ABSTRACTS of the Technical Poster Session

The 2nd International Forum on the Decommissioning of
the Fukushima Daiichi Nuclear Power Station-Moving forward together-

Mon, July 3, 2017
The Iwaki Lifelong Learning Plaza in Iwaki-city,
Fukushima-prefecture, Japan

Nuclear Damage Compensation andDecommissioning Facilitation Corporation (NDF)

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

Contents
Session A: Internal PCV Situation Analysis, Fuel Debris Characterization and PCV Internal
SurveyA01

Measurement of Lower Part of PCV of Hamaoka NPP Unit 2 by Transmission
Type Muon Radiography Using Nuclear Emulsion
Kunihiro Morishima1, Mitsuhiro Nakamura1, Kenji Tsuji2 andMasataka Ooyama2
1: Nagoya University, 2: CHUBU Electric Power Co., Inc.

A02

Technology Development to Predict Dose Rate Distribution in PCV of Fukushima
Daiichi Nuclear Power Station
Eka Sapta Riyana and Keisuke Okumura
Japan Atomic Energy Agency (JAEA)

A03

Investigation inside the Primary Containment Vessel (PCV)
Yoichi Murai1, Satoshi Okada1, 2 and Atsuyoshi Uranishi1, 3
1: International Research Institute for Nuclear Decommissioning (IRID),
2: Hitachi GE nuclear Energy, Ltd, 3: Toshiba Corporation

A04

OECD/NEA International Joint Research Project regarding Fuel Debris
Characterization PreADES& TCOFF
Akira Nakayoshi, Tadahiro Washiya, Masaki Kurata, Kimihiko Yano,
Yuji Nagae, Masahiko Osaka and Hideki Ogino
Japan Atomic Energy Agency (JAEA)

A05

Development of Ultrasonic Underwater Imaging Method for Decommissioning ofFukushima Daiichi Nuclear Power Station
Ryosuke Kobayashi, Hayato Mori, Takahiro Nagai and Naoyuki Kono
Hitachi Ltd., Research & Development Group

Session B: Remote Technology and Radiation Tolerant Equipment
B01

3-D Sensing Technologies of Radiation and Environmental Structure at the
Fukushima Daiichi NPS
Yuta Terasaka, Yuta Tanifuji, Masaaki Kaburagi, Yuki Sato,
Hiroshi Usami, Kuniaki Kawabata, Hiroko Nakamura Miyamura andTatsuo Torii
Japan Atomic Energy Agency (JAEA)

B02

Development of a Small and Modular UROV for Surveying Radioactive Materials
on the Lakebed
Mizuki Tanaki, Kanno Ryo, Luis Canete and Takayuki Takahashi
Fukushima University

B03

Decommissioning Operations Planning Using Proven TechnologiesiYasushi Kikkawa, Daphne Ogawa and Bastien de Lazzari
AREVA ATOX D&D SOLUTIONS Co. Ltd.
B04

Radiation Tolerant Camera by Mirion Technologies
The World’s Highest Radiation Tolerant Performance Camera
Ken Takahashi and Tadashi Sato
Cornes Technologies Limited

B05

3D Information Technology Based Support for Nuclear Decommissioning
István Szőke
Institute for Energy Technology / OECD Halden Reactor Project

B06

Activities

of

the

Fukushima

ConsortiumofRobotics

ResearchforDecommissioning and Disaster Response
Yoshiro Owadano
Fukushima Consortium of Robotics Research for Decommissioning andDisaster Response
B07

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

Session C: Analysis Technologies
C01

Mirion Technologies (Canberra) Experience and Success for Efficiency and Safety
Decontamination and Decommissioning
Atsuo Suzuki, Lou Sai Leong and Helene Lefebvre
Mirion Technologies (Canberra) KK

C02

Active/Passive Neutron/Gamma Assay Systems for Fuel Debris and Nuclear
Waste Activity and Isotopic Mass Determination
Lou Sai Leong, Atsuo Suzuki and Helene Lefebvre
Mirion Technologies (Canberra) KK

C03Development of Simple and Rapid Analytical Methods for Sr-90 in theContaminated and Treated Water in the Fukushima Daiichi Nuclear Power
StationTomomune Matsunaga1, Takuya Nakagawa1, Takuyo Yasumatsu1,
Hirofumi Tazoe2 and Masatoshi Yamada2
1: Tokyo Power Technology Ltd., 2: Hirosaki University

C04

ICP Mass-Spectrometric Analysis of Radioactive Strontium without Calibration
Curve
Mizuki Odashima1, Makoto Furukawa1,2,3, Katz Suzuki4 andYoshitaka Takagai1,5ii

1: Fukushima University, 2: The University of Tokyo, 3: Perkin Elmer,
4: Japan Agency for Marine-earth Science and Technology (JAMSTEC),
5: IER Fukushima University
C05

