Nuclear Damage Compensation andDecommissioning Facilitation Corporation (NDF)
ndf-poster@ndf.go.jp
http://www.dd.ndf.go.jp/eindex.html
July 2019

Contents
Session A: Evaluation, Activities and Experiences related to Decommissioning ･････････････1
A01

A Study of Risk Analysis and Evaluation during a Debris Removal Process
Akio Yamamoto1,2, Takashi Takata1,3, Satoshi Takeda1,4, Takashi Ino5 andKazuki Hida5
1: Atomic Energy Society of Japan (AESJ), 2: Nagoya University,
3: Japan Atomic Energy Agency (JAEA), 4: Osaka University,
5: Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF)

A02

Research and Development of the Project of Decommissioning and Contaminated
Water Management and Connection to Preliminary Engineering
Masaki Kawai and Fumiya Sato
Mitsubishi Research Institute, Inc.

A03

Value Chain in the Field of Decommissioning and Dismantling of Nuclear
Facilities
Mamoru Numata, Sadaaki Abeta, Takeshi Nitta, Ken Kojima and
Erwan Hinault
Kurion Japan K.K.,

A04

Remote Access Technologies for Decommissioning of Nuclear Facilities
Mamoru Numata1, Sadaaki Abeta1, Takeshi Nitta1, Ken Kojima1,
Erwan Hinault1, Mark Sharpe2, David Loughborough2, Scott Martin3,
Matt Cole3 and Marc Rood3
1: Kurion Japan K.K., 2: Veolia Nuclear Solutions (UK) Ltd.,
3: Veolia Nuclear Solutions Inc.

A05

GeoMelt® Application to 1F Waste
Sadaaki Abeta1, Erwan Hinault1, Jeremy Whitcomb2, Brett Campbell3 andKevin Funicane3
1: Kurion Japan K.K., 2: Veolia Nuclear Solutions Inc.,
3: VNS Federal Services Inc.,

A06

MITSUFUJI Wearable HRV Monitoring System with Garment Type Sensor
Predicts Workers Health Problems in Real Time
Hiromichi Osoegawa
MITSUFUJI Corporation

Session B: Remote Technology and Radiation Tolerant Equipment ･･････････････････････7
B01

The Fukushima Consortium of Robotics Research for Decommissioning andDisaster Response
Yoshiro OwadanoiFukushima Consortium of Robotics Research for Decommissioning andDisaster Response
B02

R&D of an Automatic Suppression Robot that can Pinpoint the Source of an
Incipient Fire and Immediately Suppress the Fire for Nuclear Facilities
Yukio Akatsu, Shigeo Kanke and Kaoru Akatsu
Tokyo Bosai Setsubi Co., Ltd.

B03

Development of underwater robot Named “RADHOTAR” RADiation HOsTile
semiAutonomus Robot
Shigekazu Suzuki1, Shinji Kawatsuma1, Eiji Aoki1, Masanori Takahashi1,
Koki Watanabe2, Shuzo Nakano3, Kosuke Haga3 and Hideyuki Hnawa4
1: National Institute of Technology (NIT), Fukushima College,
2: Takawa Seimitsu Co. Ltd., 3: East Japan Accounting Center Co. Ltd.,
4: Ascend Inc.

B04

Disaster Response Robots ‘Giraffe’ Developed by the Members of Fukushima
Consortium of Robotics Research for Decommissioning and Disaster Response
Noritaka Baba
Aizuk, Inc.

B05

Organization Profile of IRID
Kazufumi Aoki
International Research Institute for Nuclear Decommissioning (IRID),

B06

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

B07

Radiation soft PVC materials-resistant
Huminori Egusa and Tomoyuki Yamanaka
Tigers Polymer Corporation

B08

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

B09

Monitoring technology for fuel debris removal process
decommissioning
Akira Kobayashi and Hideyuki Suzuki
Hamamatsu Photonics K.K.

B10

Introduction of Mirion Technologies’ Radiation Tolerant Camera
The World’s Highest Radiation Tolerant Performance Camera
Yasuo Arai, Mikio Katsura and Toshiki Matsui
CORNES Technologies Limitedii

in Fukushima

Session C: Analysis, Detection and Evaluation for Radioactive Materials ･･････････････17
C01

Development of Dosimeter for Severe Radiation Environment near Reactor
Pressure Vessel
Tamotsu Okamoto1, Yasuhito Gotoh2, Masafumi Akiyoshi3,
Mitsuru Imaizumi4, Tomohiro Kobayashi5 and Yasuki Okuno6
1: National Institute of Technology (NIT), Kisarazu College,
2: Kyoto University, 3: Osaka Prefecture University,
4: Japan Aerospace Exploration Agency (JAXA), 5: RIKEN,
6: Japan Atomic Energy Agency (JAEA)

C02

Progress of PYRAMID project - Piping system, risk management based on wall
thinning monitoring and prediction –Toshiyuki Takagi1,2, Philippe Guy3, Yutaka Watanabe2,4, Hiroshi Abe4,
Shinji Ebara4, Tetsuya Uchimoto1,2, Takayuki Aoki4, Mitsuo Hashimoto1,
Ryoichi Urayama1, Hongjun Sun1, Thomas Monnier3, Jérôme Antoni3,
Bernard Normand3, Nicolas Mary2, Ryo Morita5, Shun Watanabe5,
Atsushi Iwasaki6, Hiroyuki Nakamoto7, Christophe Reboud8, Pierre Calmon8,
Edouard Demaldent8, Vahan Baronian8, Xavier Artusi8, Sylvain Chatillon8
and Alain Lhemery8
1: Institute of Fluid Science, Tohoku University,
2: ELyTMaX UMI 3757, CNRS – Université de Lyon – Tohoku University,
3: Université de Lyon, 4: Graduate School of Engineering, Tohoku University,
5: Central Research Institute of Electric Power Industry (CRIEPI),
6: Gunma University, 7: Kobe University,
8: French Alternative Energies and Atomic Energy Commission (CEA)-LIST

C03

Development of directional gamma-ray detectors for Fukushima decommissioning
Mitsuhiro Nogami1, Keitaro Hitomi1, Tatsuo Torii2, Yuki Sato2,
Yoshihiko Tanimura2, Kuniaki Kawabata2, Yoshihiro Furuta2,
Hiroshi Usami2, Kenichi Watanabe3, Toshiyuki Onodera4, Keizo Ishii1,
Kenji Shimazoe5 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

C04

Can cesium be adsorbed through the leaf and bark of trees?
– Estimation of the dominant absorption route of cesium in treesTomoko Ohta1,2, Junji Torimoto3 and Yasunori Mahara4
1: Central Research Institute of Electric Power Industry (CRIEPI),
2: The University of Tokyo, 3: Japan Agency for Marine-Earth Science andTechnology (JAMSTEC), 4: Kyoto University

C05

Radiochemical analysis of actual samples collected at the Fukushima Daiichi NPS
iii

site
Masashi Ichikawa1, Masaaki Kaneko1, Yoshikazu Koma2,3, Hiroshi Hinai2,3,
Ayako Nitta2,3 and Atsuhiro Shibata2,3
1: Nippon Nuclear Fuel Development Co., Ltd. (NFD),
2: Japan Atomic Energy Agency (JAEA),
3: International Research Institute for Nuclear Decommissioning (IRID)
C06

Estimation of Aging Properties of F1 Fuel Debris
Oleg Bagryanov1 and Sergei Poglyad2
1: TENEX JSC, 2: SSC RIAR JSC

Session D: Decontamination Technology and Treatment for Contaminated Water ･･････23
D01

Cesium decontamination of soil by combined techniques of elution, adsorption andelectromigration
Haruaki Matsuura1, Uladzimir I. Kislyi2, Vladimir M. Dudarchik2 andYasuhiro Kawashima2, Paul G. Molchanov2, Paul M. Bulai2 andVictor V. Boksha2
1: Tokyo City University, 2: Nuclear water soil solutions

D02

Adsorption Properties of Actinoids (U, Am, Np) for Various Inorganic Adsorbents
Hitoshi Mimura1, Minoru Matsukura1, Tomoya Kitagawa1, Fumio Kurosaki1,
Natsuki Fujita2, Hitoshi Kanda2, Akira Kirishima2, Daisuke Akiyama2 andNobuaki Sato2,
1: UNION SHOWA K.K., 2: Tohoku University

D03

Stable Solidification of Radioactive Wastes by Zeolites and Allophane
Hitoshi Mimura1, Minoru Matsukura1, Tomoya Kitagawa1, Fumio Kurosaki1,
Yuki Ikarashi2, Xiang-Biao Yin2, Chun-Nan Hsu2 and Yan Wu3
1: UNION SHOWA K.K., 2: Tohoku University,
3: Shanghai Jiao tong University

Session E: Progress of Decommissioning at the Fukushima Daiichi NPS ･･･････････････26
E01

Major Initiatives for Water Management at the Fukushima Daiichi NPS
Tomoya Kaneda
Tokyo Electric Power Company Holdings Inc. (TEPCO)

E02

Development of Laser Decontamination Method for Dismantling Flange Tank
Naoki Maeshiro1, Tomohiro Ogawa1, Nobushige Sasaki1, Ryohei Takeuchi2,
Fumio Ito2, Haruo Nagamine2 and Hiroshi Nakamura3
1: Tokyo Electric Power Company Holdings, Inc. (TEPCO),
2: Taisei Corporation, 3: Toyo Union Co., Ltd.

E03

Uranium-Containing Particles Found in the Fukushima Daiichi PrimaryContainment Vessel Interioriv

Masato Mizokami1,4, Kenichi Ito1,4, Akihiro Suzuki2, Ikken Sato3,4,
Masaki Kurata3,4, Mutsumi Hirai1,4 and Shinya Mizokami1,4
1: Tokyo Electric Power Company Holdings, Inc. (TEPCO), 2: Nippon Nuclear
Fuel Development Co., Ltd. (NFD), 3: Japan Atomic Energy Agency (JAEA),
4: International Research Institute for Nuclear Decommissioning (IRID)
E04

Units 1-2 Exhaust Stack: Local Company's Big Challenge -- Yes, it's possible!
Yukihide Sato, Isamu Okai and Mariko Chuman
ABLE Co. Ltd.

E05

Preparation Work for Unit-2 Fuel Removal at Fukushima Daiichi Nuclear Station
Takashi Inoue, Motohiro Taniyama, Jyunro Nakagoshi, Yasushi Kageyama,
Miho Miyazaki, Ippei Matsuo and Kihei Ogawa
Kajima Corporation

Session H: Hydrogen Treatment ････････････････････････････････････････････････････････31
H01

Variations of Hydrogen and Vapor Concentrations in a Simply SimulatedRadioactive Waste Long-Term Storage Container with a Newly Developed PAR
Kazuyuki Takase1, Genya Kawasaki1, Kohei Ueta1, Ky Tri Vu2 andIsamu Kudo2
1: Nagaoka University of Technology, 2: ADVAN ENG. Co., Ltd.

H02

Research on hydrogen safety technology utilizing the automotive catalyst
T. Aida1, T. Yamoto1, T. Kita1, H. Ono1, K. Takenaka1, D. Matsumura2,
M. Taniguchi 3, E-A. Reinecke4 and H. Tanaka1
1: Kwansei Gakuin University, 2: Japan Atomic Energy Agency (JAEA),
3: Daihatsu Motor Co., Ltd., 4: Forschungszentrum Jülich

H03

Research on hydrogen recombination catalyst under humidity environment
T. Yamamoto1, T. Aida1, K. Fujita1 T. Kita1, H. Ono1, D. Matsumura2,
M. Taniguchi3 and H. Tanaka1
1: Kwansei Gakuin University, 2: Japan Atomic Energy Agency (JAEA),
3: Daihatsu Motor Co, Ltd.

Session K: Radioactive Waste Management ････････････････････････････････････････････34
K01

Waste Management Symposia: Annual Conference in Phoenix, Arizona,
Exchanging Knowledge from Around the World
Kazuhiro Suzuki, Gary Benda and Akira Ohno
WM Symposia, Inc.

K02

Influence of Na and Ca Concentrations on Sr Adsorption Performance of Cs-Sr
Simultaneous Adsorbent
Akira Tsutsumiguchi1, Yuko Kani1, Tsuyoshi Ito1, Takashi Asano2,
Takatoshi Hijikata3, Tadafumi Koyama3 and Yasuhiro Suzuki4v

1: Hitachi, Ltd., 2: Hitachi-GE Nuclear Energy, Ltd.,
3: Central Research Institute of Electric Power Industry (CRIEPI),
4: JGC Corporation
K03

Study on Post-Accident Waste Management Scenarios for Fukushima Daiichi
Nuclear Power Station
Satoshi Yanagihara and Hiroshi Miyano
Atomic Energy Society of Japan (AESJ)

K04

SIAL® : Geopolymer solidification technology approved by Slovak / Czech Nuclear
Authority
Milena Prazska1, Marcela Blazsekova1, Hisashi Mikami2, Nobuyuki Sekine2
and Mika Mochimaru2
1: Wood, 2: Fuji Electric Co.,Ltd.

Session J: Research and Development of the JAEA ･････････････････････････････････････38
J01

Preparation of VR data for utilization to decommissioning of Fukushima Daiichi
Nuclear Power Station
Rintaro Ito, Takahiro Ohno, Yoshihiro Tsuchida and Takashi Okada
Japan Atomic Energy Agency (JAEA)

J02

Evaluation Method of Robot Mapping Performance for Decommissioning
Taichi Yamada and Kuniaki Kawabata
Japan Atomic Energy Agency (JAEA)

J03

Development of Corrosion Database in Specific Environment
Tomonori Sato1, Kuniki Hata1, Yoshiyuki Kaji1, Hiroyuki Inoue2,
Mitsumasa 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 and Technology(QST), 4: Tokyo Institute of Technology, 5: Tohoku University,
6: The University of Tokyo

J04

Development of a Robot Simulator System for Operator Proficiency Training
Kenta Suzuki, Shunji Mukaida, Fumiaki Abe and Kuniaki Kawabata
Japan Atomic Energy Agency (JAEA)

J05

Outline and Construction status of Okuma Analysis and Research Center,
Laboratory-1
Takahiro Furuse, Yuki Ota and Van-Khoai Do
Japan Atomic Energy Agency (JAEA)

J06

Development of analytical methods for 1F-samples ~Analysis of 93Zr by ICP-QQQMS and automated sample preparation instruments~
Van-Khoai Do1,2, Takahiro Furuse1,2 and Yuki Ota1vi

1: Japan Atomic Energy Agency (JAEA)
2: International Research Institute of Nuclear Decommissioning (IRID)
J07

Approach to analysis of hydrogen combustion inside of radioactive waste vessels
using CFD softwares based on an open source “OpenFOAM”
Thwe Thwe Aung, Atsuhiko Terada and Ryutaro Hino
Japan Atomic Energy Agency

J08

Detection of Alpha Particle Emitters Originating from Nuclear Fuels of the
Fukushima Daiichi Nuclear Power Station
Yuki Morishita
Japan Atomic Energy Agency (JAEA)

J09

On the hydrogen production of geopolymers under gamma irradiation
Vincent Cantarel, Makoto Arisaka and Isao Yamagishi
Japan Atomic Energy Agency

J10

Development of an Adaptive Control System for Laser Cutting of Fuel Debris andStructural Components for the Decommissioning of the Fukushima Daiichi
Nuclear Power Plant
Koichi Saruta, Naomitsu Kamei, Yuji Sato and Toshiharu Muramatsu
Japan Atomic Energy Agency (JAEA)

J11

Prediction of the dose rate of various fuel debris to be sampled from Fukushima
Daiichi Nuclear Power Station
Kenichi Terashima1,2 and Keisuke Okumura1,2
1: International Research Institute for Nuclear Decommissioning (IRID),
2: Japan Atomic Energy Agency (JAEA)

J12

Innovative laser remote analysis for surveillance of nuclear debris indecommissioning of Fukushima Daiichi Nuclear Power Station
Ikuo Wakaida1, Hironori Ohba1,2, Masabumi Miyabe1, Katsuaki Akaoka1,
Masaki Ohba1, Koji Tamura1,2, Morihisa Saeki1,2, Yuji Ikeda3 andTakunori Taira4
1: Japan Atomic Energy Agency (JAEA),
2: National Institutes for Quantum and Radiological Science and Technology,(QST), 3: Imagineering Inc., 4: National Institutes of Natural Sciences (NINS)

J13

Research and Development on Solidification Technology for Fukushima Daiichi
NPS Accident Waste
Jun Kato1,2, Taniguchi Takumi1,2, Takeshi Osugi1,2, Osamu Nakazawa1,2,
Tomoyuki Sone1,2 and Ryoichiro Kuroki1,2
1: Japan Atomic Energy Agency (JAEA),
2: International Research Institute for Nuclear Decommissioning (IRID)

J14

Sensitivity analysis of core slumping and alternative water injection in
Fukushima Daiichi Nuclear Power Plant Unit 3
vii

Xin Li1, Ikken Sato1 and Akifumi Yamaji2
1: Japan Atomic Energy Agency (JAEA), 2: Waseda University
Session F: France ･･･････････････････････････････････････････････････････････････････････52
F01

Fuel

Debriscuttingandaerosolcollection

forFukushima

Daiichi

Decommissioning
Damien Roulet1, Rémi Delalez1, Christophe Chagnot2, Ioana Doyen2,
Christophe Journeau2, Christophe Suteau2, Emmanuel Porcheron3 andThomas Gelain3
1: ONET Technologies, 2: French Alternative Energies and Atomic Energy
Commission (CEA), 3: Institut de Radioprotection et de Sûreté Nucléaire
(IRSN)
F02

Evaluation of the safety of geopolymer with respect to the hydrogen production
JAEA – CEA Collaboration
Yoshikazu Koma1, Vincent Cantarel1, David Lambertin2, Veronique Labed2,
Makoto Arisaka1 and Isao Yamagishi1
1: Japan Atomic Energy Agency (JAEA)
2: CEA – DEN – DE2D – SEAD – LCBC

F03

Orano: Working in Partnership with Local Communities Over 50 Years ofRegional Integration
Xavier Verdeil, Maryline Breton and Daphné Ogawa
Orano Group

F04

Investigation and Characterization in hostile nuclear environment
Kenny Le Flanchec, Cédric Escoffier, Simon Haen, Sébastien Bouzat,
François Catsivelas, Frédéric Chambon and Daphné Ogawa
Orano Group

F05

TEΩ600, the industrial solution for robotics in hostile nuclear environment (up to
1MegaGray cumulated dose)
Alain Coudray1, Xavier Verdeil2 and Daphné Ogawa3
1: Orano Temis, 2: Orano Cycle, 3: Orano Japan

F06

Tools and technical means for under-water decommissioning in highly
contaminated environment
Jean-Michel Chabeuf1, Malo Diennet2, Alexandra Sykora3 andDaphne Ogawa4
1: Orano Dismantling & Services, 2: Ecole Centrale Supelec, 3: Orano GmbH,
4: Orano Japan

F07

DEM&MELT In-Can Melting Technology for the Vitrification of D&D andRemediation Waste
Regis Didierlaurent1, Thierry Prevost1, Daphné Ogawa1, Hubert Alexandreviii

Turc2 and Kohei Shibata3
1: Orano, 2: French Alternative Energies and Atomic Energy Commission(CEA), 3: ANADEC
F08

New peelable coating RTV FA 878 when dismantling the nuclear facilities
J.P. Thome1 and M. Sawamoto2
1: BCSN, 2: e-Energy Corp.

F09

Introduction of ROBATEL Industries business and their experiences regarding
hot cells project
Christopher Dane2, Natalia Zolnikova2, Masahiro Sawamoto1 andKeiichi Inoue1,
1: e-Energy Corporation, 2: ROBATEL Industries

F10

Mechanical Installations and Equipment for Decommissioning and for Waste
Storage
Craire Paoloni1, Vincent Gillet1, Renaud Bourdy1, Masahiro Sawamoto2 andKeiichi Inoue2
1: Cegelec CEM, 2: e-Energy Corporation

Session GB: Great Britain ･･････････････････････････････････････････････････････････････62
GB01

Pile Fuel Cladding Silo Decommissioning Project – Collaboration
Sarah Mundy and Roger Cowton
Sellafield Ltd.

GB02

Technologies for Detection of Contaminated Particles
Christopher Carter1, Jonathan Britton1 and Mark Rouse2
1: Cavendish Nuclear Ltd, 2: Cavendish Nuclear Japan

GB03

Integrating Waste Management with Project Management: the Key to Successful
Decommissioning
Bill Miller, David Luckhurst, Darren Dudley, Steve Caley and Paul Fleming
Wood plc

GB04

Innovative Cost Effective Solutions for In-Situ Liquid Waste Processing
Timothy Milner
ATKINS

GB05

The

importanceofcorrectly

identifying

andmanaging

the

cost

of

decommissioning nuclear facilities
Gary Mills
Gleeds Energy Limited
Session SK: South Korea ･･･････････････････････････････････････････････････････････････67
SK01

External & Internal Dose Assessment on Spent Resin Treatment Facility for
Recycling C-14
ix

Woo Nyun Choi, Ukjae Lee, Bae and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)
SK02

A Transportable Ground Radiation Monitoring System for Decision-Making on
Release of Nuclear Decommissioning Sites
Chanki Lee, Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK03

Dose calculation of dismantling workers for bio-shield of Nuclear power plant
ChoongWie Lee and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK04

Comparison of Detection Efficiency according to the number of Scintillator inContinuous Tritium Monitoring System
Jae Hoon Byun, Ki Joon Kang and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK05

Real-time Monitoring of Tritium in Water with Enhanced Detection Efficiency
Jun Woo Bae, Ki Joon Kang and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK06

The Measurement and Comparation Analysis of Change of Radioactivity
Concentration of the Water Containing HTO Molecule by Proton Exchange
Membrane Electrolysis
Ki Joon Kang, Jun Woo Bae and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK07

Fundamental Experiment of In-situ Beta Monitoring System for
Decommissioning Site
Min Ji Kim, Uk Jae Lee and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK08

Pre-treatment System for In-situ Tritium Monitoring in Seawater near
Decommissioning Site
Se Won Park and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

SK09

Fit-Goodness Evaluation of Peak Shape Fitting Function for Airborne Alpha
Beta Detection System
Si Hyung Seong and Hee Reyoung Kim
Ulsan National Institute of Science and Technology (UNIST)

Session US: United States of America ･･････････････････････････････････････････････････76
US01

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

US02

EnergySolutions Waste Management ExperiencexColin Austin1 and Makoto Kikuchi2
1: EnergySolutions, 2: MKC Consulting
US03

Horizontal Storage System for Used Nuclear Fuels, Fuel Debris and Nuclear
wastes – NUHOMS® MATRIX system –Ryan Buck1, Jane He1, Prakash Narayanan1, Vincent Tran2, Hiroshi
Akamatsu2, Yasuo Yamada2
1: TN Americas, LLC, 2: Transnuclear, Ltd.

US04

DBE PLUS+

Innovative Approach to Decommissioning

Takehiko Miyazaki
BechtelUS05

Application of UK Waste Management Performance at the Fukushima Daiichi
Decommissioning
Andy Beckwith and Max Ehrhardt
AECOM

List of Organizations ･･･････････････････････････････････････････････････････････････････81

xi

A01

A Study of Risk Analysis and Evaluation
during a Debris Removal Process
Akio Yamamoto1,2, Takashi Takata1,3, Satoshi Takeda1,4, Takashi Ino5
1

and Kazuki Hida5
Risk Analysis and Evaluation subcommittee, Review Committee on

Decommissioning of the Fukushima Daiichi Nuclear Power Station, Atomic Energy
Society of Japan, 2Nagoya Univ., 3JAEA, 4Osaka Univ., 5NDF
Abstract
Risk analysis and evaluation play an important role in decision making of the decommissioning process andplan in the Fukushima Daiichi Nuclear Power Station. In the present poster, activities of the risk analysis andevaluation subcommittee are introduced so as to support an establishment of the risk analysis methodology.
1. Objectives
A radiological risk is one of the most important issues and it will vary during the decommissioning process.
However, the risk analysis methodology has not been well considered during the decommissioning process because
the plant is already damaged and thus it includes plenty of uncertainty. Furthermore, each sub-process is notdesigned in detail. Therefore, the risk analysis and evaluation subcommittee support to develop the methodology
by reviewing each fundamental technique. A near-term goal of the development is to find important scenarios andto investigate a relative comparison among different processes of debris removing.
2. Review of Fundamental Techniques
2-1. Scenario assessment

Inadvertent radioactive release

MLD
analysis

Airborne release

Taking into account a complete uncertainty

Impacts on containment

in the risk, an exhaustive scenario assessment,

Degraded containment

which includes an initiating event and its
combination of top-down and bottom-up
down approach, a master logic diagram (MLD)
is applied. A hazard and operability study(follow below)
Impacts on risk sources
Increase in airborne debris

Filter Structure
OverPowder Fragmen- Entrain- Excess
failure damage pressure airborne
tationment
filings

progress, should be carried out. Consequently, aapproaches are considered. As concerns the top-

Waterborne release

Increase in FPs(follow below)

Event progression (hydrogen explosion, temperature increase,)

HAZOP/
FMEA
Analysis

Initiating events (nitrogen supply failure, cooling failure, debris fall, heavy drop, )Threats (random failure, human error, earthquake, )Figure 1. Scenario assessment with hierarchical tree

(HAZOP) and a failure mode and effects analysis (FMEA) are chosen as a bottom-up approach (Figure1).
2-2. Expert elicitation and quantification of radioactive release
An expert elicitation will be efficient to investigate an occurrence probability and magnitude of theconsequence of physical phenomena that include plenty of uncertainty. However, a variation becomes large in
general. An effective manner of the elicitation and implementation of some fundamental physical parameter into
the elicitation are discussed. A quantification of radioactive release is one of the key challenges in the risk analysis.
As a first step investigation, a simplified method with the multiplication of factors, which is applied to anassessment of nuclear fuel cycle facility accident [1], is applied and the applicability of each factor to thedecommissioning process is discussed.
References[1] U.S. NRC, NUREG/CR-6410, 1998.

