
management approach is integrated into the entire Supply
Chain of the waste generators which allows cross-utilizationof services and significantly minimizes the cost for thefollowing Supply Chain elements:
 Safety Case Analysis
 Waste Characterization
 Transportation
 Packaging
 Metals Processing Combustible Processing
Figure 1. Reduced LLW Disposal under the NWP Very Low Level and Low Level Waste Disposal
 Waste Tracking Systems
Avoidance of Radioactive Waste generation is the best option as it has extended the life of the LLWR siteand has greatly reduced the cost of disposal, transportation, and treatment requirements in the UK. Avoidancehas best been achieved by AECOM working directly with the decommissioning sites to ensure that the correct
Decommissioning and Dismantling processes are applied to minimize the generation of as much radioactive
waste as possible. Applying decommissioning and dismantling methods that avoid co-mingling of radioactive
waste and non-radioactive waste is a primary approach to avoid radioactive waste generation. Other methods
used by AECOM to minimize disposal at LLWR include Metal Size Reduction, Decontamination via Shot
Blasting, Metal Melting, Incineration of Combustible Waste, and Shredding and Compaction.
3. Conclusion
Even though waste volumes in the UK have increased as shown in Figure 1, the risks of depleting thedisposal capacity at LLWR and the newly opened disposal vaults at the Dounreay site in Scotland are being
successfully mitigated. By utilizing the Waste Management Program, waste generators are avoiding more than
90% of their waste from requiring disposal at LLWR. Waste treatment and avoidance methods are proving to beeither cost neutral or cheaper than disposal, resulting in sustainable savings for the UK’s decommissioning
programs. The existing LLWR site will be able to service the needs of UK waste generators to the year 2130 and
potentially beyond.

97

US04

EnergySolutions Decommissioning Experience
Colin Austin1 and Makoto Kikuchi2
1

EnergySolutions, 2 MKC Consulting

Abstract
EnergySolutions (ES) is an international leader in commercial nuclear power plant (NPP)
decommissioning. It has developed this experience over a period of 30 years performing NPP decommissioning
and related nuclear waste management services in the USA and Europe. This presentation will provide somespecific examples of its decommissioning experience in the USA and identify benefits that this experience
brings to NPP decommissioning projects.
1. Decommissioning Life Cycle Experience
ES was formed 12 years ago and is now made up ofover 10 different companies that have been integrated toprovide the complete value chain of services necessary to
successfully perform NPP decommissioning projects.
The scope of work in successful NPP decommissioningthat ES has been involved in includes:⚫

Regulator compliant estimates for funding,⚫

Detailed decommissioning planning,⚫

Spent fuel management and storage,⚫

Decontamination and dismantling,⚫

Equipment, facility and site characterization,⚫

Waste management (treatment, disposal packaging, transportation and logistics), and,⚫

NPP site and waste management facility operating license holder.

This experience allows ES to understand and optimize the NPP decommissioning strategy to ensure projects are
performed safely, in full compliance with all regulatory and permitting requirements and at the lowest cost.
2. Extensive Experience in USA and Europe

3. Japanese Experience (Technology Transfer, Services & Equipment)
JAPCO

D&D CollaborationTEPCO

ALPS Water Treatment

EnergySolutions’ experienceis transferable to Japan.

1F Project Management Support
JNFL(IHI)

Vitrification Technology Support

98

US05

EnergySolutions Waste Management Experience
Colin Austin1 and Makoto Kikuchi2
1

EnergySolutions, 2 MKC Consulting

Abstract
The key to success in decommissioning is the effective management of the radioactive waste that it generates.
An effective waste management strategy is focused on the optimum approach to complying with the final
approach for dispositioning or disposing of the waste. ES developed a broad range of waste processing(decontamination and volume minimization) techniques to optimize the waste strategy to meet different national
and international disposal criteria. Examples of these techniques include:
Mechanical decontamination (grit / abrasive blasting/ high pressure spray), liquid / water treatment, chemical
decontamination, electrorefining / electropolishing, incineration, metal melting / recycling, steam reforming,
vitrification, shredding, compaction, equipment salvage and reuse, repackaging for volume reduction,
reclassification etc.
This experience allows ES to develop optimized waste management strategies for operating NPPs and NPP
D&D projects that can be delivered safely, in full compliance with all regulatory and permitting requirements
and at the lowest cost.