1

A02

Research and Development of the Project of Decommissioning andContaminated Water Management and Connection to Preliminary
Engineering
Masaki Kawai1, Fumiya Sato1
Mitsubishi Research Institute, Inc. (Management Office for the Project of1

Decommissioning 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 developments with high technical difficulty. We will show the relationshipbetween the various subsidized projects of this program and the expected contributions to the decommissioningof 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 R&Ds; therefore, the Ministry of Economy, Tradeand Industry established a fund since fiscal 2013 and has implemented the "Project of Decommissioning andContaminated Water Management" as a grant-funded program by public bid to support R&Ds with hightechnical difficulty. Various R&Ds in the program have been managed by the Management Office for the Projectof Decommissioning and Contaminated Water Management. Mutual coordination among the researches isnecessary to apply the results of R&D to the decommissioning of Fukushima Daiichi NPP.
2. Subsidized Projects of Decommissioning and Contaminated Water Management Program andConnection to Preliminary Engineering
In accordance with the R&D plan for decommissioning[1], the subsidized projects are classified into
“Internal Investigation”, "Debris Retrieval" and "Processing of Solid Waste". The R&Ds of Debris Retrieval are
carried out based on the information inside the reactor vessel obtained by Internal Investigation. In addition, theresults of R&D such as Development of Debris Retrieval Method including the concretization of fuel debris
removal process are reflected in the preliminary engineering by the operating entities. The researches ofProcessing of Solid Waste are also studied in parallel with R&Ds for debris retrieval. In this way, the current
projects are closely related and aimed at engineering.
References[1] Ministry of Economy, Trade and Industry, Agency for Natural Resources and Energy, R&D plan for decommissioning in
FY 2019, Decommissioning and Contaminated Water Management Team Meeting / Project Management Office Meeting(63rd) (Japanese only)

2

A03

Value Chain in the Field of Decommissioning and Dismantlingof Nuclear Facilities
Mamoru Numata, Sadaaki Abeta, Takeshi Nitta,
Ken Kojima and Erwan Hinault
Kurion Japan K.K.,

Abstract
Veolia Nuclear Solutions (VNS) includes the most comprehensive range of technologies and services forfacility restoration, decommissioning of plants, and the treatment of radioactive waste, all nurtured by our nuclear
experts and backed by thousands of Veolia staff worldwide.
1. An Integrated offer within Veolia for the Nuclear Industry in Facility Clean-up

1.1. Characterization
VNS operates radiological laboratories in the U.S. and France. We have all the resources required to perform
onsite operations for clients based on recognized expertise in characterization and measurements: mobile pressure
and characterization units, mobile measuring equipment, sampling techniques, mapping and modeling tools.
1.2. Robotics
VNS is developing new remote handling methods in order to ensure the safety of its operators. These methods
can be used to decontaminate sensitive sites, which cannot be accessed by operators due to high levels ofradioactivity.
1.3. Separation and waste treatment
VNS deploys effluent treatment technologies as modular skid-based systems that allow for quick deployment,plug & play functionality, off-site commissioning and deployment of facility-based technologies to the mostdynamic nuclear environments.
1.4. Stabilization
VNS offers an economical, practical way to protect the environment from the threats of nuclear and hazardous
waste. GeoMelt® is ideal for solid waste and debris, while the Modular Vitrification System (MVS®) is designed
for liquid waste.

3

A04

Remote Access Technologies for Decommissioningof Nuclear Facilities

Mamoru Numata1, Sadaaki Abeta1, Takeshi Nitta1, Ken Kojima1,
Erwan Hinault1, Mark Sharpe2, David Loughborough2,
Scott Martin3, Matt Cole3, and Marc Rood3
1
Kurion Japan K.K., 2Veolia Nuclear Solutions (UK) Ltd.,
3
Veolia Nuclear Solutions Inc.

Abstract
Veolia Nuclear Solutions (VNS) introduced the inspection manipulator, repair manipulator, remote guide pipe
and inspection boom system for the primary containment vessel of the unit 2 reactor at the Fukushima Daiichi
nuclear power plant site.
1. Introduction
VNS develops technical solutions using remote technologies from concept, to final design, fabrication, testing,and field deployment. To date, the VNS team has leveraged its engineering capabilities to deliver more than 200
innovative remote systems, robotics and tools. This skill set includes the development of robotic manipulators,
inspection systems, specialty tools, waste handling systems, mobile waste treatment platforms and modular
containment systems. Equipment is commonly custom-designed or modified to meet the needs of our clients for
deployment in high-hazard environments where failure can result in significant safety concerns.
2. Reducing worker risk by advancing remote solutions
VNS’s remote access technologies and robotics eliminate human risk in the remediation of nuclear andhazardous waste in some of the most dangerous environments.
3. Fit-for-purpose engineering solutions and an agile, creative approach
Each radioactive or hazardous environment has a unique set of challenges related to cleanup, dismantlement or
decommissioning. While remote access to such unique environments requires distinct solutions, fundamental
knowledge and problem-solving are often applicable from one project to another. That is why it’s critical to find ateam that has solved similar problems before.
4. Application of remote handling technologies to the 1F site of TEPCOVNS is working in close collaboration with IHI/Toshiba and MHI on the design and construction of a unique,remotely controlled, robotic systems to inspect and repair the
suppression chamber and vent tubes using grout (IHI/Toshiba) andalso to investigate the nature and location of nuclear fuel debris in
the highly contaminated environment of the No. 2 of the Fukushima
Daiichi(MHI).
Investigation
Boom System

Guide Pipe System
Inspection Manipulator

Repair Manipulator

4

Dexter Manipulator

A05

GeoMelt® Application to 1F Waste

Sadaaki Abeta1, Erwan Hinault1, Jeremy Whitcomb2, Brett Campbell3,
and Kevin Funicane3
1
Kurion Japan K.K., 2Veolia Nuclear Solutions Inc.,3
VNS Federal Services Inc.,

Abstract
Veolia Nuclear Solutions’ GeoMelt® technologies are a group of vitrification processes that are configured
in a variety of ways to meet a wide range of radioactive and hazardous waste treatment and remediation.
GeoMelt® vitrification decompose organic wastes and immobilizes radionuclides and heavy metals in a stable
glass. Kurion Japan K.K. is investigating the applicability of GeoMelt® to treat and stabilize radioactive waste
generated by the treatment of contaminated water at the Fukushima Daiichi nuclear power plants.
1. Introduction
GeoMelt® creates a stable glass that is typically 10 times stronger than concrete, and more durable than graniteor marble. Its leach-resistance is among the highest of all materials in the world. Through vitrification technologies,
we offer an economical and practical way to protect the environment from the threats of nuclear and hazardous
waste. GeoMelt® is ideal for solid waste and rubble.
2. Industry Challenges and Features of GeoMelt®
Despite glass’s superiority, conventional vitrification technologies have been too complicated and far too
expensive to deploy widely. Such technologies also require a homogeneous waste feed and are less flexible than
GeoMelt®. Unlike most other vitrification technologies, GeoMelt® is cost effective and can process various types
of wastes simultaneously.
3. Development of GeoMelt®
GeoMelt® have been treating nuclear and hazardous waste since the 1990s, producing over 26,000 metric tonsof glass for disposal in the U.S., UK, Australia, Japan and other countries. Initially developed by Pacific Northwest
National Laboratory in the U.S., GeoMelt® has been used successfully around the world for the U.S. Departmentof Energy (DOE) at Hanford and at Sellafield in the UK for example.
4. Efforts for Wastes Arising from 1F
In Fiscal Year 2017 and Fiscal Year 2018, Kurion Japan and Veolia Nuclear Solutions successfully completed
two phases of testing which demonstrated the suitability of its GeoMelt® In-Container Vitrification (ICV)TM
process for the treatment and stabilization of Fukushima Daiichi contaminated water treatment secondary waste(i.e. zeolite absorbents, titanate absorbents, multi-nuclide removal equipment [ALPS] slurries, water
decontamination device [AREVA] sludge). This testing was conducted under the research program entrusted to
IRID (International Research Institute for Nuclear Decommissioning) as a METI (Ministry of Economy, Tradeand Industry) subsidy program.
This year, a holistic evaluation on the implementation of GeoMelt® In-Container Vitrification ICV™ is being
carried out under subsidy program of METI.

Sellafield Central Laboratory Facility

In-Container Vitrification GeoMelt® ICV™

5

A06

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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 “theConsortium”） is working to establish a network among manufacturing industries in Fukushima
Prefecture and related organizations, through seminars and robots demonstration events.
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 to disaster response robotics by utilizing their technologies developed
and accumulated in the decommissioning projects. This Consortium organizes seminars, robots
demonstration events and other products for networking and information exchange with relatedorganizations.
2. Main activities of the Consortium
2-1. Robots demonstration events and exhibitions
The Consortium exhibits its products at the exhibitionin Tokyo where many related people visit. In addition,
the Consortium holds a robots demonstration events in
JAEA to appeal new robots by demonstrating in front ofpotential 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.
2-3. Other activities
The Consortium members can be supported by expert
coordinators. Market survey and site visit to relatedorganizations, and information gathering fromgovernmental organizations are possible. Besides, latest
project information of new governments will be provided
timely to our members.

Figure 1. Robots demonstration events

Figure 2. Technology matching events

3. Unique Features of the Consortium
Because the Consortium is managed by Fukushima Technology Centre, a public research andtesting organization established by Fukushima local government, the members can exchange
information with companies 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

R&D of an Automatic Suppression Robot that can
Pinpoint the Source of an Incipient Fire and
Immediately Suppress the Fire for Nuclear Facilities
Yukio AKATSU, Shigeo KANKE, Kaoru AKATSU
Tokyo Bosai Setsubi Co., Ltd.

Abstract
It is difficult to fight a fire once a fire has developed and spread.

Automatic suppression robots

can extinguish the fire with a high degree of accuracy and with a minimum use of foam or anotheragent. When a fire occurs, people should safely evacuate.

These suppression robots can detect afire at the incipient stage and suppress the fire before manual firefighting is needed. Intelligent
AFEX or flying AFEX, automatic suppression robots help you focus on other decommission work.
Automatic Suppression Robots
1. “Intelligent AFEX (Automatic Fire Extinguisher)” (Fig. 1) is stand-alone, portable and can beplaced in locations with hazards and can be moved as decommissioning progresses. This
device detects and extinguishes fires in protected areas (Fig. 2).

Figure 1.
Intelligent AFEX

Figure 2.
Intelligent AFEX Protected Area

2. “Flying AFEX” (Fig. 3) is movable on the ceilingrail byreceivinga firealarm
signalfrom fire
detecters(Fig. 4).

Figure 4. Flying AFEX
Extinguishing Fire

Figure 3. Flying AFEX

8

B03

Development of underwater robot
Named “RADHOTAR”
RADiation HOsTile semiAutonomus Robot

Shigekazu Suzuki1, Shinji Kawatsuma1, Eiji Aoki1, Masanori Takahashi1,
Koki Watanabe2, Shuzo Nakano3, Kosuke Haga3, Hideyuki Hnawa4
1National Institute of Technology, Fukushima College,
2Takawa Seimitsu Co. Ltd.,
3East Japan Accounting Center Co. Ltd., 4Ascend Inc.

Abstract
National Institute of Technology, Fukushima College and several loco-enterprises have beendeveloping underwater robot “RADHOTAR”. The robot is aimed to be deployed in hostile field suchas Fukushima daiichi Nuclear Power Plants. The robot is to be small, radiation tolerant and highoperable in order to use inside of Primary Containment Vessels.
1. Background
Fukushima daiichi Nuclear Power Plants, attacked by huge
Tsunami following Tohoku Region Pacific Coast Earthquake on
March 11th, 2011, fall into melt down and hydrogen explosions, which
released radioactive materials inside of primary containment
vessels.
In order to aid Decommissioning, the college and the enterprisestarted to develop under water robot “RADHOTAR”.

Fig.1 Fukushima daiichi
NPPs after explosion [1]2. Features and technology under developing
The RADHOTAR robot is planned to be small, radiation tolerant
and high operability. [2] A direct drive thruster mechanism newly
developed to be small in order to enter the vessels through narrowtube. Semiconductors are reduced and selected as far as possible inorder to increase radiation tolerance. [3] Autonomous control using
image processing and Structure from Motion technology have been
applied in order to increase operability.

Fig.2 RADHOTAR robot(110 mmΦ×600mmL)

3. Acknowledgement
This RADHOTAR is being developed under an assistant project of Fukushima Prefecture.
References[1] http://photo.tepco.co.jp/library/110316/110316_1f_chijou_2.jpg[2] Suzuki, et al, “Development of underwater robot " RADHOTAR”, (1) Overall plan”[3] Kawatsuma, et al, “Development of underwater robot “RADHOTAR”, (2) Radiation Tolerance
Examination”

9

B04

Disaster Response Robots ‘Giraffe’ Developed by the Members of
Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response
Noritaka Baba
Aizuk, Inc.

Abstract
One of the biggest task for disaster response robot is consider about peaceful use on a daily basis. Because of
the not use on a daily basis. Sometimes robot cannot action when the disaster occurs. Therefore, we havedeveloped a base robot for running which easily customize aiming at the peaceful use on a daily basis of disaster
response robots.

1.

Specifications of Disaster Response Robots ‘Giraffe’[Fig1]
Giraffe is a crawler type robot which has two main crawlers and four sub
crawlers. Four sub crawlers work independently and it enable to run roughterrain.

2.

Size(L*D*H)[mm]

1060*448.7*382.1

Gland clearance[mm]70Driving method(main crawler)

Left and right independent crawler drive

Auxiliary run unit(sub crawler)

Four independent sub crawler system

Belt drive system

Chain and sprocket drive

Battery

Li-ionOperating time

2hour

Maximum inclination angle

45°

Price(Minimum configuration)¥3,000,000

Fig1.Giraffe

Fig2.WRS

Joint research and development with University of Aizu.
Aizuk conclude cooperation agreement with University of Aizu in 2015. We have been working on
standardization of robots compliant with Open-RTM.
In 2018 we won the WRS Tunnel Disaster Response Recovery Challenge category[Fig2]

3.

Cooperation with the local manufacturing industry.
Giraffe was designed by Aizuk and manufactured by
SAKAE-SEISAKUSYO which in the Minami-Soma.
We want active development of Giraffe by each vendor.

10

B05

Organization Profile of IRID
Kazufumi Aoki1
1
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

11

B06

Overview of IRID R&D Projects
1

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

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

12

耐放射線軟質ＰＶＣ材料

B07

Radiation soft PVC materials-resistant
江草

史典、山中

智之

Huminori Egusa and Tomoyuki Yamanaka

Ｍタイガースポリマー株式会社
Tigers Polymer Corporation

Abstract
We developed radiation soft PVC materials with the durability of gamma beam
exposure dose 5MGy-resistant.
I keep elasticity, flexibility even after irradiation, and the molding at various shapes,
diameters such as hose tube seats is possible.
In addition, materials properties of matter are good and are superior in wear
resistance.
Because it is PVC materials, I am superior in cost performance.
1. Gamma ray irradiation examination【 Hardness retention after
irradiation 】
120

Tensile elongation retention（％）

Hardness retention （％）

115
110
105
100
95
90
85
80
75
70

【 Elongation retention after
irradiation 】
120
①

100

②

80

③

60
40

①

20

②
③
0

2

4

Total irradiation dose （ＭGy）

0

6

0

2
4 （Ｍ
Total irradiation
dose
Gy）

Sample①：Radiation-Resistant Flexible PVC Material（Developedmaterial）
Sample②：Soft PVC for general hose (reference data)
Sample③：Olefinic soft material (reference data)

2.Sample

Tube

Hose

13

Sheet

6

B08

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.

14

B09

Monitoring technology for fuel debris removal process
in Fukushima decommissioning
Akira Kobayashi

Hideyuki Suzuki

HAMAMATSU PHOTONICS K.K.
Abstract
We have developed radiation tolerant image pick up tube and camera for real time
monitoring in PCV and RPV in fuel debris removal process. The objective specifications ofradiation tolerance are dose rate of 10 kGy/h and cumulative dose of 2 MGy.
Newly developed image pick up tube and camera

Image pick up tube

Camera system

Evaluation of radiation tolerance
Each step on gray-scale chart can be recognized under 60Co gamma ray exposure of 10
kGy/h even after continuous operation and radiation for 200 h.

Non exposure

Start

200 h

Image samples under 60Co gamma ray exposure of 10 kGy/h

15

B10

Introduction of Mirion Technologies’ Radiation Tolerant Camera
The World’s Highest Radiation Tolerant Performance Camera
Yasuo Arai, Mikio Katsura and Toshiki Matsui
CORNES Technologies 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 is recognized as themarket 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 to meet customer needs andexpectations.

1. HYPERION Solid-State, 1M Gy Radiation Tolerant & Robust Camera

The launch of the Hyperion™ solid-state camera follows many years of extensive
research and development activity and considerable investment in new technologies.
Featuring all new Radiation Tolerant Zoom lens and optics, with integrated pan and tilt,
the Hyperion camera has been independently tested to 1MGy with Co-60 source. Using
an FPGA platform based on SoC (System-on-Chip) technology, the Hyperion camera
allows upgrades without changing hardware.
z 100Mrad / 1MGy Total Dose (Gamma)z High performance Solid-State 1 Megapixel Sensor
z Low maintenance and long lifez Superior picture geometryz High sensitivity and low noisez New Integrated Pan/Tilt/Zoom Stainless Steel Outstationz In air or underwater operationz Compatible with existing Mirion control racks and cables

2. 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 a compact sizewith a large range of viewing attachments and options.
z Radiation Tolerance: 2MGy (2 x 108 rad), 1kGy/hr dose ratez Resolution: 600 TV lines per picture height (center zone)
z 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)

3. 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.
z Radiation Tolerance: 1MGy (108 rad), 1kGy/hr doze ratez Resolution: 550 TV lines, horizontal
z 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

4. RC911 DotCam Miniature Color Camera

The DotCam is a miniature color camera designed for use in higher radiation environments where the use of standard 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.
z Radiation Tolerance: 1kGy (105 rad), 300Gy/hr doze ratez Resolution 400 TV lines (NTSC), 450 TV lines (PAL)z 4mm standard lens (FOV 90 deg)
16

C01

Development of Dosimeter for Severe Radiation Environment
near Reactor Pressure Vessel
Tamotsu Okamoto1, Yasuhito Gotoh2, Masafumi Akiyoshi3,
Mitsuru Imaizumi4, Tomohiro Kobayashi5, Yasuki Okuno6
1

NIT, Kisarazu Coll., 2Kyoto Univ., 3Osaka Pref. Univ., 4JAXA, 5RIKEN,

6

JAEA

Abstract
High dose-rate radiation detection is required for Fukushima decommissioning. Therefore, a compact
and radiation-tolerant dosimeter without power supply using the solar cells was proposed. In this work,radiation-detection characteristics of the solar cells such as CdTe and InGaP solar cells were investigated.
Radiation-induced current density due to 60Co γ-ray irradiation was evaluated in the CdTe and InGaP solar
cells, and it was found that radiation-induced current density increased linearly in the range up to
approximately 2000 Gy/h.
High dose-rate radiation detection under severe environment such as high-level radiation, hightemperature and high humidity is required for decommissioning the Fukushima Daiichi nuclear power
station (FDNPS). Gamma dose-rates near reactor pressure vessel (RPV) are predicted to be very high suchas over 100 Gy/h, because the nuclear fuel debris and 137Cs are trapped. In addition, no power supply isdesired in view of residual hydrogen. Furthermore, a compact detector is required in order to introduce
through a small hole of approximately 10 cm. Therefore, we proposed a compact and radiation-tolerant
dosimeter without power supply using the solar cells. In this work, we investigated the radiation-detection
characteristics of the solar cells such as CdTe and InGaP solar cells.

both solar cells, radiation-induced current
density increased linearly in the range up to
approximately 2000 Gy/h. This result
suggests that dose rate is easily measured
using the solar cells.
This work is financially supported by
the Nuclear Energy Science & Technology
and Human Resource Development Project(through concentrating wisdom) from theJapan

AtomicEnergyAgency/Collaborative Laboratories for Advanced
Decommissioning Science.

2

the dose rate of 60Co γ-ray irradiation. In

Induced Current Density (nA/cm )Figure 1 shows radiation-induced current density from a CdTe and an InGaP solar cells as a function of150
CdTe solar cell
InGaP solar cell
100

50

0
0.0

0.5

1.0

1.5

2.0

60

Dose Rate of Co γ-ray Irradiation (kGy/h)Figure 1 Radiation-induced current density from a
CdTe and an InGaP solar cells as a functionof the dose rate of 60Co γ-ray irradiation.

17

C02

Progress of PYRAMID project- Piping system, risk management based on wall thinning monitoringand prediction –Toshiyuki Takagi1,2, Philippe Guy3, Yutaka Watanabe2,4, Hiroshi Abe4, Shinji Ebara4, Tetsuya
Uchimoto1,2, Takayuki Aoki4, Mitsuo Hashimoto1, Ryoichi Urayama1, Hongjun Sun1, Thomas
Monnier3, Jérôme Antoni3, Bernard Normand3, Nicolas Mary2, Ryo Morita5, Shun Watanabe5, Atsushi
Iwasaki6, Hiroyuki Nakamoto7, Christophe Reboud8, Pierre Calmon8, Edouard Demaldent8, Vahan
Baronian8, Xavier Artusi8, Sylvain Chatillon8, Alain Lhemery8
1
2
3

Institute of Fluid Science, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Japan

ELyTMaX UMI 3757, CNRS – Université de Lyon – Tohoku University, Sendai, Japan

Université de Lyon, INSA-Lyon, MATEIS CNRS UMR 5510, Villeurbanne F-69621, France

4

Graduate School of Engineering, Tohoku University, 6-6-01-2 Aoba, Aramaki, Aoba, Sendai, Japan

5

Central Research Institute of Electric Power Industry, 2-6-1 Nagasaka, Yokosuka, Kanagawa, Japan
6
7

Department of Mechanical System Engineering, Gunma University, Kiryu, Japan

Graduate School of System Informatics, Kobe University, Nada, Kobe, 657-8501, Japan
8

CEA-LIST, Gif-sur-Yvette, France

1. Introduction
Cooling water circulation is an important guarantee for the safety of Fukushima Daiichi Nuclear Power Plant
during decommissioning. However, when removing fuel debris, a flow with a high concentration debris of various
kinds occurs in cooling water pipe. Pipe wall thinning by Slurry Flow induced Corrosion (SFC) under solid-liquid
two-phase has been anticipated. This may seriously affect the safety of the cooling system. 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. Progress of the project2.1. Modeling of SFC and prediction of wall thinning. The wall thinning evaluation model is developed basedon the wall thinning rate evaluation considering the mass transfer coefficient under solid-liquid two-phase flow.
Then the SFC is elucidated with experiment and numerical simulation.
2.2. Development of EMAT monitoring system. The focusing type EMAT is designed for applying to monitoring
system. The designed EMAT is evaluated with experiment and numerical simulation. In addition, a new dataprocessing method is proposed to improve the evaluation accuracy of pipeline wall thickness measurement.
2.3. Engineering risk evaluation. Based on PoF (Probability of Failure) evaluation in consideration of the various
errors (inspection / damage progress and applied force) by Bayesian estimation and Reasonable plan by integrativeevaluation of factors using a risk matrix, a quantitative evaluation method for engineering risks associated with
SFC is proposed.
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 Developmentby the Ministry of Education, Culture, Sports, Science and Technology of Japan, and ANR of France.

18

C03

Development of directional gamma-ray detectors
for Fukushima decommissioning
Mitsuhiro Nogami1, Keitaro Hitomi1, Tatsuo Torii2, Yuki Sato2
Yoshihiko Tanimura2, Kuniaki Kawabata2, Yoshihiro Furuta2, Hiroshi Usami2

Kenichi Watanabe3, Toshiyuki Onodera4, Keizo Ishii1, Kenji Shimazoe5, Hiroyuki Takahashi5
1

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

Abstract
Our group developed directional gamma-ray detectors for Fukushima decommissioning. The directional
gamma-ray detector consisted of a lead absorber and two thallium bromide(TlBr) spectrometers. The performanceof the detector was evaluated using a 137Cs checking source at room temperature. The directional gamma-ray
detector exhibited a good directional performance suitable for Fukushima decommissioning.
1. Introduction
In promoting Fukushima decommissioning, directional
gamma-ray detectors operable in intense radiation fields areindispensable for identifying radioactive materials in thecontaminated areas. Thallium bromide (TlBr) is a compound
semiconductor characterized by high photoelectric absorption
efficiency. It has been confirmed that small-volume TlBr detector
can operate in intense radiation fields. In this study, a directional
gamma-ray detector consisted of a lead absorber and two TlBr

Fig. 1 Directional detector
fabricated in this study.

spectrometers was developed for Fukushima decommissioning.
2. Fabrication of directional radiation detector
TlBr spectrometers were fabricated from TlBr crystals. The
device had a planar electrode (3.5 mm × 3.5 mm) on the cathodesurface, and a pixel electrode (0.5 mm × 0.5 mm) and surrounding
guard electrode on the anode surface. The directional detector has alead absorber with a dimension of 1 mm × 5 mm × 20 mm. Fig.1
shows a directional detector fabricated in this study.
3. Experiments and results
The directional detector was evaluated with a 137Cs

Fig. 2 Setup for directionality
measurements.

checking source at room temperature. Directional responses ofthe detector were recorded by changing the position of theradio isotope source around the device as shown in Fig.2.
Figure 3 shows the directional response of the detector.
As can be seen from the figure, the device exhibited clear
directionality. The details of the experiments will be described
Fig. 3 Directional responseof detector.

at the conference poster.

19

C04

Can cesium be adsorbed through the leaf and bark of trees?
– Estimation of the dominant absorption route of cesium in treesTomoko Ohta1,2, Junji Torimoto3, Yasunori Mahara4
1

CRIEPI, 2The University of Tokyo, 3JAMSTEC, 4Kyoto University

Abstract
After water-soluble Cs was applied to the leaves and bark of eight pine trees, the Cs was tracked in the whole bodyof the tree. Although Cs on both the bark and the leaves was transferred into the xylem of tree, Cs on the bark was
not taken up into the trunk more effectively than that on the leaves.
1. Introduction
Observing the radiocesium distribution in annual tree rings in fields was limited in early period after theaccident. It is necessary to elucidate the dominant route of uptake of radiocesium by the multidirectional
comparison methods. We are tried to tracer test to confirm the dominant route of radiocesium uptake from theleaves and bark of pine, which is a popular evergreen coniferous tree (Pseudotsuga menziesii var. glauca) in Japan,
and to estimate the translocation velocity of cesium in xylem in this study.
2. Distribution of cesium in tree
A water-soluble form of cesium was applied to the leaves or bark of pine trees to elucidate the translocationroute in the tree. A 10000-ppm solution of water-soluble cesium was made using cesium chloride. The cesium
solution was applied to all the leaf surfaces of four trees and to bark on the trunks of five trees. The sample trees
were cultivated to 30 days after the cesium solution was applied to the leaves or bark.
After cultivation, the sample(b) applied to bark(a) applied to leaf

trees

were

cut

down

and

separated into several parts.
Concentration of cesium wasmeasured by ICP-MS.Fig. 1
shows the distribution of cesium
1

2

3

4

5

1

2

3

4

5

6

in the tree.

Fig. 1 Distribution of cesium in tree.
(a)1: branch, 2: leaf (Applied to cesium) 3: leaf 4: trunk, 5: root , (b) 1: branch, 2: leaf 3: trunk, 4: root, 5:xylem,
6: bark (applied cesium)
The results of these tracer experiments indicated that stable cesium enters xylem from foliage and bark.
However, radiocesium is quantitatively taken up into the xylem of trunk from leaves than from bark. This result
supported that result of our field study the accident after 1.5 years: the amount of radiocesium in the lower trunk
derived from the leaf surfaces of Cryptomeria japonica was markedly greater than that derived from the bark,though tree species is different.
3. Conclusion
Cesium on leaves and bark was translocated to through bark and foliage to xylem absorption within 30 days.
However, in pine trees, cesium is more rapidly translocated from the leaves to the xylem of the trunk than fromthe bark to the xylem of the trunk. Dominant route of the radiocesium into xylem was leaves.

20

C05

Radiochemical analysis of actual samples collected at the
Fukushima Daiichi NPS site

Masashi Ichikawa1, Masaaki Kaneko1, Yoshikazu Koma2,3, Hiroshi Hinai2,3, Ayako Nitta2,3,
Atsuhiro Shibata2,3
1

Nippon Nuclear Fuel Development Co., Ltd., 2Japan Atomic Energy Agency (JAEA),
3

International Research Institute for Nuclear Decommissioning (IRID)

Abstract
As a study on treatment and disposal of solid waste generated by the
Fukushima Daiichi NPS (1F) accident, the radionuclide composition of samples
collected from the bottom of the primary containment vessel (PCV) of unit 1 was
acquired by radiochemical analysis. This is the first time that radionuclide
composition data of actual waste samples taken from inside the reactor have beenobtained, and the acquisition of these data will make it possible to estimate theorigins of the radionuclide contaminants.
1. Introduction
In the 1F accident, radioactive waste with a different nuclide composition
from that of conventional waste was generated. To consider the treatment anddisposal of this waste, it is necessary to know the radionuclide composition of
actual samples. In April 2017, samples were obtained of soil-like sediments
present at the PCV bottom of 1F Unit 1 [1]. Some samples were transported to
the Nippon Nuclear Fuel Development Co. Ltd. for radiochemical analysis.

Figure 1. Sample
from the PCV
Sample from the PCV←aqua regia, HF
Heat・Dissolution
Dissolved solution
Filtration

2. Radiochemical analysis
The photo of a contaminated soil-like sediment sample obtained from theFiltrate
PCV is shown in Figure 1. Figure 2 summarizes the sample preparation
procedure for radiochemical analysis. A sample was dissolved in aqua regia
Measurement
and hydrofluoric acid with heating. After filtration of the acid solution, theNuclides and elements
filtrate was analyzed for radionuclides including those of “difficult-tomeasure” and elements. Table l shows some of the measured radionuclides and Figure 2. Sample preparation
elements. Actinides (U, Np Am, Cm) and fission products (such as 90Sr, 137Cs
for radiochemical analysis
and 154Eu) were detected. Also, activation products ofZircaloy (such as 93Zr), and of stainless steel (such as Table 1. Measurement of radionuclides and elements
Co and 63Ni) were detected. Difficult-to-measurenuclides (such as 55Fe and 121mSm) were detected
after carrying out chemical separation procedures. As
analysis of these nuclides has become possible, anestimation of the origins of these nuclides will bepossible.
60

Origin
Fuel
Fuel cladding（Zry-4)Structurematerials(SUS etc.)

Radionuclide
90Sr, 99Tc, 106Ru, 110mAg, 125Sb,

126Sn, 137Cs, 154Eu, isotope
237Np, isotopeof U,of Pu, Am/Cm

Element
U

93Zr, 121mSn, 125Sb

Zr

55Fe, 60Co, 63Ni, 94Nb, 93mNb

Fe

Acknowledgement
This report includes part of the results obtained from the work “Research and Development of Processing andDisposal of Solid Waste”, carried out on a supplementary budget allocated for Project of Decommissioning andContaminated Water Management.
References[1] TEPCO, 62nd Commission on Supervision and Evaluation of the Specified Nuclear Facility (10 Aug 2018).

21

C06

Estimation of Aging Properties of F1 Fuel Debris
Oleg Bagryanov1, Sergei Poglyad2
TENEX JSC*, 2SSC RIAR JSC*

1

Abstract
Rosatom successfully realized the project «Estimation of Aging Properties of Fuel Debris», main objective of
which was to make a comprehensive forecast of FD aging in the range of up to 50 years and to estimate theprobability of such aging depending on the initial material parameters and storage conditions.
Within the project model samples with & without Cm imitating FD of F1 were manufactured and studied. Obtained
experimental data has been subjected to verification procedures based on the studies of natural geological systems,
vitrified HLW and Chernobyl lavas.
1. Introduction
The application of the prediction model had the aim to assess the worst possible case of radioactive materials
release into the media contacting with the FD (FD of the core & concrete melt zone periphery).
2. The following tasks have been completed within the project- Prior research of Chernobyl fuel containing materials summarized;- Research program for the preparation and testing of the F1 FD
model samples developed;- 14 model uranium-plutonium model samples without Cm & 22 Cmcontaining samples manufactured;- Experimental studies of such processes of FCM aging, as the change
in mechanical strength, leaching of different radionuclides, thechange in the phase composition performed;- Experiments for exploring two factors of degradation – radiation
factor & chemical factor fulfilled;- Obtained experimental data has been subjected to verification
procedures based on the studies of natural geological systems,
vitrified HLW and Chernobyl lavas.

Figure 1. Metallic Phase

2-1. Fe is corrodible
The study of the obtained experimental data allowed us to formulate a hypothesis about the possible scenario of
corrosion process. In the presence of the alpha radiation source there is an increase in the rate of Fe leaching, as
well as the generation of hydrogen peroxide. The rapid decomposition of hydrogen peroxide leads to theoxidation of Fe, which in turn leads to the “red” suspensions formation. Such “red” suspensions can put
additional stress on the safety systems that will function during the implementation of the FD retrieval activities.
3. Conclusion- Leaching rate of controlled radionuclides of silicate samples iscommensurate with leaching rates of the same radionuclides ofvitrified HLW matrixes;- Cumulative radiation-chemical damage leads to greater
degradation of fuel fragments, than total radiation and chemical
effect;- Ferrous hydroxide is a good coagulant of Pu and U and otherelements, except for Cs, it will lead to formation of larger
quantities of small particles and hazardous fine dust. Delay inremoval of fuel fragments after 20-25 years after an accident
can lead to complication of dose and radiation environment,
because of increased emission of plutonium, its further transit
into colloidal forms and solid particles of submicrometer
dimension.

*Companies of Rosatom group

22

Figure 2. Silicate PhaseD01

Cesium decontamination of soil by combined techniques of elution,
adsorption and electromigration

Haruaki Matsuura1, Uladzimir I. Kislyi2, Vladimir M. Dudarchik2,Yasuhiro Kawashima2,
Paul G. Molchanov2, Paul M. Bulai2, Victor V. Boksha,2
1

Tokyo City Univ., 2Nuclear water soil solutions

Abstract
Cesium decontamination of ground soil are now developing in a collaboration with Japanese and
USA-European joint team. Feasibility study has been launched and at first, the preliminary experiment has been
performed by using USA-European technique in Japanese facility. Two kinds of ion chelate eluents wereintroduced to extract radioactive cesium from the soil and eluted cesium were collected in sorbent wrapped
cathodes by aid of electromigration.
1. Introduction
It is one of the critical issues of volume reduction of decontaminated soil which is widely spread due to the
Fukushima Daiichi Nuclear Power Station. Our developing cesium decontamination technique for soil is asfollows; 1) radioactive cesium is eluted by two kinds of eluents, 2) a direct current electromigration technique isapplied to concentrate cesium in the solution, and 3) concentrated cesium is adsorbed into sheet form of sorbents
located near anodes. The final decontamination target is the fertile soil, and we try to develop thedecontamination technique using ecological and economical material to avoid from damage to the soil. We havelaunched feasibility study of combined technique which has been orginally developed for solving the problems
against Chernobyl accident using Cs contaminated soil which was sampled in Fukushima prefecture.
2. Experiment
Sampled soil (ca. 20000 Bq/kg) in ammonia basic solution had been well stirring mixed with two kinds ofeluents as well as without eluents for the reference at 50oC for 3days. Then an electromigration under thecondition of 40 V between carbon turbular anode and several cathodes wrapped by sorbent sheets have been
applied to two solutions for 2 hours each. The concentration efficiency has been evaluated by gamma
spectroscopy using Ge semiconductor spectrometer (SEIKO EG&G, measurement duration; 86000 s per eachsample) which is focused on 137Cs in sorbent, soil and solution.
3. Results and discussion
Concentration efficiency which is derived from supernatant is 6.13. Therefore two types of eluents (one issurface acitivated material and the other is natural degraded material) has clearly effective to decontamination ofsoil. However, concentration efficiencies which are derived by the radioactivity of sorbent devided bysupernatant are 22100 for mixture with eluents and 70500 for no eluents. The reason of this fact would be as
follows: affinity to eluents and cesium disturb the sorption or eluents affect transport phenomena under theelectromigration. As shown in preliminary experiments, it is confirmed that eluent and sorbent works well,Further task is characterization of elution and adsorption ability both in equilibrium and kinetic point of view,and finding out optimum condition of electromigration.

23

D02

Adsorption Properties of Actinoids (U, Am, Np)
for Various Inorganic Adsorbents
Hitoshi Mimura1, Minoru Matsukura1, Tomoya Kitagawa1, Fumio Kurosaki1,
Natsuki Fujita2, Hitoshi Kanda2, Akira Kirishima2, Daisuke Akiyama2,
Nobuaki Sato2, 1UNION SHOWA K.K., 2Tohoku University.

Abstract
Adsorption properties of actinoids (U(VI), Am (III), Np(V)) for 19 kinds of inorganic adsorbents (zeolites,etc.) were evaluated under different solution conditions. The adsorption of actinoids (An) are shown by a
distribution curve consisting of ion exchange, surface adsorption of hydrolysis species and sedimentation
depending on equilibrium pH; the pH at the maximum is influenced by hydrolysis pH of each actinoid, and the Kdvalue depends strongly on pore size, Si/Al ratio and coexisting cation concentration, etc.
1. Introduction
The basic adsorption behavior of An on inorganic adsorbents should be clarified from the standpoint of
accurate inventory estimation and safety disposal, and the decontamination of An leached from the crushed fueldebris becomes more important in near future. In this study, the practical adsorption properties of An for various
inorganic adsorbents were evaluated by batch method under different solution conditions.
2. Results and Discussion
2-1. Adsorption properties of An in simple solutions (HNO3, HCl, HClO4, NaCl)
The distribution of U(VI) and Am(III) ions depended on equilibrium pH; the Kd value increased markedly withequilibrium pH and had a maximum around neutral region, and then tended to decrease in alkaline region. Zeolite
A with the lowest Si/Al ratio (Si/Al=1) had a relatively large Kd value of U(VI) above 103 cm3/g around pH 6∼8,and zeolite L with large pore size (7.1∼7.8 Å) had the largest Kd value of Am(III) ions over 104 cm3/g at pH 3. On
the other hand, the Kd value of Np(V) ions having a high hydrolysis pH tended to increase with equilibrium pH up
to 8; the Kd value for zeolite X was estimated to be over 100 cm3/g in the presence of 0.1 M NaCl-10-3 M HCl.
2-2. Adsorption properties of An in boric acid/seawater㻝㻜㻜

The uptake (%) of U(VI) in 3,000 ppm boric acid/ 30%㼀㼕㻻㻞㻮㼑㼚㼠㼛㼚㼕㼠㼑㻷㼀㼍㼘㼘㼛㼜㼔㼑㼚㼑㻵㼄㻱㻙㻤㻜㻜

㻵㻱㻙㻥㻝㻝

㻵㼄㻱㻙㻝㻞㻜㻜㻳㻯㻼㻹㻯㻞㻠㻢㻜㻏㻔㻺㻹㻕㻵㻱㻙㻥㻢㻥㻜㻜㻺㼍㻔㻿㻹㻕㼁㻻㻼㻸

antimonate (IXE-1200G) in acidic region due to the ion

㻜㻝㻟㼄above 80% was obtained for CST (IE-911) and tin

㻞㻜

㼁㻿㼅

carbonate complex. Relatively large uptake (%) of Am(III)㻠㻜㼄㻙㻢㻝㻶㻴㻼

lowered below 20%, due to the formation of Am(III)㻢㻜㻭㻙㻡㻝㻶㻴㻼

The uptake (%) of Am(III) for zeolites was considerably㻤㻜㼁㼜㼠㼍㼗㼑㻌㻔㻭㼙㻘㻌㻑㻕

diluted seawater was above 20% for zeolites A, X and Y.

exchange (Fig. 1). As for Np(V), zeolite X (13X) had a

Figure 1. Uptake (%) of Am(III) in boric

relatively large Kd value around 100 cm /g.

acid/seawater3

3. Conclusion
Precise estimation of chemical species of An in the contaminated water is very important for the fuel debris
treatment in Fukushima NPS, and the inorganic adsorbent for the decontamination ofAn should be selected byconsidering the solution conditions.
References : H. Mimura, et al., Proc of GLOBAL2015, Proc. of ICONE2018, Proc. of ICAPP2019.

24

Stable Solidification of Radioactive Wastes
by Zeolites and Allophane

D03

Hitoshi Mimura1, Minoru Matsukura1, Tomoya Kitagawa1, Fumio Kurosaki1,
Yuki Ikarashi2, Xiang-Biao Yin2, Chun-Nan Hsu2, Yan Wu3
UNION SHOWA K.K., 2Tohoku University, 3Shanghai Jiao tong University

1

Abstract
Zeolites and allophane have three functional abilities effective for the treatment of radioactive wastes; nuclide
adsorption, Cs trapping and self-sintering abilities. These excellent abilities are applicable to the stable
solidification of insoluble ferrocyanide, Cs selective composites, high-level radioactive wastes and uranium oxide.
1. Introduction
Development of stable solidification method contributes to the advancement of decontamination system andenvironmental remediation at Fukushima NPP-1. The stable solidification of radioactive solid wastes using
functional abilities of zeolites and allophane has been studied by UNION SHOWA K.K. and Tohoku University[1][4]. This paper deals with the results of stable solidification of the above radioactive solid wastes.
2. Results and Discussion
2-1. Stable solidification of Cs-insoluble ferrocyanides with zeolites
By using the immobilization ability of zeolites, i.e., Cs trapping andself-sintering abilities, the discs of CsCoFC/zeolites (CP, SA-5, MC, A51J) mixtures were sintered at higher temperatures up to 1,200ºC, and theCs immobilization (%) for the sintered products at 1,000℃ is estimatedto be above 99% (Fig.1), indicating that Cs volatilization was
effectively depressed by mixing of zeolites.
2-2. Stable solidification of Cs-selective composite adsorbents

Figure 1. Cs immobilization ratio of
sintered products of CsCoFC/zeolite

Cesium selective adsorbents of insoluble ferrocyanides (NiFC, CoFC) and hetropolyacids (AMP, AWP) canbe loaded on macropores of granular zeolites, porous silica gels and alumina. These composites are expected forthe selective adsorption of Cs in contaminated water and wash waste water of incineration ash. As for loaded
composites of zeolite, the stable solidification can be accomplished by direct sintering. The loaded composites ofporous silica can be converted to the stable solid form by adding allophane.
2-3. Stable solidification of denitrated precipitates of HLLW and uranium oxide
Zeolites can be applied to the stable solidification of denitrated precipitates of HLLW. The 1:1 mixture of A
type zeolite and denitrated precipitate containing Mo, Zr and La, etc. was converted to the stable solid form
consisting of major crystal phase of Haüyne after sintering at 1,200℃ for 1 h. The 1:1 mixtures of uranium oxide(U3O8) and zeolite (chabazite, allophane) can be converted to the stable solid forms by press/sintering at 600℃.
These sintered products had excellent mechanical strength and relatively low elution ratio of U.
3. Conclusion
The excellent immobilization abilities of zeolites and allophane are widely applicable to the stable
solidification of radioactive wastes generated from the decontamination and decommissioning in Fukushima NPS.
References [1] Y. Ikarashi, et al., Proc. of WM2013, [2] H. Mimura, et al., J. Ion Exchange, 22(3) (2011)1-13,[3] H. Mimura, et al., Ion Exchange, 23(1)(2012) 1-15, [4] H. Mimura, et.al., J. Ion Exchange, 23(2)(2012) 1-14.

25

E01

Major Initiatives for Water Management at the
Fukushima Daiichi NPS
Tomoya Kaneda
Tokyo Electric Power Company Holdings Inc.

Abstract
We have taken measures against contaminated water caused by earthquake and accident. As
a result, we have reduced the rate of contaminated water converted, have made progress inremoving contamination from contaminated water, and have made progress in preventing leakageof contaminated water. I will explain the status of our measures against contaminated water.
Outline of contaminated water treatment
At the Fukushima Daiichi NPS, some of the groundwater flowing from the mountain-side to
the sea is entering into the nuclear reactor building, converting into newly contaminated water.
For this reason, we are implementing various measures to counter the risk of contaminated waterflowing into the port of the power station and the risk of contaminated water flowing out from thestorage tanks. These measures are based on our three basic policies of "Eliminate contamination
sources", "Isolate water from contamination", and "Prevent leakage of contaminated water".
■Eliminate contamination sources
・ Clean up contaminated water with Advanced liquid
Processing System (ALPS)
・ Removing contaminated water in trenches
・ Removing contaminated water in buildings■Isolate water from contamination
・ Pumping up groundwater through groundwater bypass
・ Pumping up groundwater through wells near buildings
・ Installation of frozen-soil impermeable wall on the land side・ Paving of site to curb permeation of rainwater into soil
・ Repair damaged roof
■Prevent leakage of contaminated water・ Ground improved with water glass
・ Installation of impermeable walls on the sea side・ Augmentation of tanks (replacement with welded tanks, etc)

Figure 1. Main Facilities for Water Management
By implementing these measures, we have achieved the following.


Reduce the rate of newly contaminated water converted from approximately 490 tons/day into approximately 170 tons/dayComplete the removal of contaminated water from the Unit 1 turbine building and radioactive
waste disposal building

Replace the assembly tank with a leak-proof welding tank.

We plan to continue these efforts and achieve the following by the end of 2020
Reduce the rate of newly contaminated water converted to approximately 150 tons/dayComplete the removal of contaminated water in the buildings except for the Unit 1 ~ 3reactor building, where water is being injected to cool fuel debris

26

E02

Development of Laser Decontamination Method for Dismantling
Flange Tank
Naoki Maeshiro1, Tomohiro Ogawa1, Nobushige Sasaki1, Ryohei Takeuchi2,
Fumio Ito2, Haruo Nagamine2 and Hiroshi Nakamura3

Tokyo Electric Power Company Holdings, Inc., 2Taisei Corporation, 3Toyo Union Co., Limited

1

Abstract
In this paper, the development of a method for decontaminating theinside of flange tank contaminated with radioactive materials at Fukushima
Daiichi Nuclear Power Station (1F) by laser irradiation is described.
1. Introduction
In dismantling the flange tank(Figure-1), radioactive material is fixedby applying painting on the inner surface in advance. However, we must
remove it again during decontaminating dismantled piece of flange tank. In
order to reduce dust scattering during dismantling, reduce the exposure ofworkers and improve the efficiency of work, we developed "Laser

Figure-1 Dismantling flange tankstatus

Decontamination Method (Figure-2)" controlled automatically that canremove radioactive materials and anticorrosion coatings before dismantling.
2. Laser decontamination method for dismantling flange tank2-1. Cold tests (with uncontaminated test piece)
We carried out cold tests to confirm the effective coating removal
method with continuous wave fiber laser (2kw) used for laser material

processing machines recently. We tried and selected the combination of Figure-2 Laser decontamination image
parameters (laser power, speed and frequency), with the test piece whichcoated tar epoxy resin on steel plate (SS400) same as a real tank. (Figure-3)

2-2. Hot tests (with contaminated test piece)
We performed the parameter examination like clean test to confirm thedecontamination effect by laser with the test piece cut out from thedismantled tank. Because of difference of the properties by the influence ofthe encrustations, we reset most suitable parameter. As a result of hot tests,

Figure-3 Cold tests result

we obtained decontamination factor (DF) 10~1000. Then we developed
exclusive laser irradiation and collecting waste system (Figure-4) to reduceradiation exposure. (patent application No. 2018-139336).
3. Conclusion
According to both tests, it was confirmed laser decontamination
method for inside of flange tank was effective. Utilizing these results, we
continuously execute this method to decontaminate flange tank effectively.

27

Figure-4 Laser decontamination
method in flange tank

E03

Uranium-Containing Particles Found in the Fukushima Daiichi
Primary Containment Vessel Interior
Masato Mizokami1,4, Kenichi Ito1,4, Akihiro Suzuki2, Ikken Sato3,4,
Masaki Kurata3,4, Mutsumi Hirai1,4, Shinya Mizokami1,4
1

TEPCO HD, 2NFD, 3JAEA, 4IRID

Abstract
With the recent progression of decommissioning work, radioactive contaminate samples have become obtainablefrom the Primary Containment Vessel (PCV) interior of Fukushima Daiichi accident reactors. The samples weretransported from Fukushima Daiichi (1F) to the hot laboratories of Japan Atomic Energy Agency (JAEA) andNippon Nuclear Fuel Development Co. Ltd (NFD) for general and detailed examinations. This presentation places
the focus on characterization of uranium-containing particles found in the samples.
Introduction
A small portion of the samples were used for Scanning Electron Microscope (SEM) observation. EDS (Energy
Dispersive X-ray Spectroscopy) analysis revealed several spots of high uranium EDS signal, possibly arising by
the presence of uranium-containing particles. We selected some of the spots for transmission electron microscope(TEM) examinations. About 20 spots were examined so far to characterize the particle details.
Characterization of uranium-containing particles
The microstructures and compositions of particles were characterized
by TEM/EDS, and generation mechanisms were discussed.
Type I: Particles generated probably from molten corium
Typical particles consist of dense crystalline Urania-zirconia. An
example of TEM images is shown in Fig.1. The compositions were
examined by EDS at several spots within the particles. The Zr/U ratios
obtained were relatively uniform within the particles, while in one

Fig. 1

particle Zr-rich region was detected in the matrix, probably showing
phase separation of α-Zr(O). The Type I particles have a possibility tocarry compositional features representing main fuel debris.
Type II: Vapor condensation
Some particles have round surface and low Zr content. An example isshown in Fig.2. This looks like an agglomerate of 100nm size particles.
We also found a dense spherical particle with a diameter similar to that
Fig. 2of the agglomerate of Fig.2.
Some other particles look different from the above types (Fig.3),
having a separated regions, gaps and micro cracks.
Conclusion
We detected the two types of particles: One is generation from molten
corium (Type I), and the other is from vapor condensation (Type II).
ACKNOWLEDGEMENT
This presentation is based on the results of Subsidy Program “Project ofDecommissioning and Contaminated Water Management”.

28

Fig. 3

E04

Units 1-2 Exhaust Stack: Local Company's Big
Challenge -- Yes, it's possible!
1・2号機排気筒：地元企業のビッグチャレンジ ～ Yes, it's possible!
Yukihide Sato1, Isamu Okai1 and Mariko Chuman1
ABLE Co. Ltd.

1

Abstract
The high dose rate around Fukushima Daiichi NPP Units 1-2 Exhaust Stack has been an
impediment to the D&D work. And besides, damage and cracks were detected in its support.
ABLE is challenging these difficult tasks as the prime contractor, and demonstrating its hightechnical capabilities.
1. Removal of Accumulated Radioactive Water from Stack Drain Sump Pit
 No as-build drawing, very high dose rate (max. 3.6 Sv/h), narrow space Great efforts in robot development of only 5 months, including mock-up test and training Field work completed in only 2 months resulting from the perfect preparation in off-site

2. Dismantling of the Upper Half of Stack 9 to 12 mmT steel stack, highly contaminated inside, dismantling from 120 m to 59 mH
 Never experienced method: remotely-operated cutting machine suspended from crane FS from 2017, robot development, and elaborate mock-up test & training of 7 months

Cutting of Diagonal Braces

Cutting of Principal Posts

Mock-up test & Training

Cutting of Stack Shaft

Stand-by at the Site

29

E05

Preparation Work for Unit-2 Fuel Removal
at Fukushima Daiichi Nuclear Station
Takashi Inoue, Motohiro Taniyama, Jyunro Nakagoshi,
Yasushi Kageyama, Miho Miyazaki, Ippei Matsuo, Kihei Ogawa
Kajima Corporation

Abstract
Kajima has started preparation work for Unit-2 fuel removal in 2015. Until now, west platform construction (1),new opening work to access the operating floor of the reactor building from the outside (2) and prevention ofincrease in contaminated water due to rainwater intrusion to contaminated buildings (3) have been carried out in ahigh dose rate environment. This paper summarizes the challenges in design and construction.
1. West platform construction
 Overhang design: West platform was overhung to secure adequate work space to build new opening to accessthe operating floor since the west area of the reactor building was very limited.