1. Acquisition and Integration of Companies to Realize full Support Waste Management and D&D

2. Licensed disposal Facilities and Processing Facilities in US

3. Japanese Experience
J-PARC

2500-ton shield block using recycled metal

TEPCO

Bunker processing equipment

EnergySolutions’ experience

NDF

Waste management guidelines

is transferable to Japan.

NFT

Used cask recycle study

99

List of Organizations
Organization

ID No.

ADVAN ENG. Co., Ltd.

H01, H02

AECOM

US03

AIFEN

F08

Akita University

K04

AMRC (Nuclear Advanced Manufacturing Research

GB01, GB02, GB03, GB04

Centre)ANADEC (Orano Atox D&D Solutions Co., Ltd.)

B07

ANDRA

F01

AREVA NC

F01

AtkinsGB01, GB02, GB03, GB04

ATOX Co., Ltd.

B06

AVANTech, Inc.

US01

British Embassy Tokyo

GB01, GB02, GB03, GB04

Cavendish Nuclear Ltd.

GB01, GB02, GB03, GB04

CEA (French Alternative Energies and Atomic Energy

C07, J05, F01, F02, F03,

Commission)

F04, F05, F06, F07, F08

CHUBU Electric Power Co., Inc.

D05

CORNES Technology Limited

B09

Createc

GB01, GB02, GB03, GB04

CRIEPI (Central Research Institute of Electric Power

A03, C03, C07, D07

Industry)
CTI

F06

DAIHATSU Co., Ltd.

H01

Daihatsu Motor Co., Ltd.

H03

DBD (DBD International)

GB01, GB02, GB03, GB04

EBARA Corporation

D01, D02, D03

ECM Technologies

F01

ELyTMax

C07

EnergySolutions

A06, US04, US05

FEVDI

F04

Forschungszentrum Jülich GmbH

H03

Fuji Electric Co., Ltd.

K06

Fujikura Ltd.

B10

Fukushima Consortium of Robotics Research for

B01, B02

Decommissioning and Disaster Response100Fukushima Midori Anzen Inc.

B13

Fukushima University

J08

GleedsGB01, GB02, GB03, GB04

Goodwin PLC

GB01, GB02, GB03, GB04

Gunma University

C07

Hirosaki University

K04

Hitachi, Ltd.

B08, C01

Hitachi-GE Nuclear Energy, Ltd.

C01

Hosei University

A02

IAE (Institute of Applied Energy)

A03

IHI Corporation

K01, K02, K03

Imagineeing Inc.

J06

INS (International Nuclear Services)

GB01, GB02, GB03, GB04

INSA (National Institute of Applied Sciences)

C07

IPL (Innovative Physics Limited)

GB01, GB02, GB03, GB04

IRID (International Research Institute for Nuclear

B03, B04, B06, C02, J04,

Decommissioning )IRSN(InstitutdeRadioprotection

J05, J07
etdeSûreté

F02

Nucléaire)
JAEA (Japan Atomic Energy Agency)

A03, C02, C06, H01, H03,
J01, J02, J03, J04, J05, J06,
J07, J08, J09, J10, J11, J12,
J13, J14, J15, J16

James Fisher Nuclear Ltd.

GB01, GB02, GB03, GB04

Japan Radioisotope Association

A03

JAPC (The Japan Atomic Power Company)

A06, K05

JFEE (JFE Engineering Corporation)

K05

JGC Corporation

K07

JNFL (Japan Nuclear Fuel Limited)

K04

Kajima Corporation

E01, E02

KHNP (Korea Hydro & Nuclear Power Co., Ltd.)K03

KOBELCO STUDSVIK Co., Ltd.