The platform was connectedwith the reactor building through damping device to resist earthquake.
 Modular construction: Modular construction was adopted to reduce radiation exposure of workers duringassembly of west platform.

After each module was assembled at the yard in a low dose rate environment, theassembled modules were installed by large cranes.
 Damping device: Buckling-restrained brace and ball joint were developed for the connection between west
platform and the reactor building. The earthquake resistance was achieved by plastic deformation of steel
tube which was filled with mortar to prevent bucking.
 Remote-controlled device: Remote-controlled devices for rebar position survey and drilling for anchor
placement were developed to reduce radiation exposure of workers during installation of the bucklingrestrained braces.
2. New opening work in a west outer wall
The new opening work in a west outer wall was carried out inside the temporary house for dust scattering
prevention.

After the wall was cut and divided to several blocks by concrete core cutter, the blocks were

removed by remote-controlled machine.
3. Prevention of increase in contaminated water due to rainwater intrusion to contaminated buildings
 Contaminated protection layer for waterproof roof which contaminates rainwater was removed by remotecontrolled machinery.
 Wireless active personal dosimeters (APD) were introduced to reduce further radiation exposure of works.
They enabled remote-monitoring and remote-instruction for work and evacuation.
 High flowing, fiber reinforced, and heavy concrete was developed to reduce manual repair work for damaged
reinforced concrete roof in high dose rate environment and to enhance radiation-shielding capability.

West platform construction

New opening work in a west outer wall

Removal of contaminated protection layer

Note: The remote-controlled technologies for the above project were developed in collaboration with local companies.

30

H01

Variations of Hydrogen and Vapor Concentrations in a Simply
Simulated Radioactive Waste Long-Term Storage
Container with a Newly Developed PAR
Kazuyuki Takase1, Genya Kawasaki1, Kohei Ueta1, Ky Tri Vu2 and Isamu Kudo2
Nagaoka Univ. of Technology, 2ADVAN ENG. Co., Ltd.

1

Abstract
In the decommissioning of nuclear reactors, hydrogen which is the flammable gas is generated in storage
containers by radiolysis of water. Therefore, in order to improve the safety of waste storage containers forhydrogen, it was experimentally confirmed that a newly developed passive autocatalytic recombiner (PAR) is
effective in reducing hydrogen concentration.
1. Introduction
The objective of the present study is to confirm experimentally that a

20 mm

newly developed PAR is effective for reducing the hydrogen concentration in
the waste storage containers.

Figure 1. developed catalyst

2. Evaluation of PAR on Hydrogen Concentration Reduction
A catalyst in which platinum is coated on the surface of alumina sphere
has been developed as a hydrogen recombination catalyst. By changing the

160 mmdiameter of the alumina sphere, it becomes possible to make PARs with any
size. Fig. 1 shows the appearance of the presently developed PAR with a

500 mmdiameter of 20 mm, which was made by ADVAN ENG Co. Ltd.
Figure 2 shows a container used in the experiment which is 400 mm indiameter and 900 mm in height. A hydrogen concentration sensor and a
relative humidity sensor are installed in the container. In the experiment,
hydrogen gas of a constant flow rate was injected into the container from thecentral portion of the bottom plate of the container.
Figure 3 shows experimental results of

Figure 2. A cylindrical containerhydrogen concentration and relative humidity when
the hydrogen flow rate of 50 cc/min was injected

1.5

100

during 200 s. Here, the number of 3, 6 or 9 indicatesthe number of installed catalysts. Hydrogen and80
1oxygen are combined with PAR and vapor occurs.
As a result, the hydrogen concentration is reduced.

60

Reduction of hydrogen concentration depends on

H2 concentration Relative(%)
humidity (%)

0.5

the number of catalysts.

36
9

3. Conclusion
It was clarified experimentally that the0hydrogen concentration in the container can be
reduced by the presently developed catalyst.

0

5000

40

3
6
9

10000

Time (s)20
15000

Figure 3. Measured H2 concentration

31

H02

Research on hydrogen safety technology
utilizing the automotive catalyst
T. Aida1, T. Yamoto1, T. Kita1, H. Ono1, K. Takenaka1, D. Matsumura2,
M. Taniguchi 3, E-A. Reinecke4, H. Tanaka1
1Kwansei Gakuin University, Hyogo, Japan,

2 Japan Atomic Energy Agency, SPring-8, Hyogo, Japan,

3 Daihatsu Motor Co., Ltd., Shiga, Japan, 4 Forschungszentrum Jülich, Germany

Abstract
The passive autocatalytic recombiner (PAR) for hydrogen oxidation
which does not need electric power supply has been developed.
Researches on the effect of geometric parameter of monolithic substrateon the natural convection and the catalytic activity for oxidizinghydrogen were investigated in order to exhibit sufficient performanceeven with mass production catalysts.
1. Introduction
It has been required to transport and store safely high-doseradioactive waste toward decommissioning of Fukusima Daiich
Nuclear Power Station. Generation of hydrogen in the container due toradiolysis of water is serious problem. The objective of this study is to
reduce hydrogen by recombining to water with monolithic automotive
catalyst without external power source.
2. Experimental
The experiments were conducted in REKO-4 reactor in instituteof Jülich, German. The appearance of REKO-4, 5330 L in the volume,is shown in Fig. 1. In steady state test, the natural convection
velocity was calculated from the amount of injected hydrogen since
it was equivalent to that of hydrogen oxidized by the catalyst. In a
dynamic test, hydrogen gas was injected until the concentration
reached 6%, and then the injection was stopped and all valves wereclosed, after which time changes in hydrogen concentration were
measured. These series of tests used a chimney of 300 mm.

Fig.1. REKO-4 reactor.

Fig.2. Geometric parameter.

3. Results & Discussion
3-1. Steady state test
The effect of geometric parameter on the natural convection was
evaluated (Fig.2). The flow rate of natural convection was 27 times
improved by lowering cell density from 900 to 30 cpsi (Fig. 3).

3-2. Dynamic test
By lowering the catalyst cell density to 30 cpsi, the hydrogen
reduction time could be significantly reduced and achieved to lower
hydrogen concentrations (Figs.4 and 5). It has been confirmed thatlowering the cell density of the catalyst provides significant benefits
from both the steady state and the dynamic tests.

Fig.3. Flow rate / cell density.

4. Conclusion
The flow rate was improved by coarse cell density and it can
reduce hydrogen concentration efficiently to safe level. Sufficientperformance was exhibited even with mass production lower limit
size catalysts (30 cpsi, 10 mm).
Fig.4. The catalytic activity (30 cpsi).
References[1] H. Ono, et.al., Research on hydrogen safety technology utilizing theautomotive catalyst, E-Journal of Advanced Maintenance, Vol.11, No.1
(2019) 40-45.

Acknowledgement
A part of this study is the result of “R&D on technology for reducing
concentration of flammable gases generated in long-term waste storage
containers” carried out under the “Center of World Intelligence Project for
Nuclear Science and Technology and Human Resource Development” by the
Ministry of Education, Culture, Sports, Science and Technology of Japan. Fig.5. The catalytic activity (900 cpsi).

32

H03

Research on hydrogen recombination catalyst
under humidity environment
T. Yamamoto1, T. Aida1, K. Fujita1 T. Kita1, H. Ono1, D. Matsumura2,
M. Taniguchi3, H. Tanaka1

1

Kwansei Gakuin University, Hyogo, Japan, 2Japan Atomic Energy Agency, SPring-8, Hyogo, Japan,
3
Daihatsu Motor Co, Ltd., Shiga, Japan

Abstract
The catalysts have been developed for transport and storage debris from decommissioning of Fukushima Daiichi
Nuclear Power Station. In particular, catalysts that perform well in humid environments are focusing in this study.
It was confirmed that the catalyst performance is maintained even in a high humidity environment by subjecting
the catalyst to water repellent treatment.
1. Introduction
The technology for transport and storage a high-dose radioactive waste is required for decommissioning ofFukushima Daiichi Nuclear Power Station. It is a serious problem that hydrogen generated by radiolysis of waterin the container. The hydrogen oxidation activity was influenced under 100% relative humidity. As acountermeasure for this, the catalyst was treated to be water repellent and the effect was evaluated.
2.Experimental
The catalytic activities were evaluated by using TPR(Temperature Programed Reactor).
‣Catalyst:(1) Pd/Al2O3(2) Pd/WR (Water Repellency)(3) Pd/SWR (Super Water Repellency)‣PGM loading:
0.5 wt% Pd/Supporting materials‣Pretreatment gas: Ox ;10% O2 / Balance He / 200 ℃
Red;10% H2 / Balance He / 200 ℃‣Analysis condition: -120 ℃ ~ RT / without heater
RT ~1 00 ℃ / use heater
DRY / WET (absolute10%)‣Gas composition: 6% H2, 4% O2, Balance, He(50 ml/min)‣Gas analyzer:
Q-massFig.1. Home-in equipment.

3. Results & Discussion
Fig.2 shows that H2 conversion at 20 ℃. All catalysts converted 90% or more of hydrogen under DRY conditions.
The oxidation treatment did not affect the conversion at room temperature. The performance of Pd/Al2O3 wassignificantly reduced under WET conditions. It is considered that the water generated by the reaction covered
catalytic active site. On the other hand, Pd/WR and Pd/SWR were effective. Pd/SWR converted 10% morehydrogen than Pd/WR under WET conditions. Stronger water repellency represented greater catalytic performance
improvement. The hypothesis of the water repellant effect is shown in Fig 3.
H2

O2

H2

O2

Fig.2. H2 conversion at 20 ℃.

Fig.3. Hypnosis of water repellency.

4. Conclusionü Water repellency improved the catalytic activity.
ü The research should be continued aiming at the practical use of hydrogen recombination catalyst.
Reference[1] T. Kita, et. al., “Hydrogen safety technology utilizing the automotive catalyst”, poster H03, The 3rdInternational Forum on the Decommissioning of the Fukushima Daiichi Nuclear Power Station (2018).
[2] H. Ono, et.al., Research on hydrogen safety technology utilizing the automotive catalyst, E-Journal ofAdvanced Maintenance, Vol.11, No.1 (2019) 40-45.

33

K01

Waste Management Symposia: Annual Conference in Phoenix,
Arizona, Exchanging Knowledge from Around the World
Kazuhiro Suzuki1, Gary Benda2 and Akira Ohno3
WM Symposia, Inc. Board Member, 2WM Symposia, Inc. Deputy Managing
Director and 3WM Symposia, Inc. PAC Member

1

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

Introduction

The WM2020 Conference, the 46th version, will be held March 8-12, 2020
at the Phoenix Convention Center in Phoenix, Arizona. Conference theme
is “Reducing Long Term Environmental Liability Through Efficient,
Effective Clean-up”. WM2020 will feature over 500 papers and more than
40 panel discussions in over 140 technical sessions, complemented bynearly 200 exhibiting companies, the industry’s largest.
2.

Poster
The poster provides Conference details and describes
the Technical Panel, Poster and Oral Sessions, Exhibitor,
Student and Sponsorship program, as well as theopportunity to network with over 2,000 industry
specialists and managers from more than 35 countries
and learn of trends and developments from the mostsenior industry managers around the world. WM2020
has special programs aimed at encouraging participationof the world’s leading companies in the exhibit hall,
including a program for product demonstration.
The Conference focuses also on promoting the next
generation of students and radwaste management
professionals by providing a platform for exchanges ofexpertise.

3.

Conclusion

The Conference promotes, among Japanese and professionals
from around the world, a broad exchange of knowledge ontechnologies, operations, safety, waste management issues
and the decommissioning and dismantling. The deadline forsubmittal of Abstracts for WM2020 is August 23. Details areshown on the poster.
Reference: www.wmsym.org.

34

K02

Influence of Na and Ca Concentrations on Sr Adsorption Performanceof Cs-Sr Simultaneous Adsorbent
Akira Tsutsumiguchi1, Yuko Kani1, Tsuyoshi Ito1, Takashi Asano2,
Takatoshi Hijikata3, Tadafumi Koyama3 and Yasuhiro Suzuki4
Hitachi, Ltd., 2Hitachi-GE Nuclear Energy, Ltd.,

1

Central Research Institute of Electric Power Industry, 4JGC Corporation

3

Abstract
Treating evaporation-concentrated liquid waste (evap. waste) stored in tanks at Fukushima with high radioactive
Cs and Sr concentrations has been examined using Cs-Sr simultaneous adsorbent. As a result of the dilution effectevaluation, the Sr breakthrough point of the adsorbent was markedly increased by diluting the evap. waste 4 times
or more.
1. Introduction
In this study, we investigated the applicability to treatment of evap. waste stored at Fukushima which is present
in a small amount but has high Na (max 50000 ppm) and Ca (max 300 ppm) concentrations using Cs-Sr
simultaneous adsorbent [1]. The Na and Ca concentrations are higher than the Sr concentration, and even if theadsorbent has high Sr selectivity, the lifetime is shortened by the influence of Na and Ca. Therefore, we
experimentally evaluated the influence of Na and Ca concentrations on Sr adsorption performance of the adsorbent
to apply the Cs-Sr simultaneous adsorbent efficiently for the evap. waste.
Sr = 2 ppm
const.

A (Na 6500 ppm)
B (Na 13000 ppm)
C (Na 26000 ppm)
D (Na 46500 ppm)

15

20

30

15

20
25 30
30
20
Time (days)
Time(days)A column test was conducted to evaluate the Sr adsorption
performance of Cs-Sr simultaneous adsorbent. Na and Ca
concentrations of the test liquid were set as parameters based on
the evap. waste. From the test results, lowering of Na

C / C0 (Sr)0

C/C0 (Sr) C / C (Sr)

2. Results and Discussion

11
0.8
0.8
0.6
0.6
0.4
0.4
0.2
0.2
00
11 0 5 10
0.8
0.8
Break0.6
through
0.6
point
0.4
0.4
0.2
0.2
00
10
00 5 10concentration was effective from the start even at highwhen decreased to 80 ppm or less (Figure 1). As a result of

evaluating dilution effect of concentration reduction, Sr

breakthrough point of the adsorbent was markedly increased bydiluting the evap. waste 4 times or more (Figure 2).
3. Conclusion
The applicability of the Cs-Sr simultaneous adsorbent to

35

40

45

50

35

40
40

45

50
50

a (Ca 20 ppm)b (Ca 80 ppm)c (Ca 300 ppm)

Figure 1. Sr breakthrough curve
Markedly increased

6060
conc.
4040 NaNa and Ca conc.
reduction
reduction effect2020 effect
000
2
4
10
6
8
0
2 Dilution
4 ratio
6 (-) 810
Figure 2. Dilution ratio andDilution ratio(-)
breakthroughpoint correlation

Breakthrough point (a.u.)

Throughput ratio (-)

concentration, and lowering of Ca concentration was effective

25(days)
Sr = 0.4Time
ppm
Na = 550 ppm
const.

treatment of the evaporation-concentrated liquid waste was confirmed from the results of the column test.
Reference[1] Y. Kani et.al, Proc. WM2014 Conference, Phoenix, Arizona, March 2-6, No. 14110 (2014).

35

K03

Study on Post-Accident Waste Management Scenarios
for Fukushima Daiichi Nuclear Power Station
Subcommittee on Radioactive Waste Management
Review Committee on Decommissioning of the Fukushima Daiichi NPS
Satoshi YANAGIHARA and Hiroshi MIYANO
Atomic Energy Society of Japan

AbstractLong-term radioactive waste management strategy is integral to decommissioning and site remediation at
Fukushima Daiichi Nuclear Power Station (1F). Potential waste management approaches were identified andcharacterized in order to support future decision making on the strategy. Four radioactive waste management
scenarios were developed considering alternative interim and final end states of decommissioning and remediation.
They were semi-quantitatively evaluated from waste generation through disposal in terms of activity timelines,
amount of waste, waste management options, and timescale to achieve the end state.
1. Introduction
The strategy for managing radioactive waste and materials at Fukushima Daiichi Nuclear Power Station (1F)
needs to be developed alongside with the decommissioning and site remediation strategies and plans. The objectiveof the subcommittee is to identify the requirements and constraints on decommissioning and remediation of 1F
from the waste management perspective, and provide information on alternatives of the radioactive waste
management. There are multiple approached to manage waste that will be generated from decommissioning andsite remediation of 1F, and optimization of radioactive waste management scenarios is essential. This study aims
to semi-quantitatively evaluate potential waste management scenarios for multiple potential interim and final endstates of decommissioning and remediation.
2. Methodology
The site at 1F was divided into six zones in terms of decommissioning of nuclear facilities, and three zones interms of site remediation. In order to construct potential decommissioning/site remediation scenarios,
decommissioning/site remediation activities at 1F were divided into four phases: 1) dismantling of reactor andturbine facilities (Phase 1), 2) dismantling of other nuclear facilities (Phase 2), 3) site remediation (Phase 3), and4) nuclear site management/radioactive waste storage and environmental stewardship (Phase 4). Four scenarios
were developed with two possible interim end state of i) complete removal of all facilites/contamination, and ii)
partial removal and continued care and maintenance, and two decommissioning strategy of i) immediate
dismantling, and ii) deferred dismantling.
3. Results and Discussion
The outline of the
scenarios are as follows;
S-1: Immediate dismantling/
complete removal of allradioactive contamination,
S-2: Immediate dismantling/
partial removal of facilities/
contamination followed bycontinued
maintenance/environmental
stewardship, S-3: DeferredFig. 1 Timeline of various scenarios divided into four phases
dismantling/ complete removal
after safe enclosure of nuclear facilities, S-4: Deferred dismantling/ partial removal after safe enclosure followedby maintenance/ environmental stewardship (Fig 1). The S-1 can achieve the end state at the early stage, butdisposal of a large quantity of radioactive waste also needs to be realized. Only a limited area can be released fromthe regulatory control in the S-2 because of the continued C&N and environmental stewardship. The achievementof the end state will be further delayed in S-3 and S-4 when all of the activities occur after safe enclosure period.
4. Concluding Remarks
Based on the Mid-and-long-term roadmap towards the Decommissioning of 1F, fuel debris removal will startin 2021 to complete decommissioning activities by around 2050. These activities will generate waste, andradioactive waste management strategies need to be developed in time. The development involves not only
technical considerations, but discussions with stakeholders on important matters including the end state conditionsof the 1F site and allocating waste storage and disposal facilities. This scenario analysis on the waste management
alternatives aims to support future discussion and decision making from the long-term perspective.

36

K04

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 times as much wastes
as Cement matrix equivalent on average.
Theequipment

used

to

deploySIAL®

solidification technology is also modular, flexible andversatile.

It can encapsulate waste streams quicker

than Cementation, and can be applied under water.
SIAL solidification technology can realize®

Figure 1 SIAL® solidification treatment process

higher compressive strength and lower leachability
compared to

Cementation, and posing a low firerisk and excellent physical stability in the presenceof frost and water (no distortion or cracking).
Figure 2 Sludge and slurry waste streams crosssection observation
3. Performance Record Example
About 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. This comprehensive scope of works started
with licensing processes, solidification, and then were followed by decontamination and cleaning of theworkplace post cleanup and transports of all equipment to off-site.

37

J01

Preparation of VR data for utilization to decommissioning ofFukushima Daiichi Nuclear Power Station
Rintaro Ito1, Takahiro Ohno1, Yoshihiro Tsuchida1, and Takashi Okada1

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

1

Abstract
The Naraha Center for Remote Control Technology Development (Naraha Center) provides anoperation training system for decommissioning work using immersive virtual reality technology (VRsystem). The current status of the preparation of VR data of Fukushima Daiichi Nuclear Power Station (1F)in the system is described below.
1. Introduction
The VR system installed at the Naraha Center can display the equipment and structure of the 1F facility
in full scale with three dimensions and with dose distribution. This system is utilized for planning andtraining of works related to 1F decommissioning. In order to make more effective use of this system, it isrequired that the internal situation is accurately reproduced so that users can plan and train in the sameenvironment as the site. Therefore, continuous preparation works on the 1F VR data have been carried out.
2. Current status of the VR data
The data prepared so far are from thefirst floor to the basement floor in the reactor
buildings of Units 1 to 3. These data arecreated by the point cloud data measured by
the robot which entered in the reactor
buildings after the accident, and the
deficiency data are complemented with thedata before the accident. In the Unit 2, someparts of PCV have been prepared for the
future debris retrieval works. Furthermore, as
for the inside of the pedestal, it was difficult
to obtain the data, then the Unit 5 data wereobtained by point cloud measuring and
Fig1. Status of the Unit 2 data
converted it into 3D CAD data.
In this fiscal year, the preparation of the VR data is planned from 3D-CAD, a part of Unit 2 PCV
acquired last year and small rooms on the first floor of the reactor building. In addition, it is also planned
to collect survey data, which are inside of the reactor building acquired by International Research Institutefor Nuclear Decommissioning, and to expand the data.
The prepared data can be used in the system, and can also be used outside of our center for the 1F
decommissioning work as well as the research and develop on the 1F decommissioning.
3. Conclusion
By utilizing the data in this system, it is believed that this system contributes to the safe and reliable
decommissioning work, and also to the improvement of the decommissioning work. Our future works are
to expand the data based on needs of users, and to develop a mechanism for applying the ever-changing onsite situation.

38

J02

Evaluation Method of Robot Mapping Performance for Decommissioning

Taichi Yamada1 and Kuniaki Kawabata1
1Japan Atomic Energy Agency

Abstract
For decommissioning of Fukushima Daiichi nuclear power plant (1F), workers need to know about worksite information for working efficiency and safety of themselves. Mapping technology in robotics called
Simultaneous localization and mapping (SLAM) is useful for sharing the site information. Though, there aretechnical problems for applying SLAM to 1F tough environments. This paper shows the development of the
SLAM evaluation method aimed to enhance the research of SLAM for 1F.
1. Introduction
For 1F decommissioning, workers can work only for a very short time due to high radiation exposure.
Therefore, workers need to transfer the task to next workers, and the information sharing is important. The
visualization of site data makes it easy to share the information. SLAM allows to visualize data of the site bycreating maps based on robot sensor data. However, there are technical problems to apply SLAM to 1F
environments because of high radiation, darkness, drops of water and so on. Furthermore, because theexperiment on the real 1F site is impossible, research to apply SLAM to 1F is difficult. Therefore, in order toenhance SLAM research for 1F decommissioning, we aim to develop the evaluation method and test data for
SLAM algorithms. This paper represents the idea to facilitate SLAM algorithm test without actual field and thequantitative evaluation of SLAM algorithm.
2. SLAM Evaluation Method
SLAM algorithms compute the robot position based on the robot

Grating Corridor Mockup

Camera Image in PCV

sensor data such as camera images. Therefore, by using the log ofrobot sensor data, it is possible to test SLAM algorithm without any
actual equipment, robot and testing field. Moreover, the performancecomparison between SLAM algorithms can be conducted based on
testing results using the same log, that means, under the same

Copyright: TEPCO

Camera Image in Dark

condition. Thus, we prepare the log of sensor data when a robot moves
on the mockup field simulating 1F site in Naraha Center for Remote
Control Technology Development [1] for SLAM algorithm test.
Regarding the SLAM evaluation, we can evaluate quantitatively
SLAM algorithm by comparing the estimated robot position by the

Figure 1. Mockup Filed of PCVSLAM algorithm and the accurate robot position. In this research, weobtain the accurate robot position using motion capture at the same time of the sensor data logging. Specifically,
our current research is focused on the collecting the log of robot sensor data and the accurate robot position databy motion capture when the robot moves on the mockup field of platform in PCV like shown in Figure 1.
3. Conclusion
This paper describes a SLAM evaluation method using the sensor data log when the robot moves on
mockup field. In the future works, we will verify this SLAM evaluation method with the sensor data log.
References[1] Naraha Center for Remote Control Technology Development, URL: https://naraha.jaea.go.jp/en/index.html

39

J03

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

Japan Atomic Energy Agency, 2Osaka Prefecture Univ., 3National Institutes for Quantum and1

Radiological 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
are developed based on corrosion mechanism.
Preventing methods for these corrosion risks are

Fig. 1 Implementation systemdeveloped based on corrosion mechanism. The implementation
system of this research is shown in Fig.1.
2. Research plan and results
2-1. Data area for evaluation
In 1F decommissioning process, corrosion data is needed inrelatively high dose rate (100 - 10,000 Gy/h) and 1 μS/cm - 10mS/cm in conductivity conditions (Fig. 2). The gamma-ray
irradiation test environment was constructed in the gamma-ray
irradiation facility, Takasaki advanced radiation research institute,
QST as a research hub for the corrosion under irradiation.

Fig. 2 Data area for evaluation of 1F corrosion

2-2. Major results
The obtained major results are listed in follows;●The corrosion test environment under irradiation with
controlling pH, temperature dissolved O2 concentration andion concentration was constructed (Fig.3).
●The corrosion behavior changes with [B4O72-]. The
de-passivation occurred with the increase of the [Cl-] in

Fig. 3 Gamma-ray irradiation test system

B4O7 and Cl co-existent condition.
2--●Corrosion rate at the waterline for half immersion was larger than simple immersion. Corrosion rates underirradiation were larger than non-irradiation.
This work was conducted as a part of "Analysis of Corrosion Mechanism in Specific Environment" funded by the Ministryof Education, Culture, Sports, Science and Technology (MEXT) of Japan.

40

J04

Development of a Robot Simulator System
for Operator Proficiency Training
Kenta Suzuki, Shunji Mukaida, Fumiaki Abe and Kuniaki Kawabata
Japan Atomic Energy Agency

Abstract
This paper describes the current development status of a robot simulator system for nuclear decommissioning
operations by using remotely operated robots. This system is utilized for operator proficiency training andimplemented function of underwater behavior, communication traffic control, image view presentation, etc. arefor simulating the situations and conditions of decommissioning tasks.
1.

Introduction
After the accident at Fukushima Daiichi Nuclear Power Station (FDNPS), remotely operated robots were
deployed to execute the tasks such an environmental and a radiation survey [1]. However, maneuvering the robots
was sometimes difficult and also doing the training within FDNPS by using remotely operated robots is difficult
because it may worsen the condition of the work site. Therefore, we are developing a robot simulator system forcontributing an operator proficiency training.
2.

Implemented functions and Current status of our study
In actual decommissioning tasks, remotely operated robots were deployed to execute environmental survey
inside the reactor building, dose rate measurement, underwater investigations in the cooling water and so on. The
communication condition also affects to maneuver the robots by the operator. A gamma camera carried by the
robots was utilized for radiation measurement at remote site. By referring to the conditions and the lessons learned
from past tasks, we have started developing a robot simulator system based on Choreonoid and developed extended
functions for simulating above-mentioned situations. Figure 1 shows the implemented functions on the system andeach function is a plug-in function of Choreonoid. FluidDynamicsPlugin is to calculate the buoyancy and the fluid
resistance applied to the robot model [2]. TCPlugin is to emulate communication conditions like time delay,
bandwidth limitation and packet loss. ImageNoisePlugin is to provide a magnification, the distortion, the gaussian
noise and the salt-pepper noise on captured image from the camera device. DosePlugin is to provide simulated
view of a pinhole-type gamma
camera utilizing on radiation dose
rate data. In addition, digitalmodels of 1st and basement floor of
the rector building of FDNPS, unit
1, 2 and 3. Also the model of mockup testing facilities of Naraha
Center for Remote Control
Technology Development, Japan
Atomic Energy Agency are
installed. By combining thesefunctions, the users can set various
situations and conditions and do
operator proficiency training.
Figure.1 Implemented functions on robot simulator
3.

Conclusion
We described the current status of development of the robot simulator and we introduced implemented
functions and the digital work site models. The users can train to improve an operator’s skill for robot operation.
In future work, we will attempt to develop the functions for simulating a decommissioning task by using
multiple robots and also a surface contamination of the robot body.
References[1] K. Nagatani, S. Kiribayashi, Y. Otake, K. Yoshida, S. Tadokoro, T. Nishimura, T. Yoshida, E. Koyanagi, M. Fukushima
and S. Kawatsuma, “Energency response to the nuclear accident at the Fukushima Daiichi nuclear power plants using
mobile rescue robots”, Journal of Field Robotics 30 (1), 44-63, DOI: 10.1002/rob.21439, (2013)[2] K. Suzuki, K. Kawabata “Development of a multi-copter simulator and a projection system for virtual operation
experience”, 2019 IEEE/SICE International Symposium on System Integrations (SII 2019), Paris, France, 2019

41

J05

Outline and Construction status of
Okuma Analysis and Research Center, Laboratory-1
Takahiro Furuse1, Yuki Ota1, Van-Khoai Do1
1
Japan Atomic Energy Agency (JAEA)

Abstract
The Japan Atomic Energy Agency has been developing Okuma Analysis and Research Center, a research anddevelopment facility, in order to ascertain the properties of radioactive wastes and fuel debris towards thedecommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Station (1F). In this presentation, we arereporting outline and construction status of Laboratory-1 which will analyse low and medium level samples in
the Okuma Analysis and Research Center.
Outline of Laboratory-1
Radioactive materials with surface dose rate up to 1 Sv/h such as rubble, incinerated ash, and secondary
waste from water treatment will be analyzed. Laboratory-1 equips iron cells, glove boxes and fume hoods(Fig. 1).

Fig. 1: Conceptional drawing of Laboratory-1


Table 1: Main apparatus to be installed

Radiochemical analysis of 38 nuclides isanticipated necessary to establish technicalprospects on safe treatment and disposal ofradioactive waste from 1F. Facility design
was made to enable variety of analyses,
which are listed in Table 1.

Construction status of Laboratory -1
Construction work began on April 19, 2017, we have been adopting precast concrete method. This method
is that construction products are made in factories in advance to reduce work time in construction site.

The building is under construction, we are making the third floor now (Fig. 2).

October, 2018

February, 2019
Fig. 2: Construction status of Laboratory-1

42June, 2019

J06

Development of analytical methods for 1F-samples~Analysis of 93Zr by ICP-QQQ-MS and automated sample
preparation instruments~
Van-Khoai Do1,2, Takahiro Furuse1,2, Yuki Ota1
Japan Atomic Energy Agency (JAEA)International Research Institute of Nuclear Decommissioning (IRID)
1

2

Abstract
This presentation outlines our idea of method development for the analysis of 93Zr with application of ICPQQQ-MS after a simple chemical separation using automated sample preparation instruments

Introduction
The analysis of radioactive samples collected from Fukushima Daiichi Nuclear Power Station (herein referred
as 1F) is in a great demand in order to promote the decommissioning of the nuclear facilities. In general, theanalysis of those samples often includes a chemical separation step, which eliminate completely influence ofsample matrix, and a quantification step using an appropriate technique. 93Zr is one of long-lived fission productnuclides, which emits low beta energy Sample from
dissolution
particles. 93Zr is, therefore, a difficult-tomeasure isotope. Conventional methods
91Zr carrier
require a time-consuming chemical separation
oxalic acidof it from all other beta emitting nuclides. In
ZR resin
the previous report [1], it was found thatinductively coupled plasma triple quadrupole
mass spectrometry (ICP-QQQ-MS) is a
SPE mini workstation TBS-3
promising method for 93Zr quantification• Automated control by a computer
Evaporation
because it can simplify the chemical separation• 4 separated lines for 4 parallel columns• Precise flow rate
step and be capable of enhancing detection
Dilution• Precise injected amount of reagents
limit.
Fig. 1 Separation scheme and automated SPE instrument

ICP-MS

Sample preparation
instruments

using

automated

NH3m/Z ≠ 93

Figure 1 shows a simplified separation
scheme for Zr (left) using an automated SPE
instrument (right). Samples (rubble, ash, soil) are
first dissolved in acids using a microwave. Then,
Zr is separated from the sample matrix by a solid
phase extraction (SPE) process using ZR-resin(Triskem) before measured by ICP-QQQ-MS.

m/Z ≠ 195

m/Z = 195

Q1
(m/Z = 93)
93Zr ：

NH3：

Reaction cell
93Mo ：

other ：

93Nb ：

m/Z ≠ 195

Q2
（m/Z = 195)

Fig. 2 Interference removal using ICP-QQQ-MS

Quantification of 93Zr using ICP-QQQ-MS.

Interference from natural 93Nb and activation product 93Mo is a major problem in quantification of 93Zr. Theinterference is expected to be removed partly in the previous chemical separation. Additionally, our preliminary
data indicate that ICP-QQQ-MS with two mass filters (Q1 and Q2) and a reaction cell can remove effectively Nb
using NH3 as a reaction gas as illustrated in Fig. 2 [1]. According to our test with non-radioactive isotopes, at mass
shift of 102, instrumental sensitivity for 90Zr is 5 orders of magnitude greater than that for 93Nb and 98Mo. After
separated from the sample matrix and interferences, collected 93Zr is quantified through 91Zr standard after a mass
bias correction.
Reference [1] T. Horita et at, AESJ Spring meeting 2019, 2D10.
This report includes part of the results obtained from the work “Research and Development of Processing and Disposal of
Solid Waste”, carried out on a supplementary budget allocated for Project of Decommissioning and Contaminated Water
Management.
43

J07

Approach to analysis of hydrogen combustion inside of radioactive
waste vessels using CFD softwares based on an open source“OpenFOAM”
Thwe Thwe Aung1, Atsuhiko Terada1 and Ryutaro Hino1
1

Japan Atomic Energy Agency

Abstract
As an approach to analysis of hydrogen (H2) combustion inside of radioactive waste vessels, we performed
the simulations for propagation of H2-air premixed flame by XiFoam solver of OpenFoam package. A new laminarflame speed model deduced from H2-air explosion experiments by T. Katsumi et al. [1] was implemented in thesolver. Flame radius obtained by simulation agreed with the experimental results [1-2].
1. Introduction
Since hydrogen is continuously generated and releases inside of high-level radioactive waste vessels, theawareness must be taken on the risk of hydrogen combustion and explosion. In premixed combustion of H 2-air,flame front becomes unstable and wrinkle due mainly to the hydrodynamic and diffusive-thermal effects, whichcauses the increase in flame surface area and the faster in flame propagation speed, and the flame accelerates itself.
In severe cases, fire and gas explosion occur. In hydrogen safety management, besides the experimental
investigations, CFD approaches in predictions of flame propagation phenomena are of important role. Therefore,
we introduced the new laminar flame speed model in XiFoam solver [3] and reproduced the propagation of H2-air
flame [1].
2. Simulation Setup and Results
By using the modified XiFoam solver, propagation of H2-air flame was simulated with setting the parameterobtained by experiment [1]. Mesh for one-eighth of model (0.13m in each side) with uniform mesh size of 0.65mm
was generated. Total number of cells was 8,000,000. Equivalence ratio of H2-air premixture was set to unity, andinitial temperature and pressure was set to 298K and 101,325Pa, respectively. Figures 1 and 2 show the temperature
distribution of flame at t = 0.007s, and the comparison of flame radius obtained by XiFoam and experiments,
respectively. Wrinkle flame formation was observed as in Fig. 1 when the flame propagated outwardly as in
1/8 of model

full model

Flame Radius [cm]experiments [1-2].
15

XiFoam_newSu-model

10

T.Katsumi et al.
M.Ilbas et al.

50Figure 1. Temperature distributions of H2-air premixed flame.

XiFoam_default-Su-unstrained
0

0.002

0.004
0.006
time [s]

0.008

0.01

Figure 2. Comparison of flame radius.

3. Conclusion
The results indicated that the modified XiFoam solver of OpenFOAM has capability to reproduce thespherically expanding H2-air flame and propagation phenomena as in experiments within 0.005s.
References[1] T. Katsumi et al., Journal of the Combustion Society of Japan, 2017, Vol. 59 No.189, pp. 210-215 (in Japanese)[2] M. Ilbas et al., International Journal of Hydrogen Energy, 2006, Vol.31, pp. 1768-1779[3] OpenFOAM User Guide v1712, https://openfoam.com/releases/openfoam-v1712/

44

J08

Detection of Alpha Particle Emitters Originating from Nuclear Fuelsof the Fukushima Daiichi Nuclear Power Station
Yuki Morishita1

Japan Atomic Energy Agency/Collaborative Laboratories for Advanced Decommissioning Science

1(CLADS)
Abstract
We measured alpha emitters obtained from a reactor building in the Fukushima Daiichi Nuclear Power Station(FDNPS) by using an alpha particle imaging detector which combines a very thin (0.05-mm-thick) a cerium-doped
Gd3(Ga,Al)5O12 (Ce:GAGG) scintillator and silicon photomultipliers (SiPMs). The alpha spectrum was in theenergy range of 5–6 MeV, which corresponds to the alpha particle energy of 238Pu (5.5 MeV alpha) originated fromthe nuclear fuel.
1. Introduction
Several papers have reported the presence of alpha particle emitters originating from Nuclear Fuels at the FDNPS[1],[2]. A commercial ZnS(Ag) scintillator-based alpha counter cannot show where the alpha emitters are
originating from: nuclear fuels or naturally occurring radionuclide such as Radon (Rn) and progeny. Therefore, wedeveloped an alpha particle imaging detector which is capable of measuring an alpha particle spectrum and 2dimensional distribution. Some research results were recently published in the journal of Scientific Reports [3].
2. Materials and Methods
Actual measurements for smear samples were performed at the FDNPS site. The detector was composed of a0.05-mm-thick cerium-doped Ce:GAGG scintillator and SiPM arrays. The floor of the reactor building in FDNPS
was wiped off by using smear papers, and the radioactivity of these papers was measured by the alpha particle
imaging detector. In addition, we measured a Plutonium (Pu) sample
obtained from a nuclear fuel facility by using the same detector forcomparison with the smear papers.
3. Results
Figure 1 shows the comparison of alpha spectra between the smear
paper and the Pu sample. The alpha spectrum was in the energy range
of 5–6 MeV, which corresponds to the alpha particle energy of 238Pu.
Moreover, the 60 keV gamma-ray peak from 241Am was identified using

Fig.1 Comparison of alpha spectra

a CZT-based gamma spectrometer (Fig.2).

between the smear paper and the

4. Conclusions

Pu sample

We measured the contamination of the alpha emitters of the FDNPS smear
papers by using the alpha particle imaging detector. Based on the measured

241Am peak

results, we report actual findings of alpha emitters in the FDNPS reactor
buildings originating from nuclear fuels.
References[1] K. Shozugawa, et al. Environmental pollution 163 (2012) pp. 243-247.
[2]S. Schneider et al. Scientific reports 3 (2013): 2988.

Fig.2 Gamma-ray spectrum[3] Y. Morishita et al. Scientific reports 9.1 (2019), 581.

of smear paper

45

J09

On the hydrogen production of geopolymers undergamma irradiation
Vincent Cantarel1, Makoto Arisaka1, Isao Yamagishi1
1Japan Atomic Energy Agency

Abstract
We measured the hydrogen production of geopolymers under gamma irradiation and confront
the experimental result to a simple model. Using this model, we were able to highlight theimportance of hydrogen recombination [1].

1. Introduction

To ensure the absence of hydrogen explosion risk, it is important to quantify the amount of H 2
gas produced by radiolysis. This is a key factor for intermediate activity wastes in hydraulicbinders. To enhance the accuracy of hydrogen production prediction, experiments were carriedout to quantify the effects of gas diffusion throughout geopolymers, interesting matrix for in the
nuclear waste management field [2].

2. Experimental Results – Irradiation of geopolymers

Under 60Co irradiation, the hydrogen production yield (mol of H 2/Joule of energy absorbed) was
measured for water saturated geopolymer of different size to increase the diffusion length for thegas (Figure 1). From powder to 40 cm high cylinders a strong decrease of the hydrogen yield wasmeasured, even after the hydrogen trapped inside the sample was allowed to diffuse out of thematrix.

3. Modelling of the hydrogen gas in geopolymers

For the modeling of the hydrogen production, 3 phenomena were taken in account, production,
recombination and diffusion of H2. With f a function representing the hydrogen concentration, it
was found that approximation of the solution of equation 1 allows to predict the hydrogen of all
samples with a reaction constant of 6.10-4
s-1, all other parameters fixed on literature
values.

߲ ଶ݂߲݂ൌ ܲ െ ݂݇ ൅ ‫ ܦ‬ଶ ሺͳሻ
߲‫ݐ‬
߲‫ݔ‬

4. Conclusions

When the sample is large, the H2
accumulates in the sample and is available
for recombination, making the global yielda lot lower than powdered samples.
References[1] Cantarel, V.; Arisaka, M.; Yamagishi, I. On thehydrogen production of geopolymer wasteforms underirradiation J. Am. Ceram. Soc. 2019 00:1–11.
https ://doi.org/10.1111/jace.16642[2] Cantarel, V.; Motooka, T.; Yamagishi, I.
Geopolymers and their potential applications in the
nuclear waste management field; A Bibliographicalstudy- JAEA-Review 2017-014

Figure 1 : Hydrogen production of geopolymers
samples of different sizes and model predictions

46

J10

Development of an Adaptive Control System for Laser Cutting of Fuel
Debris and Structural Components for the Decommissioning of the
Fukushima Daiichi Nuclear Power Plant
Koichi Saruta, Naomitsu Kamei, Yuji Sato, and Toshiharu Muramatsu

Applied Laser and Innovative Technology Institute, Japan Atomic Energy Agency (JAEA)
Abstract
An adaptive control system for laser cutting of fuel debris and structural components is studied for thedecommissioning of the Fukushima Daiichi Nuclear Power Plant.

An overview of the research and developmentof the adaptive control system and some preliminary experimental results are presented.
1. Introduction
Laser cutting is one of the promising cutting techniques in the decommissioning of nuclear power plants
since it provides many advantages such as remote controllability, high efficiency, and good compatibility withrobot technologies.

To make the best use of these capabilities and successfully apply laser cutting to fuel debris

and structural components in damaged nuclear power plants such as the Fukushima Daiichi Nuclear Power Plant,
an adaptive control system will be required because of the inherent difficulty and insufficient knowledge of cutting
those targets.
2. Overview of adaptive control system
As shown in Fig. 1, the proposed adaptive control system [1] employs
various environmental sensors such as a laser scanner to capture indefinite
shapes, a laser-induced breakdown spectroscopy sensor to identify multiingredient materials, and a two-color thermometer to monitor melting
conditions.

Based on the information from the sensors, the robot motion,laser irradiation, and assist gas flow are adaptively controlled to perform laser
cutting processes under various conditions.

To demonstrate the adaptive

control performance, a series of cutting experiments were conducted using atest piece simulating fuel debris, which consists of a SUS304 plate and alumina Figure 1.
system.
pellets, as shown in Fig. 2 [2]. The results confirmed that the adaptive control

Adaptive control

system enables laser cutting of the simulated fuel debris in both air and water.
3. Summary
The preliminary results obtained from the experiments are very
encouraging in the sense that the proposed adaptive control system is usefulfor laser cutting of the fuel debris and structural components at the Fukushima
Daiichi Nuclear Power Plant.

In the next step the adaptive control system Figure 2. Demonstration of thewill be integrated into the robot arm system installed in the Fukui Smart adaptive control system for laser
cutting of the simulated fuelDecommissioning Technology Demonstration Base [3] of JAEA to debris.
demonstrate laser cutting of simulated fuel debris and structural components
in full-scale experiments.
References[1] Muramatsu et al., JAEA-Research 2013-024 (in Japanese). [2] Muramatsu et al., JAEA-Research 2014-018 (in Japanese).
[3] https://fsd.jaea.go.jp/

47

J11

Prediction of the dose rate of various fuel debris to be sampled from
Fukushima Daiichi Nuclear Power Station
Kenichi Terashima1,2, Keisuke Okumura1,2
1International Research Institute for Nuclear Decommissioning (IRID),

Japan Atomic Energy Agency (JAEA)

2

Abstract
The dose rate around the surface of retrieved fuel debris is important information for radiation shielding,handling, transport, analysis and storage of the fuel debris. Therefore, we have developed the dose rate prediction
method applicable to various fuel debris sampled from the Fukushima Daiichi Nuclear Power Station in the near
future. Using the method, we predicted the dose rate of the fuel debris to be sampled in 2021 and clarified thesensitivities of contributing parameters.
1. Introduction
Samplings of a small amount of fuel debris are planned at the Fukushima Daiichi Nuclear Power Station (1F).
However, it is anticipated that various types of fuel debris will be sampled depending on the unit and the sampling
position, which makes it difficult to predict the dose rate of fuel debris required for radiation shielding and workers’
exposure management. Therefore, by combining the fitting formula based on the results of many photon transport
calculations by the Monte Carlo code PHITS with theoretical models, we have developed an efficient prediction
method of the surface dose rate defined as ambient dose equivalent around the surface of fuel debris.
2. Method
It is considered that the surface dose rate depends on fuel debris size, radiation source for each nuclide, releaseratio of volatile FP at the 1F accident, burnup, elapsed time, etc. The combination of these parameters is enormous,and the information to estimate the value of each parameter is quite limited. Therefore, we have developed an
equation for predicting the surface dose rate by the following procedure: (1) Making a radiation source composition
by burnup and activation calculations, (2) Making a photon source (line spectrum) for each radioactive nuclide,(3) Dose rate calculations by PHITS for various combinations of parameters for spherical fuel debris assuming
homogeneous composition with unit photon source for each nuclide. (4) Making a prediction equation based on
the fitting formula obtained from the results of many PHITS calculations, and on the theoretical models onradiation source decay.
3. Conclusion
The obtained equation for the prediction of the surface dose rate D is expressed as follows,�⃗� = � 𝑔𝑔𝑔𝑔𝑖𝑖𝑖𝑖 (1 − 𝑓𝑓𝑓𝑓𝑖𝑖𝑖𝑖 /100) ∙ 𝑃𝑃𝑃𝑃𝑖𝑖𝑖𝑖 �𝑟𝑟𝑟𝑟, 𝜌𝜌𝜌𝜌, 𝑢𝑢𝑢𝑢, 𝑡𝑡𝑡𝑡, 𝐵𝐵𝐵𝐵𝐵 𝐵 𝐵𝐵𝐵𝐵𝑖𝑖𝑖𝑖 (𝑟𝑟𝑟𝑟, 𝑢𝑢𝑢𝑢, 𝑁𝑁𝑁𝑁𝑍𝑍𝑍𝑍 )𝐷𝐷𝐷𝐷 𝐷𝐷𝐷𝐷𝐷, 𝜌𝜌𝜌𝜌, 𝑢𝑢𝑢𝑢, 𝑤𝑤𝑤𝑤𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍 , 𝑤𝑤𝑤𝑤𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹 , 𝑡𝑡𝑡𝑡, 𝑁𝑁𝑁𝑁𝑍𝑍𝑍𝑍 , 𝐵𝐵𝐵𝐵, 𝑓𝑓𝑓𝑓
𝑖𝑖𝑖𝑖

𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹
𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍∙ 𝑃𝑃𝑃𝑃𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶60
�𝑟𝑟𝑟𝑟, 𝜌𝜌𝜌𝜌, 𝑤𝑤𝑤𝑤𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍 , 𝑡𝑡𝑡𝑡𝑡 + 𝑔𝑔𝑔𝑔𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶60∙ 𝑃𝑃𝑃𝑃𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶60
�𝑟𝑟𝑟𝑟, 𝜌𝜌𝜌𝜌, 𝑤𝑤𝑤𝑤𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹 , 𝑡𝑡𝑡𝑡𝑡𝑡 𝑡𝑡𝑡𝑡𝑡𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶60 (𝑟𝑟𝑟𝑟, 𝑢𝑢𝑢𝑢, 𝑁𝑁𝑁𝑁𝑍𝑍𝑍𝑍 )+ �𝑔𝑔𝑔𝑔𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶60
𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍+ 𝑔𝑔𝑔𝑔𝑆𝑆𝑆𝑆𝑆𝑆𝑆𝑆125 ∙ 𝑃𝑃𝑃𝑃𝑆𝑆𝑆𝑆𝑆𝑆𝑆𝑆125 �𝑟𝑟𝑟𝑟, 𝜌𝜌𝜌𝜌, 𝑤𝑤𝑤𝑤𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍 , 𝑡𝑡𝑡𝑡𝑡𝑡𝑡𝑡𝑡𝑡𝑆𝑆𝑆𝑆𝑆𝑆𝑆𝑆125 (𝑟𝑟𝑟𝑟, 𝑢𝑢𝑢𝑢, 𝑁𝑁𝑁𝑁𝑍𝑍𝑍𝑍 ).

where, r: radius, ρ: bulk density, u, 𝑤𝑤𝑤𝑤𝑍𝑍𝑍𝑍𝑍𝑍𝑍𝑍 , 𝑤𝑤𝑤𝑤𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹 : concentrationof uranium, zirconium, iron in fuel debris, respectively,t:time elapse from the accident, Nz: number density of proton,

B: burnup, 𝑓𝑓𝑓𝑓𝑖𝑖𝑖𝑖 : release ratio from fuel debris, i: contribution

radioactive nuclide, 𝑔𝑔𝑔𝑔𝑖𝑖𝑖𝑖 : correction coefficient (usually 1.0),

𝑃𝑃𝑃𝑃𝑖𝑖𝑖𝑖 : Function of photon intensity, 𝑅𝑅𝑅𝑅𝑖𝑖𝑖𝑖 : Response function of
surface dose rate for unit photon source by nuclide.

Figure 1 shows an example of the surface dose rate forthe fuel debris assuming sampled from Unit-2 in 2021.

Figure 1. An example of surface dose rate offuel debris and contributing nuclides

This study includes the result of Subsidy Program “Project of Decommissioning and Contaminated Water Management (Development ofAnalysis and Estimation Technology for Characterization of Fuel Debris)” in the FY2016 supplementary budget.

48

J12

Innovative laser remote analysis for surveillance of nuclear debris
in decommissioning of Fukushima Daiichi Nuclear Power Station
Ikuo Wakaida1, Hironori Ohba1,2, Masabumi Miyabe1, Katsuaki Akaoka1,
Masaki Ohba1, Koji Tamura1,2, Morihisa Saeki1,2, Yuji Ikeda3, Takunori Taira4Japan Atomic Energy Agency/CLADS, 2QST/QuBS, 3Imagineering Inc., 4NINS/IMS

1

Abstract
Radiation resistant optical fiber based laser induced breakdown spectroscopy (Fiber LIBS) and laser
ablation resonance absorption spectroscopy are developing for elemental and isotope analysis of sampled
nuclear debris in the on-site hot cell facility at the Fukushima Daiich Nuclear Power Station. Microwave
assisted LIBS for higher sensitivity and Microchip laser use LIBS (MC-LIBS) for flexible and easy lightdelivery, are also performed for the surveillance.
1. Introduction
For the decommissioning of “Fukushima Daiich Nuclear Power Station (F1NPS)” development of rapid, easy,
onsite and especially in-situ remote analysis/surveillance techniques under the severe environments such as extremely
high radioactive condition will be strongly required. In order to accomplish this requirement, the concept of “Probing
by light and Analyzing by light” with radiation resistant optical fiber will be one of the simple, powerful andapplicable choices as the innovative development based on LIBS and related technologies. In near future, at the firsttime, analysis/surveillance of sampled debris with strong radiation activity will be required before the transportation ofdebris samples to the Hot-facility at the outside of F1NPT to make fine and detailed analysis. And the next, we will
introduce again the planning of in-situ surveillance of debris in damaged core.
2. Schematics of surveillance system
2-1. Basic concept of laser remote analysis
Basic concept of laser remote analysis will beshown in Fig1. LIBS for elemental analysis and laser
ablation resonance absorption spectroscopy in reducedpressure for isotope analysis are integrated as the
non-contact, no-preparation, rapid and in-situ analysis
system. The laser light and the atomic emission will beable to delivery through an optical fiber, so, this concept
will be suitable for remote analysis even under a highradiation activity.
2-1. Surveillance system for the on-site hot cell facility

For fine and detailed analysis of debris in the
Hot-facility at the outside of F1NPT, confirmation of theexistence of the nuclear fuel material in sampled debris by
pre-analysis/surveillance will be required before
transportation. The basic system for the on-site hot cell
facility at F1NPT are now introduced and under developing
as shown in shown in Fig.2. The activity of microwave
assisted LIBS and MC-LIBS are also under developed.