K10KRI (V.G. Khlopin Radium Institute)

R01, R02

Kurion Japan K.K

B05

Kwansei Gakuin University

H01, H03

Kyoto University

C03, J06, J08

Mirion Technologies (Canberra) K. K.

C04, C05

101

Mitsubishi Research Institute, Inc.

A04

MKC Consulting

US04, US05

MNF (Mitsubishi Nuclear Fuel Co., Ltd.)

D04

Mott MacDonald

GB01, GB02, GB03, GB04

Nagaoka University of Technology

H01, H02, H03

Nagoya University

C06

National Museum of Nature and Science

J08

NDA (Nuclear Decommissioning Authority)

GB01, GB02, GB03, GB04

NDF

(Nuclear

DamageCompensation

andA01, D07

Decommissioning Facilitation Corporation)
NFD (Nippon Nuclear Fuel Development Co., Ltd.)C02

NIA (Nuclear Industry Association)

GB01, GB02, GB03, GB04

NINS (National Institutes of Natural Sciences)

J06

NIS Ltd.

GB01, GB02, GB03, GB04

NNL (National Nuclear Laboratory)

GB01, GB02, GB03, GB04

Nuclear Valley

F08

Nucléopolis

F08

NUKEM Technologies Engineering Services GmbH

K08, K09

Nuvia

GB01, GB02, GB03, GB04

Obayashi Corporation

D06

OECD/NEA (Organisation for Economic Co-operation

ON01, ON02, ON03, ON04

and Development/Nuclear Energy Agency)
ONET Technologies

F02

ORANO

F04, F06

Osaka Prefecture University

J09

OSMOS Association

C10

OSMOS Group SA.

C10

PCubedGB01, GB02, GB03, GB04

PentekUS02

QST (National Institutes for Quantum and Radiological

J06, J09

Science and Technology)
RIAR (Research Institute of Atomic Reactors)

R02

RLB (Rider Levett Bucknall)

GB01, GB02, GB03, GB04

Rolls Royce

GB01, GB02, GB03, GB04

RosRAO

R03

Sellafield Ltd.

GB01, GB02, GB03, GB04

Shibaura Institute of Technology

A05

102

Socium Ltd.

R01

TAISEI CORPORATIONE03

TAIYO KOGYO CORPORATION

E03

TENEX

R02

TEPCO (Tokyo Electric Power Company Holdings, Inc.)

E03, E04, E05

The Graduate School for the Creation of New Photonics

D05

Industries
The University of Shiga Prefecture

K04

The University of Tokyo

C06, C08, C09, J09

Tigers Polymer Corporation

B12

Tohoku Institute of Technology

C06

Tohoku University

C06, C07, J09

Tokai University

A02, A05

Tokyo City University

A05

Tokyo Institute of Technology

A05, K04, J09

Tokyo Medical and Dental University

A05

Toshiba Corporation

A02

Toshiba Energy Systems & Solutions Corporation

A03

Toyokoh Co., Ltd.

D05

Toyo Rubber Chip Co, Ltd.

B11

TOYO UNION Company Limited

E03

UKAEA (UK Atomic Energy Authority)

GB01, GB02, GB03, GB04

UNIST (Ulsan National Institute of Science andSK01, SK02, SK03, SK04,

Technology)

SK05

Utsunomiya University

H01

Veolia Nuclear Solutions, Inc.

B05, GB01, GB02, GB03,
GB04

Wood

D06, K06

Wood PLC

GB01, GB02, GB03, GB04

Nuclear Damage Compensation andDecommissioning Facilitation Corporation (NDF)
ndf-poster@ndf.go.jp
http://www.dd.ndf.go.jp/eindex.html
July 2018103