Fig.1 Basic concept of laser remote analysis

Fig.2 LIBS system for on-site Hot Cell
Facility

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).

49

J13

Research and Development on Solidification Technology for
Fukushima Daiichi NPS Accident Waste

Jun Kato1,2, Taniguchi Takumi1,2, Takeshi Osugi1,2, Osamu Nakazawa1,2,
Tomoyuki Sone1,2, Ryoichiro Kuroki1,2
1Japan Atomic Energy Agency (JAEA), 2International Research Institute for
Nuclear Decommissioning (IRID)

Abstract
We extracted task for selecting solidification technologies available to contaminated water treatment
secondary wastes occurred by Fukushima Daiichi NPS accident. The extracted task is lack of information to
select solidification technology available to contaminated water treatment secondary wastes. The information
lacked was obtained by investigation and experiment for solve this task.
1. Task for selecting solidification technology
Among solid wastes caused by Fukushima Daiichi NPS Accident, rubble, the wastes account for the majority
of total (i.e. trees and the waste generated by power plant dismantling (i.e. concrete, metals, and so on)) wasremoved from target examined because these properties are similar to the existing nuclear facility waste except
radioactive inventory. On the other hand, contaminated water treatment secondary wastes (i.e. zeolite, slurry,
resin, and so on) have diversity and they have not processed yet. The purpose of this study is to selectsolidification technology, which will be applicable to the contaminated water treatment secondary wastes. Initial
candidate technologies were suggested based on our technology catalog [1] and they were further limited at thepast approach. The candidate technologies were limited to high temperature processing (i.e. vitrification andmelt-solidification) and low temperature processing (i.e. cementation and Alkali Activated Material (AAM)
solidification). The task on these technologies applied to the contaminated water treatment secondary wastes was
extracted, and it was found that obtainment of the information lacked is necessary to select candidate technology.
2. Examination of task solution(1) Preparation of comparative approach between different technologies
Though comparative of different solidification technologies is necessary to select applicable technology, theinformation which is necessary for preparation of the comparative approach is lack. Therefore, existing
information (i.e. processing achievement, the number of steps, processing capacity, operability, economic
efficiency, solidified waste size, leaching rate, and so on) was investigated on each candidate technologies. The
items which is comparable between deferent technologies were selected from information obtained, comparable
items (horizontal axis) was arranged for candidate technologies (vertical axis), and the table was created forcomparison these technologies.
(2) Investigation for supplement of unsatisfactory information concerning low temperature processing
Among low temperature processing, the information lacked is concerning the influence factor, liquidity
evaluation method, influence of chemical components, and waste temperature in the case containing radioactive,in cementation. These unsatisfactory informations were supplied by investigating available information except
for concerning temperature of waste with radioactive. The information concerning the temperature of waste
containing radioactive was investigated by simulation of solidified radioactive waste.
(3) Experiment for supplement of unsatisfactory information concerning low temperature processing
Among low temperature processing, the information lacked is concerning usability evaluation (i.e.
measurement of fluidity and hardening characteristic), analysis concerning solidified substance performance (i.e.
measurement of compressive strength, dissolution, and hydrogen generation by γ-ray irradiation), and analysis
concerning standard (i.e. comparison of materials synthesized multiple institutions), in AAM processing. We did
experiment for supplement these unsatisfactory informations. The data obtained were comparison with the dataof cementation.
4. Next approach
In the next stage, the unsatisfactory information on high temperature processing is going to be supplemented.
The information concerning solidified waste which contains the contaminated water treatment secondary wastes
is lacking. Therefore, the information is going to be supplemented by investigation and experiment.
This report includes part of the results obtained from the work “Research and Development of Processing andDisposal of Solid Waste”, carried out on a supplementary budget allocated for Project of Decommissioning andContaminated Water Management. This work was studied by Central Research Institute of Electric Power
Industry, Taiheiyo Consultant Corporation, and JAEA.
References[1] Jun Kato，Akinori Nakagawa, Takumi Taniguchi, Tetsuro Sakakibara, Osamu Nakazawa, Yoshihiro Meguro (2017)：“TheCatalog of solidification and volume reduction technologies for the treatment of radioactive waste generated by thedecommissioning of Fukushima Daiichi Nuclear Power Station”, JAEA-Review 2017-015. (in Japanese)

50

J14

Sensitivity analysis of core slumping and alternative water injection
in Fukushima Daiichi Nuclear Power Plant Unit 3
Xin Li1, Ikken Sato1 and Akifumi Yamaji2
1

Japan Atomic Energy Agency, 2Waseda University

Abstract
This study aims at identifying the modeling uncertainties and addressing the sensitivity parameters inFukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core region has
been developed to better simulate the thermal-hydraulic behavior of liquid water and steam, which is considered
to be crucial in simulating the core uncovery and degradation process.
1. Introduction
In order to predict the possible status of the corium materials in Fukushima NPP Unit 3, by utilizing severe
accident analysis code MELCOR, the current study aims at identifying the modeling uncertainties andaddressing the sensitivity parameters in the first 55 hours after SCRAM of Unit-3.
2. Results
2-1. Core degradation behavior before RPV depressurization (09:00, March 13th, i.e. 42:14 h after
SCRAM)
Good agreement of the calculated RPV pressure and the water level with the measurement data wasachieved by tuning the the flow rates of steam extracted by RCIC/HPCI turbine from the core and the flow rates
of water injected to the core by RCIC/HPCI pump (see Fig. 1). Water level history shown in Fig. 2 suggested
that water level should have stayed above BAF before RPV depressurization.

Water level stayed above BAF

Figure 2. RPV Water level before
RPV depressurization

Figure 1. RPV pressure history before
RPV depressurization
2-2. RPV water level history before depressurization
Possibly 6 SRVs remained open to reproduce RPV
peak assuming whole core collapsed

Table 1 Summary of lower head failure time and RPV
water inventory conditions in sensitivity cases of

Figure 3. Sensitivity analysis of RPV history with

3. Conclusion•

different opening SRVs

6 SRVs opened during RPV depressurization at ca. 42:09 h after SCRAM and possibly remained open
when the major core slumping took place with the current MELCOR modeling conditions.

•Freshwater injection reaching RPV is most likely to be no more than 50% of that reported.

51

F01

Fuel Debris cutting and aerosol collection for
Fukushima Daiichi Decommissioning

Damien Roulet1, Rémi Delalez1, Christophe Chagnot2, Ioana Doyen2,
Christophe Journeau2,Christophe Suteau2, Emmanuel Porcheron3, Thomas Gelain3
1

ONET Technologies, 2CEA, 3IRSN

Abstract: A French consortium studies laser-cutting technology for Fukushima Daiichi fuel debris. Dedicatedlaser heads have been optimized and validated on simulants. Preliminary investigations of on-site
implementation of such cutting systems have been conducted. Two approaches are being studied to mitigate therisk linked to the generated radioactive aerosols during laser cutting: spray scrubbing and local collection near
the cutting point.
1. Introduction
Fuel debris in Fukushima Daiichi must be cut into pieces before their retrieval. A French consortium (ONET,
CEA, IRSN) has been studying laser-cutting technology for cutting of fuel debris [1]. This project is in line with
the implementation experience in UP1 French reprocessing plant decommissioning. Laser cutting has a great
advantage compared to many cutting techniques for its deployment by robots and remote systems. Tools arecompact, without contact with the material to be cut, and well fitted for being deployed by remote manipulators
in extreme conditions. Figure 1 sketches a possible implementation to cut ex-vessel fuel debris.

Figure 1: Example of laser head deployment in a 1F pedestal [1]2. Laser cutting technology
Dedicated laser heads have been optimized for in-gas andunderwater applications. Experimental validations on metallic andceramic fuel debris simulants have been successfully carried out.
3. Aerosol collection
One of the important safety issue is linked to the generated
aerosols. Two approaches are being studied to mitigate the risk
linked to the generated radioactive aerosols during laser cutting.
One concept is the use of spray scrubbing to significantly
reduce the aerosol concentration. This technology has been
investigated and the preliminary results show good
performance for a possible implementation during fuel debris
Figure 2: Laser head for underwater
retrieval [3]. Another approach is related to local collection
cutting in the DELIA test facility [2]of aerosols near the cutting point.
References:[1] D. Roulet et al., Development of innovative laser cutting technology for fuel debris
retrieval by robotic systems, FDR2019, J-Village, Naraha, Japan, 2019.
[2] C. Journeau et al., Fukushima Daiichi fuel debris simulant materials for the developmentof cutting and collection technologies, J. Nucl. Sci. Technol. 55:9, 985-995, 2018.
[3] E. Porcheron et al., Study of spray scrubbing as mitigation means against aerosol
dispersion during the Fukushima Daiichi fuel debris retrieval, ICONE27, Tsukuba, Japan,
2019.
Acknowledgements: This work has been carried out thanks to subsidized project ofDecommissioning and Contaminated Water Management (Advancement of
Figure 3: Spray
Fundamental Technologies for Retrieval of Fuel Debris and Internal Structures) fundedscrubbingexperimentby the Japanese Ministry of Economy, Trade and Industry (METI) and managed by thein TOSQAN facility [3]Mitsubishi Research Institute.

52

F02

Evaluation of the safety of geopolymer with respect tothe hydrogen production
JAEA – CEA Collaboration

Yoshikazu Koma2, Vincent Cantarel2, David Lambertin3,
Veronique Labed3, Makoto Arisaka2, Isao Yamagishi2
2 Japan Atomic Energy Agency
3 CEA – DEN – DE2D – SEAD - LCBC
Abstract
From 2019, a scientific collaboration is conducted between CEA and JAEA, aiming at
enlarging the knowledge on hydrogen production of geopolymer wasteforms. The hydrogen
recombination in geopolymers [1] will be further investigated to have a precise estimation of
the effective hydrogen accumulation in realistic storage conditions.

1. Introduction

Recently, geopolymer gained interest for the treatment of nuclear wastes, due to their specificchemistry and their ability to retain Cesium. However, this matrix contains water, susceptible
to produce hydrogen gas by radiolysis. For the immobilization of waste with non-negligible
activity, accurate prediction of the hydrogen accumulation around the waste packages isessential to mitigate the hydrogen explosion risk.

2. CEA experiment

Geopolymer samples will be irradiated at CEA site Marcoule, using the Gammatec facility(Figure 1). It is possible to measure in situ the hydrogen production using an experimental
setup designed in CEA. The hydrogen generated by the sample under 60Co irradiation will bemeasured during the experiment in a laboratory contiguous to the irradiation cell.

3. JAEA samples and hydrogen recombination model
JAEA will prepare the samples for irradiation and diffusion experiment. JAEA will be incharge of the data treatment and the use of a model to extract the effective hydrogenproduction and recombination rates. At last, diffusion experiment will be carried out to obtaina precise value of the hydrogen gas diffusion constant in the geopolymer wasteform.
Simulations in 3D will be used toestimate the hydrogen accumulation
around geopolymer based wasteforms.
References[1] Cantarel, V.; Arisaka, M.; Yamagishi, I. On thehydrogen production of geopolymer wasteforms
under irradiation J. Am. Ceram. Soc. 2019 00:1–
11. https ://doi.org/10.1111/jace.16642

Figure 1 : Gamma irradiation facility(Gammatec – CEA Marcoule Site) with 60Co source

53

F03

Orano: Working in Partnership with Local Communities
Over 50 Years of Regional Integration
Xavier VERDEIL, Maryline BRETON, Daphné OGAWA
Orano Group

Abstract
For many decades, Orano has been actively involved in promoting regional economic development andengaging with stakeholders at all of its sites worldwide. Orano’s experience with the la Hague site: going beyond
regulatory requirements for successful integration in the region.
1 – Introduction
With more than 5200 employees, the Orano Group is the number one industrial employer in theNormandy region. Approximately €400 million are spent in purchases from Norman companies each year.
The Orano La Hague site also participates in economic development in the region by actively supporting thecreation of jobs in all sectors of the economy, including the Social and Solidarity Economy.
Furthermore, Orano La Hague engages in regular dialogue and consultation with all stakeholders.
2 – Involvement of Orano Group: the Example of La Hague site
2-1 – Dialogue with Stakeholders
The Orano La Hague site communicates on a regular basis with its external stakeholders, at the Local
Information Commission for instance, or using the Nuclear Safety Transparence report. Thousands of visitors
come to the site every year, including many journalists and decision-makers. The site also regularly exchanges
with regional actors and elected officials, communicates news from the site to the press or on social media, andparticipates in public meetings.
2-2 – Attractiveness and Employment
Orano La Hague has established close connections with certain universities and schools in the region
and with all employment actors, including the main occupational branch. It has contributed to the launch of
several training programs and employees teach specific courses. Orano La Hague participates in many
professional or job fairs for a variety of audiences (schools, universities, job seekers). In 2018, the site organised
its own job fair, welcoming over 4000 visitors.
2-3 – Supporting the Nuclear Industry
Orano La Hague is a founding member of Nucleopolis byNormandie Energies, a cluster dedicated to economic developmentof the Norman nuclear industry. It therefore participates in various
actions organised in the sectors of business (Norman Nuclear
Industry Business Day, Focus’Business Day on the site, etc.),
innovation and technical skills, such as the operational excellencelabel, Exc’op, which Orano sponsors.
Figure 1. BtB meeting as part of the
Focus’Business Day with suppliers

2-4 – Participation in the Local Economy

Orano is a partner in organising several sporting and cultural events in the local community. It is also amember of economic development structures: the Alizé Manche Committee (which provides small companies
with free access to the skills of larger companies) or Normandie Incubation, a startup incubator.
To encourage employment opportunities, Orano provides financial support to local companies and non-profits,
including those participating in the Social and Solidarity Economy. Since the end of 2016, around 220 local jobs
were created with the support of €500,000 from Orano.

54

F04

Investigation and Characterization in hostile nuclear environment
Kenny Le Flanchec, Cédric Escoffier, Simon Haen, Sébastien Bouzat,
François Catsivelas, Frédéric Chambon, Daphné Ogawa
Orano Group

Abstract
Investigation and Characterization are key activities to build robust Decommissioning scenario. Orano has
designed a significant number of innovative solutions and successfully implemented them in very complex
environments. The characterization and the sampling of high activity deposit settled at the bottom of fission
product evaporators is one of those Orano solutions which are currently being adapted to perform corium (Fueldebris) sampling for Fukushima reactors.
1.

Introduction

Collecting reliable input data is essential to build robust decommissioning scenario and define waste routes.
Therefore, many innovative solutions have been designed by Orano and successfully implemented in very
hostile environment, such as the investigation within fission product evaporators.
2.

Investigation within fission product evaporators at the CEA Marcoule site

Between 2016 and 2018, on behalf of the French Atomic Energy Commission (CEA), Orano DS teamconducted investigation work at Marcoule on three evaporators that were operated from the 1960s to the 1990s
to concentrate fission products due to the fuel reprocessing operations. The first step was to core drill through a
one-meter thick heavy concrete block wall and then drill through the evaporator to access inside of the vessel. A
dedicated technology of trepanning was developed to cut the 12 mm thick Uranus steel vessel.
Different kind of investigation had to be performed: visualinspection, dose rate measurement, smear, vessel bottom headdeposit depth measurement, deposit sampling. Due to thehighly radioactive environment (dose rate from 1 to 10 Gy/h at
the vessel bottom), Orano designed, procured and tested two
multitask mechanical robotic arms able to carry customized
tools like depth measurement device, scraper, mechanicalsampler, suction sampler, collimated radiological probe, from
top and side access.
To limit the ambient dose rate at the operating position and meet the dosimetry objective initially defined for theworkers, a new biological protection called IRIS (Investigation Radiological Innovation Safety), designed in theframework of Orano’s R&D activities, was applied for the first time. IRIS is a diaphragm-like radiologicalshield which allowed Orano to divide by ten the integrated dose by our team and thus meet the ALARA
objective. The project is now successfully completed. As specified by our client, nineteen deposit samples havebeen taken from the three evaporators and more than one hundred radiological and physical measurements havebeen performed. Forecasted operator radiation exposure was respected.
3.

Fuel Debris (corium) sampling tool for Fukushima reactors

Based on its large experience in the design of various sampling tools and their implementation in several projects
like the Marcoule evaporator one, Orano is now adapting a specific solution to meet the need of characterizationof fuel debris fragment within Fukushima reactors. Indeed, in cooperation with MHI within IRID project, Oranois currently performing the design of a grasping-type device to recover several samples of corium fragments.
Conceptual design and preliminary trials with 3D-printed sampling tools have already been successfullycompleted. A new phase concerning basic design and steel body prototyping is ongoing.
4.

Conclusion

Orano D&D teams have acquired over 20 years of hands-on experience in investigation and characterization inhighly radioactive environment. Innovative robotic equipment was designed, tested and successfully
implemented to support the D&D program of the first generation of French reprocessing plants. These provensolutions can now be shared and adapted to the needs of new challenges such as Fukushima-Daiichi.

55

F05

TEΩ600, the industrial solution for robotics in hostile nuclear
environment (up to 1MegaGray cumulated dose)
Alain Coudray1, Xavier Verdeil2, Daphné Ogawa3
1
Orano Temis, 2 Orano Cycle, 3Orano Japan

Abstract

The TEΩ600 is a modular, economic and robust solution, based on an industrial arm to guaranty
performance, reliability and spare part availability. TEΩ600 is designed for high precision and dexter distantwork using remote control with force feedback, and for complex and/or repetitive tasks in automatic mode byrunning programmed cycles. A wide range of tool (handling, cutting, cleaning, machining, welding…) can beused with TEΩ600 thanks to its performant payload of 60kg in all its workspace. Embedded on a mobile carrier,it can be easily adapted for Fukushima requirements with a long-distance remote control station (several
kilometers).

1. Introduction

Decommissioning works at 1F site are very challenging in terms of dose rate, contamination risk andaccess difficulty. The TEΩ600 is a telerobotics solution with force feedback technology designed for remote
operations requiring sensitivity and contact-like control to be performed in hostile environment as nuclear sites.
D&D projects are often singular and have restraint budget, thus, adapting off-the-shelf industrial robot
allows proposing reliable solutions at low price with little development.

2. Innovation

Based on Orano Temis extensive experience,TEΩ600 is specially designed such as follows fornuclear environments:
 Radiation hardening at high cumulated dose Improved maintainability (rapid, semiautomatic) in hostile environment(contaminated, high dose rate, limited space
and time) Management of the contamination (avoid
contamination insertion, decontamination)
 Evolved handling of malfunctions
 Accessibility to target area thanks to the useof a robust and powerful industrial crawler


Improvement of operability (visibility,
sensors, 3D model vision), etc.

Figure 1: TEΩ600 Architecture

3. Key features









Cartesian position and force-feedback Master-Slave control
Prevention of internal contamination by air purging to the casing
Surface smoothing and dedicated for easy decontamination
Water proof, dust proof
Equipped with cameras in the proper position
Can be attached to various tools by modular design
Able to be embedded to various carrier as lifting crane, or crawler
Able to be operated with protect gear and mask

Figure 2: TEΩ600 on a mobile carrier(MC-TEΩ600)

4. Conclusion

Orano
has
acquireduniqueexperience in remote solutions for
performing D&D projects. Orano can
propose integrated solutions (soft andhard) for Fukushima-Daiichi and its
ever more demanding challenges
ahead.
Figure 3: ATOX, ANADEC and Orano collaboration:
Operator training and operation tests type for Fukushima (ATOX, Kashiwa)
56

F06

Tools and technical means for under-water decommissioning in highly
contaminated environment
Jean-Michel CHABEUF1, Malo DIENNET2, Alexandra SYKORA3, Daphne
OGAWA4
Orano Dismantling & Services, 2Ecole Centrale Supelec, 3Orano GmbH,
Orano Japan

1
4

Abstract
Segmentation of Reactor pressure vessels and their internals have been performed by Orano on a variety ofreactors in Germany and the United States Mainly. The underwater segmentation of core waste is also routinely
performed with industrialized solutions. Water purification is a key success factor for under water segmentation.
Finally, innovation allows improving reliability and safety of segmentation operations.
1. Introduction
In the areas of the remote segmentation of reactor pressure vessels (RPV) and internals, as well as
associated metallic structures such as control rods or fuel racks, Orano has developed and deployed successful
technologies since the end of the nineties for nuclear power plants and research reactors worldwide, delivering
projects on time and within budget.
2. Remotely operated under water segmentation
2.1 RPV Internals & RPV
Dismantling, conditioning and packaging activities of RPV and internals are performed with proven remoteoperated technology and underwater. Operations have been conducted in several BWR and PWR NPP such as
Würgassen, Stade, Mülheim-Kärlich, Brunsbüttel, Phillipsburg and Neckarwestheim in Germany.
2.1 Pool internals and core waste D&D
Orano has delivered the segmentation of core waste inseveral reactors, such as Biblis or Isar 1. In the latest
project at Brunsbüttel facility, controls rods, neutron flux
measuring lances, fuel element racks, a basket withtechnological waste and filter cartridges, have been cut
and packed as shown on the figure 1.

Water
purification
systemSaw(horizontal
cut)Baskets

Telescopic mast

2.2. Water filtration and decontamination
Figure 1 : Control rods
cutting

Figure 2 : Roughfiltration

Figure 3 : Fine filtration

The setting up of an auxiliary filtration anddecontamination system is often necessary to decrease
the visibility loss due to particles in suspension. During
the D&D activities of Orano, both fine and roughfiltration are used (resp. figure 2 and 3). The first one isa flexible machine with several setting possibilities
allowing to filtration of fine particles of only 2 μm; andthe second one is a rough filter which carries out
particles of 20 μm through hydrocyclones, to be
separated in a basket retrieved remotely.

3. Conclusion
Industrialized and proven segmentation and filtration solutions are deployed on a variety of projects, and Oranocontinues innovation to further improve the timely delivery and safety of its projects.

57

F07
Abstract

DEM&MELT In-Can Melting Technology
for the Vitrification of D&D and Remediation Waste
Regis Didierlaurent1, Thierry Prevost1, Daphné Ogawa1, Hubert Alexandre Turc2,
Kohei Shibata3
1
Orano, 2CEA, 3ANADECThe DEM&MELT In-Can vitrification process is designed to process intermediate and high level waste
coming for D&D operations. It is designed to treat liquid and solid waste, to produce a small amount of
secondary waste and to minimize investment and operating costs. This In-Can vitrification process is also
developed with a compact and modular design and can be adapted regarding to nuclear operators’ needs andrequirements. Furthermore, this robust and simple process is flexible enough to accommodate most uncertainties
on waste composition. Considering the DEM&MELT features, this process is deemed as an appropriate tool forthe vitrification of the Fukushima Daiichi water treatment secondary waste. The DEM&MELT process isdeveloped through the French consortium which gathers Orano, CEA, ECM technologies and Andra.
1.

Introduction

Following the earthquake and the tsunami on March 11 th 2011 in Fukushima, large quantities ofcontaminated water have been generated due to the cooling of the damaged reactors at Fukushima Daiichi
Nuclear Power Plant (1F). Several water treatment systems have been used to decontaminate this water and have
led to the production of many secondary solid wastes so called FETW (Fukushima Effluent Treatment Waste).
These waste, mainly contaminated with cesium and strontium, are temporarily stored, waiting for a process tostabilize them for final disposal. ANADEC, Orano and CEA have proposed to stabilize the FETW using the InCan vitrification technology. A dedicated tests program on Fukushima Daiichi wastes, supported by Japanese
government funding, has started in Japanese FY2018 and will continue until FY2020 in the framework of The
Subsidy Program “Project of Decommissioning and Contaminated Water Management”. The main goals are todemonstrate the feasibility and to provide a first preliminary design of an industrial unit adapted to 1F waste
treatment.
2.

Evaluation of the In-Can melter for Fukushima water treatment secondary waste – Main results

To demonstrate the FETW treatment feasibility by the In-Can vitrification technology, a step-by-step work
process has been implemented. The first step has consisted in studying various waste assembly scenarios andglass matrix formulation by smartly optimizing treatment time, cost and final volume of waste packages. During
this step, glass samples were elaborated at small scale and characterized (~100 g of vitrified material for eachsample). It has been shown that it is possible to obtain a dense vitreous matrix with high waste loadings, ranging
from 50wt% up to 80wt% depending on the waste scenario assembly. In a second step, bench scale tests havebeen performed on intermediate scale mock-ups (~1kg of vitreous matrix). They have confirmed the feasibilityof the various scenarios and have led to the determination of a first set of operating parameters. In the final step,a pilot scale demonstration test has beensuccessfully performed on the In-Can-Melter
vitrification pilot (~100kg of glassy material) for areference scenario with representative nonradioactive surrogate waste. Analysis of glass andoff gas samples performed during pilot test allowed
assessing cesium behavior in the process. The
waste is well incorporated into the vitrified
material. The efficient temperature control and thehighly efficient Off Gas Treatment System allowed
an efficient volatility management for Cs, which Figure 1. R&D method from laboratory to pilot scale tests
remains lower than 0.5 wt%.
3.

Conclusion

An evaluation of the In-Can vitrification process industrial implementation and the associated waste
management was performed. The feasibility of the In-Can vitrification process for FETW has been
demonstrated for 90 % in volume of the total inventory and 98.6 % of the total activity during FY2018 study. A
complementary R&D Program is currently ongoing to complete the demonstration that the In-Can vitrification
is fully adapted to the FETW characteristics.
58

F08

New peelable coating RTV FA 878
when dismantling the nuclear facilities
J.P. Thome, BCSN & M. Sawamoto, eEnergy Corp.

Abstract: newly developed coating now used at nuclear facilities for the purpose of surface
coating against contamination spread-out, strengthening leak-tightness as well as
decontamination
1 Characteristics of coating:Translucent elastomeric compound
Low Hardness Shore A (28)Sprayed with instant cure without linear shrinkage
Good mechanical properties (tensile, strength, elongation at break, tear, resistance)
Outstanding lifetime under irradiation
Injectable underwater
Easy retrieval
Self repairable
2 Use
Protective flexible skin against shocks or contamination
Reinforce depression tightness of airlocks & vent ducts
Fixing of loose contamination & high efficiency decontaminating strippable coating
Waterproof coating for submerged equipments in pools
Tightness coating against corrosion
3 Conclusion
We hope this material will be helpful for the cases at Fukushima site where there are needs
against dispersion of contamination, for leak-tightness, and for decontamination.

59

F09

Introduction of ROBATEL Industries business and theirexperiences regarding hot cells project
Christopher DANE2, Natalia ZOLNIKOVA2, Masahiro Sawamoto1
and Keiichi Inoue1,
ROBATEL Industries, 1e-Energy Corporation,

2

Abstract
The company ROBATEL Industries is a French SME founded in 1830, whose activities are exclusively turned
towards the nuclear industry for over 60 years. ROBATEL has developed a comprehensive range of skills andexperience.

1. Introduction
The activities of the company cover both aspects "design" (studies, calculations, safety assessment…) and"production" (production, commissioning tests, maintenance, on-site interventions and loading assistance). Thanks
to its long experience, ROBATEL is capable of putting its expertise to the service of major high stake projects forthe industry; developing, case-by-case, hot cells (among other fields) and all their associated operating tools(remote handling equipment, control equipment and others...) to meet our customer specific needs andrequirements.
2. Our experiences
Quantities of hot cells fabricated in recent history by ROBATEL:Over the past thirty years, 46 hot cell facilities and cementing units were completed:

More than 200 hot cells



For 14 different clients

In 8 countries

And more precisely since 2000:


More than 45 hot cells

5 for different customers

In 7 countries.

3. Conclusion
We hope above our experiences will be helpful
for decommissioning program in Fukushima 1st
nuclear power station.

60

F10

Mechanical Installations and Equipment for Decommissioning andfor Waste Storage
廃止措置および廃棄物保管用機械設備および機器

Craire PAOLONI1, Vincent GILLET1, Renaud BOURDY1, Masahiro Sawamoto2,
and Keiichi Inoue2
Cegelec CEM1, e-Energy Corporation2

1. Introduction
Cegelec CEM is a French company, belonging to VINCI Energies Corporation, specialized in mechanicalprocesses, facilities and equipment for Decommissioning and Nuclear Plants operations.
2. Handling system for the French Future Final Disposal (CIGEO)
Cegelec CEM realizes concept design and studies of all the handling process equipment of high and medium
activity nuclear wastes canisters from the level 0 to the level -500 m (under the ground level): cranes, waste hoods
and maintenance facilities.
Airtight
door

Hood

Wagon

Hood forwaste

Remote crane

Nuclear
waste

Handling
cell /
equipment
beforestorage
Remote wagon into tunnel‐500 m lower from level 0

Handling equipment under the ground‐500 m lower from level 0

3. Nuclear Waste interim storages and in-cell handling equipment
Cegelec CEM performs engineering and turnkey projects for nuclear operators in France and overseas, fromstudies to manufacturing, testing, and on-site installation and commissioning.
Cegelec CEM has in house engineering skills for design and calculations (mechanical, seismic, thermic andvarious nuclear calculations), to propose innovative solutions to its customers.

Waste lorry
docking on cell

Equipment / handling system in cells

In cell carrying solutions

4. Demonstrators and specific equipment for nuclear activity
Coring machine

HLW
ILW

Experimental
corroded area

Handling
equipment

Experimental remote extraction of HLWCoring machine (waste sampling)

Extraction
robot

Pushing machine for HLW
into disposal

5. Conclusion
By its long track record in Nuclear Waste management in Europe, Cegelec CEM can bring to Fukushima
decommissioning project its know-how and proven solutions.
6. Referenceswww.cegelec-cem.com

61

GB01

Pile Fuel Cladding Silo Decommissioning Project –Collaboration
Sarah Mundy, Roger Cowton
Sellafield Ltd.

The Pile Fuel Cladding Silo (PFCS) was built in Sellafield in 1949/1950 for the purpose of receiving
and storing the fuel cladding material from the Windscale Pile fuel reprocessing. Its secondary purpose
was to store miscellaneous waste from the UK nuclear programme. Due to its construction, PFCS is
unsuitable to store nuclear inventory for an indefinite period. Coupled with the unsuitable storage
qualities, the uncertainty of the nuclear inventory causes PFCS to be one of the most hazardous nuclear
facilities in Western Europe.
To mitigate the ongoing risk associate with operating PFCS, Sellafield Ltd have been designing andbuilding a retrievals system. To manage the uncertainty associated with inventory and structure without
the excess of a bounding design, the programme has devised a 2 step approach - early retrievals andfull retrievals where early retrievals will use a lead and learn approach. The lead and learn approachallows feedback of knowledge regarding the waste, retrievals process and equipment gathered during
early retrievals to be implemented within the full retrievals design. This aims to reduce design periods,
failure rates and streamline the retrievals process.
To deliver PFCS retrievals, Sellafield Ltd have partnered with supply chain partners Bechtel Cavendish
Nuclear Solutions (BCNS - a joint venture between Bechtel Inc and Cavendish Nuclear Ltd). Runninga project with such high levels of uncertainty on a traditional client contractor basis proved ineffective.
It was recognised that the strengths of all organisations working in a highly collaborative relationshipwold be needed to deliver the work. The project was overhauled in 2015 to focus on collaboration;
since then, the projected cost has reduced by £100m with early retrievals being delivered 17 months
ahead of schedule.
These improvements would’ve not been achievable without collaborative working between Sellafield
Ltd and BCNS, innovation solutions and a Fit for Purpose mind-set. Stakeholder confidence has been
re-affirmed and Sellafield is on the brink of retrieving legacy waste from PFCS!

Figure 1 - The retrievals equipment - from a model to fully built.

62

GB02

Technologies for Detection of Contaminated Particles
Christopher Carter1, Jonathan Britton1 and Mark Rouse2
1

Cavendish Nuclear Ltd, 2Cavendish Nuclear Japan

Abstract
During decommissioning activities the area surrounding the decommissioning site is monitored to confirm
that contamination has not been released into the environment.
Cavendish Nuclear has developed technologies and techniques that are used for monitoring the air and theground for detection of very low levels of radiation that are associated with small particles of contamination.
1. Introduction
Cavendish Nuclear is an experienced operator and developer of radiological environmental monitoring
systems.

Cavendish Nuclear’s expertise ranges from overall strategy development to investigation planning andimplementation, laboratory testing, data interpretation, visualization using geographic information system (GIS),
risk assessment and data/records management.
Hot particles with alpha emissions are of particular concern if they have entered the environment due to thehigh potential for harm if they come into contact with operational staff or public.
2. Technologies for Monitoring and Detection of Contaminated Particles
The poster will describe examples of equipment and techniques that can be applied for environmental
monitoring at decommissioning sites, including the three technologies shown below.
Large Area Radiological Characterization (LARCH)
The LARCH system deploys a high-resolution gamma-ray detector in conjunction
with automated surveying technology to produce a series of spectra that are stampedwith the date, time and GPS location from which they were acquired.
Environmental Monitoring and Characterization – RadScan®
RadScan is a portable gamma ray imaging system that remotely locatesand characterizes gamma radiation from a wide variety of environments
including building surfaces, land, cells, gloveboxes, drums and process
vessels.
RadScan automatically scans and records gamma spectra from themeasurement area and displays this as a colored overlay on a color video
image.

The image clearly shows where the radioactive contamination is

located and its intensity.
Environmental Monitoring – Drop & Go Monitor
The ‘Drop & Go Monitor’ is a portable instrument which autonomously measures Alphaand Beta contamination and Gamma Dose Rate.
This radiometric data, together with positional and status information, is transmitted via
the 3G/4G mobile phone network back to a secure cloud service which can be accessedfrom any location via a standard internet browser with access to the internet.

63

GB03

Integrating Waste Management with Project Management:
the Key to Successful Decommissioning
Bill Miller, David Luckhurst, Darren Dudley, Steve Caley& Paul Fleming
Wood plc

Abstract
Decommissioning is a waste generation activity. But decommissioning projects are too often
underprepared for the challenge of managing and disposing of wastes, leading to project delays
and cost increase. At Wood we work with site owners to integrate decommissioning plans with
waste management plans. This integrated approach delivers substantial safety and cost benefits,
and can often lead to a significant reduction in the overall volume of waste to be disposed.
1. The challenge of waste management
At its most simple, a decommissioning project is a waste generation activity that can produce verylarge quantities of many different waste materials, both radioactive and non-radioactive.
Unfortunately, there are several examples around the world of decommissioning projects thathave underestimated the difficulties associated with waste management, and have failed
adequately to plan and prepare. As a result, decommissioning projects are too often delayed
because waste treatment and disposal facilities are not available when they are needed or are notsuitable for the wastes that are actually generated. This can lead to schedule delays and greatly
increased costs, and the loss of regulatory and stakeholder confidence.
Learning from experience, it is important to recognise that waste management (and disposal) isvery often the rate limiting step on the critical path for a decommissioning project.
2. The need for waste management plans
It is important to prepare for decommissioningby first developing a robust decommissioning
plan and work breakdown structure (WBS), andthese must be underpinned by a sensible
engineering approach and reliable estimates forcost and schedule.
However, it is recommended that thedecommissioning plan is fully integrated with adetailed waste management plan. As a minimum
the waste management plan should specify:• an estimate of the waste inventory, identify
all waste materials (both radioactive and non-radioactive) and their volumes• the best available techniques (BAT) for characterisation, segregation, treatment, condition andpackaging of each waste material• the anticipated interim storage and disposal routes for each waste, and the acceptance criteria• the infrastructure requirements, including any new facilities needed for all waste treatment,
storage and disposal activities• a reasonable estimate of the schedule for waste management and disposal (which may takedecades), and the associated costs for each activity.
3. Integration of decommissioning and waste management plans
At Wood plc we work with nuclear sites to develop and integrate decommissioning plans with
waste management plans. This integrated approach (sometimes called waste-led
decommissioning) has delivered substantial benefits, including costs savings, by optimising work
while at the same time reducing the overall volume of waste to be disposed.
For example, we have undertaken several projects to plan and implement waste retrievals anddecommissioning of legacy silos on nuclear sites. Our work integrates all planning, licensing,
options assessment, safety case, engineering design for retrievals equipment and waste
immobilisation equipment, in some cases using Wood’s own SIAL® geopolymer.

64

GB04

Innovative Cost Effective Solutions for In-Situ Liquid Waste
Processing
Timothy Milner1
ATKINS

1

Abstract
The poster details the concept of using a Rotating Bed Reactor (RBR) that was pioneered by Swedishcompany SpinChem and has achieved success in the pharmaceutical and other non-nuclear industries. Atkisn has
teamed with SpinChem to develop the SpinIonic® RBR for use in the nuclear industry.
Outline of System
The Atkins and SpinChem team developed the nuclear application for the SpinIonic® RBR and is now
actively demonstrating the technology at client nuclear sites. The RBR device retains the solid phase as a packed
bed inside a rotating cylinder. As the RBR spins, a continuously circulating flow develops. Reaction solution is
rapidly aspirated into the bottom of the RBR vessel, percolated through the solid phase, media, and quickly
returned to the storage tank or vessel being treated. By the intelligent design of the SpinIonic® RBR, the axial
mixing and convective transport are maximized. The resulting efficient mass transfer minimizes reaction time,
boosts material capacity and increases process flow rates.
Conclusion
With the SpinIonic® RBR it is possible to minimize slow reaction kinetics caused by poor mass transfer
between the solution and solid phase. The SpinIonic® RBR design is flexible and can be used for numerous
applications in situ to treat tanks for radioactive waste or in process to treat continuous waste streams. Utilising
the SpinIonic® RBR typically results in faster processes, higher decontamination factors or reduced generation of
secondary waste, depending on the type of process. In addition, the SpinIonic® RBR extends the lifetime of thesolid phase particles by minimizing grinding and fines, while at the same time simplifying the solid phasecollection and recycling or disposal.

65

GB05

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

The importance of correctly identifying and managing the cost of decommissioning
nuclear facilities
Gary Mills, Director, Gleeds Energy Limited, UK

ABSTRACT
It is important that we understand the real cost of decommissioning a nuclear facility, so the correct level of funding can bearranged and then correctly managed to ensure the activities are safely completed. This poster focuses on how the collection
and application of detailed data has enabled Gleeds to provide accurate estimates that influence procurement and contracting
strategies and enable effective management and control the cost.

66

SK01

External & Internal Dose Assessment on Spent Resin Treatment
Facility for Recycling C-14
Woo Nyun Choi, Ukjae Lee, Bae and Hee Reyoung Kim

School of Mechanical Aerospace and Nuclear Engineering, UNIST, 44919 Ulsan, Republic of Korea
Abstract
External & Internal dose assessment are conducted to ensure radiological safety during the treatment anddisposal of spent resin. This study performed a radiological evaluation of the operation of spent resin waste
treatment facilities of Wolseong Nuclear Power Plant, a heavy-water reactor nuclear power plant in Korea.
1. Introduction
In the heavy-water reactor, ion exchange resin is used to purify the liquid radioactive waste generated while
operating, about 10,000 L or resin is generated annually. The resin stored in the storage tank should be handled inaccordance with future decommissioning plans [1].
2. Method and Result2-1. Modelling the spent resin treatment device
The spent resin treatment device which is a demonstration
scale (10 L) (Fig.1.) is modelled based on basic structures which
are supported by VISIPLAN (SCK-CEN, Belgium). The device
was modelled with stainless steel material.

Fig. 1 Actual resin treatment device

2-2. Assessment of external and internal dose

The 100 % leakage of radionuclides is assumed. The worker radiation exposure dose was calculated inaccordance with the worker scenario using VISIPLAN. A typical radiation worker’s annual working time of 2,000h was used to derive the annual worker dose. And the internal exposure dose was evaluated based on the activity
concentration of the actual spent resin from Wolseong nuclear power plant.
2-3. Result
For 1 h of working time, the received dose was calculated as 5.3E-05 mSv/h. The typical annual working timeof radiation workers (2,000 h) was used to derive the annual worker dose based on the received dose per hour. The
derived dose value was 0.106 mSv. The 137Cs had the largest influence on the external dose value at 6.30E-02 mSv.
And the value of internal dose of for the first year was 10.34 mSv.
3. Conclusion
In evaluating the external & internal exposure of workers to radioactive waste from the spent resin treatment
device, the resulting radiation dose was 10.446 mSv, which is below the Korea annual worker dose limit of 50 mSv.
Thus, the radiological safety of workers in the spent resin treatment process was confirmed.
References[1] Fiskum SK, Blanchard DL Jr., Steele MJ, Wagner JJ. Analysis of spent SuperLig® 644 resin used for cesium removalfrom Hanford tank wastes. Solvent Extra Ion Exc. 2006;24(1):65-79.

67

SK02

A Transportable Ground Radiation Monitoring System for
Decision-Making on Release of Nuclear Decommissioning Sites
Chanki Lee, Hee Reyoung Kim

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology
Abstract
We develop a transportable ground radiation monitoring system, which adopts Compton suppression

method with NaI and PVT scintillators. In order for the system to be efficient to making decision on therelease of nuclear decommissioning sites, we simulate scan minimum detectable concentration (MDC)
related to a preliminarily derived concentration guideline level (DCGL) of specific site [1], that is, Kori-1
nuclear power plant (NPP).
1. Introduction to the monitoring system
The system is based on a trailer as well as a gamma spectrometer that analyzes signals coming from a main
detector and a guard detector. The main detector is composed of two NaI scintillator, and the guard detector is ofthree PVT scintillator. Using anti-coincidence circuit, the system reduces background radiation from Compton
scattering.
2. Performance evaluation methods
2-1. Detection efficiency
Two cases are set to evaluate detection efficiency with distance from 0.1 m to 0.5 m: Surface
contamination and vertical contamination with different depth profiles. Depth profiles are assumed by using
Gaussian distribution [2] of MCNP code.
2-2. Estimation of suppressed background
Compton suppression factor (CSF) is calculated to quantify how much background counts are suppressed.
Actual data are used to select proper energy bin of MCNP PHL card.
2-3. Scan MDC calculation
By using conservative assumptions from evaluated performances, scan MDCs are calculated. Especially,
those of 137Cs and 60Co isotopes are compared with their DCGLs of Kori-1 NPP.
3. Conclusion

Simulation results of scan MDC show that the proposed system should be operated at velocity below
3 m/s, which is the maximum optimal velocity for application to Kori-1 NPP decommissioning. Overall,
the results imply that the system can possibly be a rapid but integrated (i.e., gamma radioactivity and doserate) evaluation tools on site. In addition to the simulation results, more strict validation based on actual
test should be conducted.
References[1] Abelquist, E.W., 2001. Decommissioning health physics: A handbook for MARSSIM users. Taylor & Francis.
[2] Waters, L.S., 2002. MCNPX User’s Manual. Los Alamos National Laboratory.

68

SK03

Dose calculation of dismantling workers for bio-shield of Nuclear
power plant
ChoongWie Lee and Hee Reyoung Kim
Ulsan National Institute of Science and Technology

Abstract
The doses of the workers at the time of bio-shield dismantling with various radioactivity distributions were
evaluated. The working scenario for dismantling of Bio-shield a was derived, and the total dose during
dismantling Bio-shield were evaluated accordingly. In case of external exposure, VISIPLAN, a dose assessment
tool, was used. In case of internal exposure, the amount of dust production was calculated, and the dose wascalculated.
1. Introduction
The Kori Unit 1 in Korea is planned to be closed and decommissioning accordingly. In the situation of
dismantling of nuclear power plant, evaluation of worker's dose is essential for securing worker's safety. A
radiological safety assessment is required for the safety of workers and authorization in the decommissioning.
2. Methods and Results
2-1. Dose calculations methods
Since there is no current radioactivity information for Kori Unit 1, it is necessary to calculate the current
radioactivity distribution based on the initial conditions of the core. For this purpose, the radioactivity
distributions for the five major radionuclides (3H, 55Fe, 60Co, 152Eu, 154Eu) were calculated using MCNP codeusing information such as neutron flux, nuclide concentration, and operation history. Then, the dose rate wascalculated using the dose assessment tool, VISIPLAN, and the external exposure was evaluated according to thedismantling work scenario. The internal exposure was calculated by modeling the amount of dust production by
concrete dismantling and taking into consideration the dose coefficients.
2-4. Results
The results of simulating the dose rate inside the radioactive concrete using VISIPLAN are shown in Fig 1.
As a result of the dose evaluation, the external exposure of the worker was 80 mSv and the internal exposure
was 3 mSv. It was shown that the external exposure was more affected than the internal exposure.
3. Conclusion
As a result of the dose evaluation, the calculated dose was morethan annual dose limit in Korea (20 mSv/y). Therefore, it is expectedthat protective measures such as remote control and protective
clothing will be needed for the safety of workers dismantling thebio-shield.
Figure 1. Dose rate at Bio-shield

References[1] D. Lee, H.R. Kim, Preliminary Source Term Assessment on Kori unit 1 Radioactive Bioshield for Decommissioning,
Transactions of the Korean Nuclear Society Spring Meeting, (2018).

69

SK04

Comparison of Detection Efficiency according to the number of
Scintillator in Continuous Tritium Monitoring System
Jae Hoon Byun, 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
The detection efficiency was compared according to the number of scintillators inside the chamber used in thecontinuous tritium monitoring system. The 13-slot chamber’s efficiency is 13.12% and that of 7-slot chamber is
19.41%. The efficiency of the 7-slot chamber was 6.29% higher than that of 13-slot chamber.
1. Introduction
In decommissioning sites, underwater tritium should be monitored. The energy generated by tritium is so lowthat the detection efficiency is low due to the self-absorption [1]. So, the tritium monitoring system using
electrolysis was used. In this study, the detection efficiency was analyzed for both conditions of the detection
chamber.
2. Materials and Methods
2-1 Experimental Setup
To compare the detection efficiencies of 13-slot chamber and 7-slot
chamber, the study was carried out. The 170 kBq/m3 of tritiated gas wasproduced and measured for 3 mins.
2-2 Comparison of two chambers

Figure 1. 7-slot detector chamber

Comparing the characteristics of the two chambers, the 13-slot chamber has a relatively large number ofscintillators, but the amount of gaseous tritium in the chamber is small. In contrast, the 7-slot chamber has arelatively small number of scintillators, but the amount of gaseous tritium in the chamber is large.
2-3 Results
13-slot chamber’s average efficiency is 13.12%, 7-slot chamber’s average efficiency is 19.41%. The efficiencyof the 7-slot chamber was 6.29% higher than that of 13-slot chamber. Of course, as the number of scintillators
decreases, the volume of the scintillators decreases, but, it was found that the amount of tritiated gas inside thechamber influenced the detection efficiency more. Also, 7-slot chamber is cheaper than 13-slot chamber.
3. Conclusion
The detection efficiency was compared according to the number of scintillators inside the chamber. The
efficiency of the 7-slot chamber was higher than 13-slot chamber. So, the device with simple composition, lightmass and low-cost for practical applications can be used in real time detection in decommissioning sites. In the
future, MDA calculations should be carried out, and further studies should be conducted to increase theefficiency.
References[1] Soreefan, Aurelie M., and Timothy A. DeVol. "Proportional counting of tritium gas generated by polymer electrolyte
membrane (PEM) electrolysis." Journal of radioanalytical and nuclear chemistry 282.2 (2009): 517.

70

SK05

Real-time Monitoring of Tritium in Water with Enhanced
Detection Efficiency
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 plastic scintillation-based tritium detection chamber was improved. The volume of the chamber was
enlarged and designed with consideration of higher detection efficiency. The maximum detection efficiency andminimum detectable activity (MDA) for tritiated water was evaluated as 43±4.3% and 3.4 kBq/L, respectively.
1. Introduction
Tritium contamination is one of major concern of Fukushima Daiichi Nuclear Power Plant site [1]. In
previous study, an electrolysis-based real-time tritium detection system was developed to monitor the abundanceof samples and constant monitoring of tritium in water. The MDA of the system was 17.8 kBq/L for 15,000 s ofmeasurement time, but it was quite high to be used as a low-level contamination monitor for aquatic
environment. In this context, the detection chamber was improved to lower the MDA.
2. Methods and Results
Figure 1 shows the improved tritium detection chamber. The
volume of the chamber was 32 times increased comparing with theprevious study. To increase the detection efficiency, 13 pieces ofplastic scintillators were placed parallelly. The 170 kBq/m3 oftritiated gas was produced by electrolyzing 82 kBq/L tritiated water
sample. The produced gas was measured for 3 mins1). The
maximum detection efficiency and MDA was evaluated as
43±4.3% and 7.0 kBq/m3 for tritiated hydrogen gas, or 3.4 kBq/L

Figure 1. Improved tritium detection chamber

as a concentration of the tritiated water sample.2) respectively. The MDA was 5.28 times lowered comparing
with previous study where the measurement time was 50 times shorter.
3. Conclusion
Tritium detection chamber was improved which was used in the electrolysis-based real-time tritium
monitoring system developed in previous study. The tritium detection chamber was newly designed, and it
showed better efficiency and lower MDA. Though the improved detection system does not satisfy an averagetritium concentration of north west Pacific Ocean (~0.1 Bq/L), it is expected to be applicable for regularmonitoring of release of contaminated water during decommissioning of the Fukushima Daiichi Nuclear Power
Plant.3)
References[1] Kumamoto, Yuichiro, et al. "Impact of Fukushima-derived radiocesium in the western North Pacific Ocean about ten
months after the Fukushima Dai-ichi nuclear power plant accident J. Environ. Radioact. 140, 114-122 (2015).

71

SK06

The Measurement and Comparation Analysis of Change ofRadioactivity Concentration of the Water Containing HTO
Molecule by Proton Exchange Membrane Electrolysis
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
The liquid tritium solution should be electrolyzed by proton exchange membrane (PEM) cell. And the change ofradioactivity concentration should be checked. The change before and after electrolysis can be detected by liquid
scintillator counter (LSC), quantulus 1220 from Parkin Elmer [1]. The change of liquid tritium mass should bealso measured by electronic scale. And then finally consumed radioactivity of liquid tritium can be calculated.
1. Introduction
In decommissioning sites of nuclear power plant, underwater tritium should be monitored and protected bysafety monitoring system. Existed detection method that need a long time for sampling, and treatment cannot
deal with realtime monitoring. However, if PEM cell is used in decommissioning sites of nuclear power plant andgets steady power supply, realtime onsite monitoring of underwater tritium can be detected.
2. Method and Result2-1. The Change of Radioactivity Concentration
LSC cocktails using liquid scintillator, liquid scintillator and the masses of the water containing HTO before andafter electrolysis are measured. The change of radioactivity before and after electrolysis is same with the numberof HTO molecules electrolyzed. The current condition of electrolysis is changed by 2 A, from 9 A to 5 A. For 45
minutes, radioactivity (Bq) of the water containing HTO from each current were changed from 535970, 476197,
455842 to 480810, 460586, 440269. The ratios of HTO molecules electrolyzed to total HTO molecules were 10.29%, 3.27 %, 3.42 %. These results can be used as reference value for the calibration of realtime detection bybetaionix monitor. And because the ratio of difference is not large, the results can be used to detect tritium only.
2-2. The comparation with PMT measurement
HT gases were moved as follow Figure 1. And it was also detected byusing plastic scintillator and photomultiplier. After detection chamber, HTgases are measured by beta ionix monitor. The radioactivity by betaionix
monitor was 25.4 Bq, 12.7 Bq, 11.8 Bq. The detection efficiencies were
35.01 %, 13.04 %, 6.43 %. It was a higher than the number of HTO
molecules electrolyzed per 1 second, 20.43 Bq, 7.49 Bq, 7.02 Bq.
However, the calculation values of detection efficiency by LSC, 43.56 %, Figure 1. The scheme of experimental
22.70 %, 10.82 % were higher than the calculation by betaionix.

process.

3. Conclusion
Direct measurement by betaionix is more incorrect than measurement by LSC. Because HT gas diffusion can beaffected by temperature or pressure. Measurement by LSC is more correct, however it also takes a long time formeasurement. If LSC is used, realtime detection will not be possible. Therefore, it is aimed to ensure the validityof the value of betaionix through repeated experiments and correction factors, so that it can be used in realtime
onsite detection.
References[1] 1220 QUANTULUS Ultra Low Level Liquid Scintillation Spectrometer, Perkin Elmer.
72

SK07

Fundamental Experiment of In-situ Beta Monitoring System for
Decommissioning Site
Min Ji Kim, Uk Jae Lee and Hee Reyoung Kim

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology
Abstract
Radiation monitoring is required to ensure the safety of decommissioning sites. In this study, beta
radionuclide 90Sr in water samples is monitored by using plastic scintillators and multiple photomultiplier tubes.
1. Introduction
The measurement system in which the source directly attaches to the plastic scintillator is proposed
because it is difficult to measure beta radionuclides in the field. The plastic scintillators do not react well withwater and gamma rays, but react well with the beta rays. The detection efficiency and MDA (Minimum
Detectable Activity) of 90Sr are calculated.
2. Method and Result2-1. Experimental System
Figure 1 shows the connection of experimental system.
This system uses coincidence counting technique withsimultaneous circuit in order to increase detection
Figure 1. Connection of systemefficiency.
The MDA is calculated as shown in the Eq (1).
𝑀𝑀𝑀𝑀𝑀𝑀𝑀𝑀𝑀𝑀𝑀𝑀 =2.71 + 4.65 × �𝑁𝑁𝑁𝑁𝑏𝑏𝑏𝑏 × 𝑇𝑇𝑇𝑇
(1)
𝑇𝑇𝑇𝑇 × 𝜖𝜖𝜖𝜖 × 1/100 × 𝑉𝑉𝑉𝑉𝑐𝑐𝑐𝑐

𝑁𝑁𝑁𝑁𝑏𝑏𝑏𝑏 is background count rate (cps), 𝜖𝜖𝜖𝜖 is efficiency (%), 𝑇𝑇𝑇𝑇 is count time (sec), and 𝑉𝑉𝑉𝑉𝑐𝑐𝑐𝑐 is volume of sample(cm3).

2-2. Results

Table 1. Detection Efficiency and MDA
Case

Activity(Bq)Efficiency(%)

MDA(Bq/g)

3.15

11.91816.305±1.234

0.014±0.0010

6.3

23.83719.327±1.511

0.011±0.0009

9.45

35.75517.854±1.328

0.013±0.0012

The experiments are conducted by adding 3.15 mlof90

Sr source whose activity is 3.7765 Bq/g to confirm

activity increases according to the amount of source. The
Table 1 shows result of 3 case of experiments. The
activities,

detection

efficiencies,

and

MDAs

are

calculated, respectively.
3. Conclusion
A system for the monitoring of beta radionuclide in water samples was proposed and MDA and detection
efficiency about 90Sr were estimated. Through this study, it is possible to develop a monitoring system for waterbeta radionuclide such as tritium using a coincidence counting technique.
References[1] Lee, UkJae, et al. "In situ beta radiation monitoring system with enhanced efficiency for water samples fromdecommissioned nuclear environment." Review of Scientific Instruments 90.2 (2019): 025103.

73

SK08

Pre-treatment System for In-situ Tritium Monitoringin Seawater near Decommissioning Site
Se Won Park and Hee Reyoung Kim

Ulsan National Institute of Science and Technology, Ulsan 689-798, Republic of Korea
Abstract
The Seawater pre-treatment system for in-situ tritium monitoring was developed, which can be applied todecommissioning sites. This pretreatment system consists of Micro Filtration filter (MF filter), Active Carbon
filter (AC filter), Reverse Osmosis membrane (RO membrane), ion-exchange resin, 0.2μm final filter etc. In thisstudy, the performance of seawater pretreatment system is verified using water quality standards.
1. Introduction
Tritium should be monitored strictly to ensure radiological safety, while its short range makes in-situ
monitoring difficult. Here, proton-exchange membrane (PEM) electrolysis method is considered, wherehydrogen ions act as an electrolysis medium. Therefore, pure water should be provided by an efficient seawater
pretreatment system.
2. Methods
2-1. Description of Seawater pretreatment System
The MF filter removes most impurities such as silica andbacteria, the AC filter removes organic matter, and the RO membrane
removes salinity of the seawater. The ion-exchange resin is used toenhance the purity of water, and 0.2μm final filter to remove microorganisms and particles once more [1].

Figure1 Design of Seawater
Pretreatment System

2-2. Water Quality analysis
To identify the purity of water generated from the system, pretreated water was analyzed based on water quality standards.
Table 1 Water Quality Standards

3. Results

For the results, conductivity is 0.080μS/cm-1 (ASTM TypeⅡ conductivity value is 1.0 μS/cm-1). Other
analysis items were analyzed as a detection limit, which is lower than the Type Ⅱ standard value.
4. Conclusion

The purity of seawater through the pre-treatment system can be predicted between Type Ⅰ (Ultra-Pure
Water) and Type Ⅱ. Therefore, seawater pre-treatment system provides proper water quality to be applied to thePEM electrolysis method.
References[1] Nikolay Voutchkov, “Considerations for selection of seawater filtration pretreatment system”, Desalination
261(3), October 2010, Pages 354-364.

74

SK09

Fit-Goodness Evaluation of Peak Shape Fitting Function for
Airborne Alpha Beta Detection System
Si Hyung Seong, Hee Reyoung Kim

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology
Abstract
An airborne alpha beta detection system using passivated implanted planar silicon (PIPS) detector was
proposed. Comparison of three fitting functions was performed to improve the accuracy of alpha beta spectrum
analysis of the detection system. The spectrum was simulated using Monte-Carlo simulation.
1. Introduction
The difficulty of alpha beta spectrum measurement in air should be considered because of alpha beta
particle energy loss easily due to high reactivity1. The application of various fitting functions should be
evaluated to good-fit the peaks of each nuclide through background subtraction in the spectrum2.
2. Materials and methods
2-1. Monte-Carlo simulation
The structure of the airborne detection system shown in figure 1
was simulated using the MCNP6 code.
2-2. Fitting functions

Figure 1. Diagram of detection system

The three fitting functions are applied to spectrum. Gaussian and Lorentz, peak shape fitting as shown Eq(1).
y(x) = 𝐺𝐺𝐺𝐺(𝑥𝑥𝑥𝑥) + 𝑇𝑇𝑇𝑇(𝑥𝑥𝑥𝑥), 𝑤𝑤𝑤𝑤𝑤𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒 𝐺𝐺𝐺𝐺(𝑥𝑥𝑥𝑥) = 𝐻𝐻𝐻𝐻𝑔𝑔𝑔𝑔 × 𝑒𝑒𝑒𝑒2�𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑔𝑔𝑔𝑔 �
2
2𝜎𝜎𝜎𝜎

−

𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑔𝑔𝑔𝑔, 𝑇𝑇𝑇𝑇(𝑥𝑥𝑥𝑥) = 𝐻𝐻𝐻𝐻𝑠𝑠𝑠𝑠 × 𝑒𝑒𝑒𝑒 𝜎𝜎𝜎𝜎×𝑡𝑡𝑡𝑡𝑠𝑠𝑠𝑠 × 𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒𝑒(𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑥𝑔𝑔𝑔𝑔√2𝜎𝜎𝜎𝜎+1√2𝑡𝑡𝑡𝑡𝑠𝑠𝑠𝑠)(1)

G(x) is gauss equation which describe peak of pulse, T(x) describe tailing of energy, 𝐻𝐻𝐻𝐻𝑔𝑔𝑔𝑔 is peak

amplitude, x is channel number in interest region, xg is peak channel, 𝐻𝐻𝐻𝐻𝑠𝑠𝑠𝑠 is peak tailing amplitude, 𝑡𝑡𝑡𝑡𝑠𝑠𝑠𝑠 is

extent of peak tailing.

3. Conclusion
Table 1 shows the comparison of the root mean square and the Chi-Squared test. As a result, it was
confirmed that the peak shape fitting function showed 3 ~ 6% higher fit-goodness compared to other fitting
functions. Since background removal and nuclide analysis are performed based on peak discrimination ability, it
is expected to improve the accuracy of nuclide monitoring when using peak shape fitting method.
Table 1. RMS and Reduced Chi-Squared Test
Fitting Functions

RMS

Chi-Squared Test

Gaussian

0.9652

6,79E-9

Lorentz

0.9382

4.53E-8

Peak Shape

0.9965

3.42E-10

References[1] 1. E. Garcia-Torano (2006) “Current status of alpha-particle spectrometry”, Applied Radiation Isotope, 64, 1273-1280[2] Bortels. G (1987) “Analytical function for fitting peaks in alpha-particle spectra from Si detectors”, IJRAI P38, 831-837

75

US01

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

Abstract
EnergySolutions (ES) is a leading international Decommissioning (D&D) company with a rich history indecommissioning and waste management, both in the US and internationally, stretching back over 30 years.
1. Successful Project Performance
Worldwide, only 15 commercial nuclear power plants have completed D&D, 12 of these were in the US. ES
managed, or played a major role in all US projects. Since the international decommissioning industry is relatively
new, successful performance requires critical evaluation of past performance and adoption of lessons learned. By
applying this practice, ES is currently
completing decommissioning of two 1,090
MWe PWR reactors (Zion) for less than $1B
and in just over 8 years.
2. Project Success through Integration
Successful D&D projects need clear understanding and definition of the scope, strong disciplined program
management, and extremely active stakeholder engagement and participation.
a. Fully Integrated Project Scope, Cost, Schedule and Risk
There is a major difference between D&D projects andtraditional predictable projects. D&D projects carry significantly
more risk. Traditional projects focus on scope, cost and schedule,
the so-called “three-headed dragon”, but in less predictableprojects, risk is a major component of the project management
model. By their nature D&D projects are dependent on thelimited industry experience to develop the baseline and are
subject to significantly more uncertainty and unpredictability. ES
has developed its own approach and tools necessary to manage
decommissioning activities in this higher risk environment, andthe success of this approach has been demonstrated on the Zion project.
b. Integration of Experience, Project Management and Local Capabilities
A major component of reducing D&D project risk is the implementation of lessons
learned from prior D&D projects. International experience has demonstrated
inexperienced companies significantly exceed budget, sometimes as much as 400%.
c. Integration of all stakeholder
Stakeholder participation and support is critical to success. Poor relations
and badly formulated stakeholder management plans will cause extreme problems
in the decommissioning project and potentially result in major cost escalation andschedule slippage. Early, open, honest and continuous communication is essential.

76

US02

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

Abstract
Key to success in decommissioning is an effective radioactive waste management strategy. EnergySolutions(ES) has more than 35 years’ experience developing and executing nuclear waste management, treatment, andwaste volume minimization processes to address the challenges of optimizing waste strategies to meet different
national and international disposition criteria.
1. Full Range of Waste Treatment Capabilities
In the past, the US had the same issue as the rest of the world high cost waste disposal and limited available
volume. To meet this challenge, ES developed comprehensive waste management, treatment and volume
minimization techniques. These included the full range of mechanical, chemical and thermal techniques including;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.
⚫

Processes can be Local or Centralized, Permanent or Temporary (Modular or Mobile)⚫

Effective Waste Strategy requires Successful Integration of Multiple Processes

2. Recycle of Major Component of Radioactive Waste for Volume Minimization
ES has been recycling waste materials from the nuclear industry for over 30 years, including over 25 years’
experience in international metal recycling. ES takes ownership of themetal, produces products used in nuclear licensed facilities, and no waste
is returned to customer. JPARK (Tokai) uses 2500 tons of shield blocks
made of recycled metal.

4. Conclusion
As D&D progresses in Japan, the need for new and effective waste management practices will be essential.
⚫

Radioactive Waste Management, a New Opportunity for Communities⚫

EnergySolutions Approach – Partner with Local Communities and Industry

77

US03

Horizontal Storage S
System for Used Nuclear Fuels,, Fuel Debrris andNuclear wastesw-N
NUHOMS® MATRIX
X system Ryan BUCK
K(i) , Jane HE
E(i) , Prakash NARAYAN
NAN(i)(ii)(ii)
Vincent TRA
AN , Hirosshi AKAMATTSU , Yasuo YAMAD
DA(ii)(i) TN Amerricas, LLC (ii) Transnuclear, Ltd

Abstractt
The NUHHOMS® MAT
TRIX Horizon
ntal Storage M
Module (HSM)) is a two-tiered concrete sttorage overpaack systemfor dry shhielded canisteers (DSC) to storesused nucclear fuels, fu
uel debris, andd nuclear waste
tes. It is an evolutionary
design off the existing NUHOMS® HSM, MATR
RIX is characcterized by itss small footprrint, storage efficiency,eseismic rresistance, andd self-shieldin
ng. Combineed with the TMI-2
T
fuel deebris canister design know
wledge, the
NUHOMMS® Matrix is ana optimal and cost-effectivve solution for the Fukushim
ma fuel debriss and used nucclear fuel.
HOMS® expeerience with fuelfdebrisThe NUHThe only example of dry
d storage off fuel debris toook place in the USA for th
he TMI-2 reacctor. Debris iss stored atL) within thee NUHOMS® system whichh consists of a metal canister and a
the Idahoo National Laaboratory (INL
horizontaal concrete stoorage overpack
k. Apart from this technology, all other storage methodd would be a FirstF
Of A
Kind (FOOAK) achievvement. The NUHOMS® horizontal concretecstorrage system went throug
gh several
developm
ments and its performancep
wasw significanntly enhanced
d since its ado
option for TM
MI-2 melted fu
uel debris.
For instannce, the use ofo a dual level structure reduuced considerrably the footp
print compareed to the systeem used in
TMI (45%% reduction).
HOMS® expeerience with usedufuel
The NUH
Orano TNN’s NUHOMS
S® systems haave securely sttored used nucclear fuel in th
he United Stattes for more th
han two
decades, with installatiions at 33 sites around the ccountry repressenting more thant
45,000 sttored used fueel
OMS® MATR
RIX HSM systtem is schedulled for approv
val by US NRC
C in 2019 andd wasassembliees. The NUHO
already seelected by a US
U Utility, Wo
olf Creek Nucclear Operatingg Company (WWCNOC), forr loading in 20
020.
HOMS® expeerience with wastesw
The NUH®
The NUHHOMS® Matriix System is alsoa capable oof storing the NUHOMS
N
RadwasteR
Cannister (RWC), a canisterthat is used to store dryy irradiated annd/or contamiinated nonfueel hardware. The
T NUHOM
MS® system dry
y shielded
canisters can be adapted internally to safely storee many forms of nuclear waaste (intact ussed fuel to enccapsulated
a of which ccould be stored in the NUH
HOMS® MAT
TRIX HSM arrray. This
fuel debris to radioactive wastes), allmmissioning process;p
its flexiblefdesignn makes it a universal
technology can be applied for the whole decom
overpackk that can meett nuclear operrators’ needs aand requiremeents.

Conclusiion
More thaan 1,200 systeems were load
ded in the US
SA, all associiated hardwarre, auxiliary ssystems, proceedures forloading/uunloading, trannsfer, and mon
nitoring are w
well developedd. The small footprintfand mmultiple uses makes the
NUHOMMS® MATRIX perfectly suitted for decomm
mmissioning acctivities.

Loading operation in h orizontal

Dual level horizo
ontal concrete module
78

US04

DBE PLUS+
Innovative Approach to Decommissioning
Takehiko Miyazaki, Bechtel1

Abstract: This poster illustrates Bechtel’s new approach to developing site-specific Decommissioning Baseline
Estimates using over 20 years of historical decommissioning results and site-specific data. A feature of thisapproach takes the 3D site-specific model developed from historical waste and construction quantities andintegrates schedule/cost to create 4D/5D modeling solutions that demonstrates D&D sequencing, cost, schedules,
waste generation, and manhours.

Screenshots from Bechtel’s DBE PLUS+ 4D/5D modeling solution.

Introduction: Bechtel’s Decommissioning Baseline Estimate methodology, DBE PLUS+, is an advanced
schedule and estimating baseline tool developed to bring site-specific D&D solutions to our customers. DBE
PLUS+ leverages more than 100 years of project execution history and decommissioning baseline estimating
expertise. It draws upon the experiences and historical data across the nuclear lifecycle of commercial andgovernment customers—as well as the latest data from current nuclear new build and decommissioning projects
such as Vogtle Units 3 & 4, Sellafield Pile Fuel Cladding Silo, and Savannah River Site. Bechtel Subject Matter
Experts (SMEs) analyze each aspect of a nuclear power plant including the following:•

End State Requirements•

Waste Management Requirements•

Large Component Removal•

Risk Management•

Reactor/Internals Segmentation•

Site Security Requirements•

Cold & Dark Requirements•

Regulatory Requirements•

Spent Fuel Requirements•

Work Force Requirements

Then the Bechtel SME’s integrate cost and schedule into a site-specific, executable plan that creates a timelinebased output of:•Generated Waste Quantities•

Staffing Plans•

Equipment Requirements•

Stakeholder Management Plans•

Work Breakdown Structure•

Risk Management Plans•

Waste Management Costs•

Labor Hours Projected and Expended

Conclusion: Bechtel’s DBE PLUS+ utilizes actual D&D results from completed projects and lessons learned tocreate an accurate execution plan for managing and optimizing project resources, schedules, and cost. The
3D/4D/5D outputs provide an effective method to demonstrate progress and D&D sequencing to all stakeholders.
References: [1] Bechtel is a trusted engineering, construction and project management partner to industry and government.
Since 1898, we have helped customers complete more than 25,000 projects in 160 countries on all seven continents.

79

US05

Application of UK Waste Management Performance at the
Fukushima Daiichi Decommissioning
Andy Beckwith, Max Ehrhardt
AECOM

Abstract
The United Kingdom’s radioactive waste management approach minimizes the volume of waste that
requires final disposal in their limited available space. As Fukushima Daiichi and other domestic
decommissioning projects face similar waste management challenges, application of this approach to radioactive
waste in Japan can replicate the successful transformation in decommissioning seen in the UK.
1. Introduction
Waste management is a significant challenge at Fukushima Daiichi both in stakeholder concerns, logisticalchallenges, and long term costs. From a safety point of view, the higher activity waste presents the most acuterisk, but the volume of Level 2, Level 3, and Clearance Waste presents a long term drain on resources. Thissituation is similar to the UK, in that Japan has a large source of radioactive waste without authorized disposal
locations. In response to this limitation, the radioactive waste management approach used in the UK is tominimize the volume of waste that requires disposal at their only radioactive disposal sites: Low Level Waste
Repository (LLWR) and Dounreay’s Low Level Waste Vaults.
2. United Kingdom Waste Management Approach
2-1. 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, the UK National Waste Program has been developed andexecuted to accomplish the goals of the Policy. Before the UK National Waste Program was developed, it was
projected that the LLWR site’s disposal capacity would be fully exhausted by 2023. As a result of the successful
implementation and execution of the UK’s National Waste Management Program, the LLWR site’s usefulcapacity and life has been extended by one hundred years.
2-2. Waste-Informed Decommissioning
The success of the UK’s LLW Policy has resulted in an 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. The “Waste-Informed Decommissioning Process” ensures that
limiting disposal waste volume is a key objective of the decommissioning strategy to provide the best overallsolution for the country.
2-3. Diverse Technologies for Waste Processing
The National Waste Program (NWP) operates through individual services contracts between the LLWR
and all UK waste generators. The waste management approach is integrated into the entire Supply Chain of thewaste generators which allows cross-utilization of services and significantly minimizes the cost for thefollowing Supply Chain elements.
• Safety Case Analysis• Waste Characterization• Transportation• Packaging• Metals Processing• Combustible Processing• Very Low Level and Low Level Waste Disposal• Waste Tracking Systems
2-4. Applying the Waste Hierarchy
Avoidance of Radioactive Waste generation is the best option as it has extended the life of the LLWR site
and greatly reduced the overall nuclear liability from generation through disposal. Avoidance has best beenachieved by dispatching waste management experts to work directly with the decommissioning sites to ensurethat correct Decommissioning processes are applied to reduce generation of as much radioactive waste as
possible. Other methods used to minimize disposal at LLWR include Metal Size Reduction, Decontamination
via Shot Blasting, Metal Melting, Incineration of Combustible Waste, and Shredding and Compaction.
3. Application to Fukushima Daiichi
The method used at the UK was a combination of government action (waste hierarchy), a good contracting
model for the generators and repository, and creation of alternative methods for disposal of the waste. At
Fukushima, incineration already exists and is being expanded. However, inclusion of metal recycling andsupercompaction, application of a “Very Low Level Waste” arrangement, central transport capability, standard
radioactive characterization, and low level liquid processing, all managed through a single group incentivized toreduce the volume of low level waste could replicate the performance of the UK over the next 10 years.

80List of Organizations
Organization

ID No.

ABLE Co. Ltd.

E04

ADVAN ENG. Co., Ltd.

H01

AECOM

US05

AESJ (Atomic Energy Society of Japan)

A01, K03

Aizuk, Inc.

B04

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

F07

Ascend Inc.

B03

ATKINS

GB04

BCSN

F08

BechtelUS04

Cavendish Nuclear Ltd.

GB02

Cavendish Nuclear Japan

GB02

CEA (French Alternative Energies and Atomic Energy

C02, F01, F02, F07

Commission)
Cegelec CEM

F10

CORNES Technologies Limited

B10

CRIEPI (Central Research Institute of Electric Power

C02, C04, K02

Industry)
Daihatsu Motor Co., Ltd.

H02, H03

East Japan Accounting Center Co. Ltd.

B03

Ecole Centrale Supelec

F06

e-Energy Corporation

F08, F09, F10

EnergySolutions

US01, US02

Forschungszentrum Jülich

H02

Fuji Electric Co., Ltd.

K04

Fukushima Consortium of Robotics Research for

B01

Decommissioning and Disaster Response
Fukushima Midori Anzen Inc.

B08

Gleeds Energy Limited

GB05

Gunma University

C02

Hamamatsu Photonics K.K.

B09

Hitachi, Ltd.

K02

Hitachi-GE Nuclear Energy, Ltd.

K02

Imagineeing Inc.

J12

81

IRID (International Research Institute for Nuclear
Decommissioning )IRSN(InstitutdeB05, B06, C05, E03, J06,
J11, J13

RadioprotectionetdeSûreté

F01

Nucléaire)
JAEA (Japan Atomic Energy Agency)

A01, C01, C03, C05, E03,
H02, H03, J01, J02, J03,
J04, J05, J06, J07, J08, J09,
J10, J11, J12, J13, J14, F02

JAMSTEC (Japan Agency for Marine-Earth Science and

C04

Technology)
JAXA (Japan Aerospace Exploration Agency)

C01

JGC Corporation

K02

Kajima Corporation

E05

Kobe University

C02

Kurion Japan K.K

A03, A04, A05

Kwansei Gakuin University

H02, H03

Kyoto University

C01, C04

Mitsubishi Research Institute, Inc.

A02

MITSUFUJI Corporation

A06

MKC Consulting

US01, US02

Nagaoka University of Technology

H01

Nagoya University

A01, C03

NDF(Nuclear

DamageCompensation

and

A01

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

C05, E03

NIT (National Institute of Technology), Fukushima

B03

College
NIT (National Institute of Technology), Kisarazu College

C01

NINS (National Institutes of Natural Sciences)

J12

Nuclear water soil solutions

D01

ONET Technologies

F01

Orano

F07

Orano Cycle

F05

Orano Dismantling & Services

F06

Orano GmbH

F06

Orano Group

F03, F04

Orano Japan

F05, F06

Orano Temis

F05

82

Osaka Prefecture University

C01, J03

Osaka University

A01

QST (National Institutes for Quantum and RadiologicalJ03, J12

Science and Technology)
RIKEN

C01

ROBATEL Industries

F09

Sellafield Ltd.

GB01

Shanghai Jiao tong University

D03

SSC RIAR (Research Institute of Atomic Reactors) JSC

C06

Taisei Corporation

E02

Takawa Seimitsu Co. Ltd.

B03

TENEX JSCC06

TEPCO (Tokyo Electric Power Company Holdings, Inc.)

E01, E02, E03

The University of Tokyo

C03, C04, J03

Tigers Polymer Corporation

B07

TN Americas, LLC

US03

Tohoku Institute of Technology

C03

Tohoku University

C02, C03, D02, D03, J03

Tokyo Bosai Setsubi Co., Ltd.

B02

Tokyo City University

D01

Tokyo Institute of Technology

J03

Toyo Union Co., Ltd.

E02

Transnuclear, Ltd.

US03

UNION SHOWA K.K.

D02, D03

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

Technology)

SK05, SK06, SK07, SK08,
SK09

Université de Lyon

C02

Veolia Nuclear Solutions, Inc.

A04, A05

Veolia Nuclear Solutions (UK) Ltd.

A04

VNS Federal Services Inc.

A05

Waseda University

J14

WM Symposia, Inc.

K01

Wood

K04

Wood plc

GB03

83