Development of Quantitative Method for Strontium-90 Using Thermal Ionization
Mass-Spectrometry
Chihiro Ito1, Takashi Miyazaki2, Shigeyuki Wakaki2, Katz Suzuki2 andYoshitaka Takagai1, 3
1: Fukushima University, 2: Japan Agency for Marine-earth Science andTechnology (JAMSTEC), 3: IER Fukushima University

C06

Simultaneous Quantification Method for Radioactive Strontium and Radioactive
Iodine in ICP-MS
Hiroaki Ogata1, Makoto Furukawa1,2,3 and Yoshitaka Takagai1,4
1: Fukushima University, 2: The University of Tokyo, 3: Perkin Elmer,
4: IER Fukushima University

C07

Rapid Detection of Radioactive Sr Using Inexpensive Reagent and a Simple-MassSpectroscopy
Kaori Chiba1 and Masaki Koike2
1: National Institute of Technology, Ibaraki College,
2: Hiroshima Institute of Technology

C08

Laser Remote Analysis for In-core and In-situ Monitoring of Debris in
Decommissioning of Fukushima Daiichi Nuclear Power Station
Ikuo Wakaida, Hironori Ohba, Katsuaki Akaoka, Masaki Ohba,
Ayumu Matsumoto and Masabumi Miyabe
Japan Atomic Energy Agency (JAEA)

Session D: R&D for Materials
D01

Development of the Insoluble Neutron Absorbers Using Self-Assembled
Surfactant Aggregates
Noriaki Ushio, Kouji Koyanagi and Hiroyuki Terazaki
Kao Corporation

D02

Corrosion Inhibition of Carbon Steel Piping by Addition of Sodium Nitrite andSodium Pentaborate
Kazutoshi Fujiwara1, Junichi Tani1, Joji Ohta1, Yuichi Fukaya2 andToshifumi Hirasaki2
1: Central Research Institute of Electric Power Industry (CRIEPI),
2: Tokyo Electric Power Company Holdings, Inc.

D03

Fundamental Study for Maintaining Integrity of Cooling Water Circulation
System: Experimental Examination of Carbon Steel Corrosion under Flow
Kazuki Kushida, Shohei Sato, Hiroshi Abe and Yutaka Watanabeiii

Graduate School of Engineering, Tohoku University
Session E: Fuel Removal from Spent Fuel Pool
E01

SFP Fuel Removal at Fukushima-Daiichi NPS
Ritsuro Tokumori, Masashi Fukushima and Akimasa Miyamoto
Tokyo Electric Power Company Holdings, Inc.

E02

Unit-3 Covering Structure Plan for Spent Fuel Removal and Work Progress
Shinya Okada, Kenichiro Go, Takashi Inoue, Norio Ryoki, Taimei Hida,
Atsushi Kikuchi, Kazuhiro Kato, Miho Miyazaki, Ippei Matsuo andKihei Ogawa
Kajima Corporation

Session H: Fuel Debris Retrieval Method
H01

Study for the Bottom Access Route to Remove the Debris from the Nuclear
Reactor Buildings
Atsuo Yamada1, Hisashi Imai1, Masanori Imazu2 and Toshihiko Fukuda2
1: Hazama Ando Corporation, 2: Nuclear Damage Compensation andDecommissioning Facilitation Corporation

H02

A Proposal of the New Method to Retrieve Fuel Debris with Geopolymer
Taichi Sakai, Takayuki Saimu, Shunichi Suzuki and Koji Okamoto
The University of Tokyo

Session J: Decontamination Technology
J01

Practical Uses of the CW Single-Mode Fiber Laser Decontamination Device
Technology
Eisuke J. Minehara
Laser Decontamination and Decommissioning (LDD) Corporation

J02

Direct Decontamination Technique by Fiber Laser
Ryuichiro Yamagishi
The Wakasa Wan Energy Research Center

J03

Sorting of Contaminated Soil Based on Radiological Separation-Experiences witha High Performance Mobile Belt Conveyor SystemF. Langer and M. Sokcic-Kostic
NUKEM Technologies Engineering Services GmbH

Session K: Technologies for Contaminated Water Treatment and Frozen-soil Walls
K01

Development of Polymeric Aluminosilicate Agent for Cation Removal
Takenori Masada1,2, Masahiko Matsukata2 and Atsushi Yamazaki2
1: Azmec Inc., 2: Waseda University
iv

K02

Environmental Risk Analysis for Radionuclides Transported in Groundwater ofFukushima Daiichi NPS
T. Koyama1, Y. Tanaka1, D. Tsumune1, T. Hijikata1, Y. Tateda1,
T. Hattori1, M. Imazu2, M. Kamoshida2, A. Hasunuma2, K. Hida2, A. Ono2
and T. Fukuda2
1: Central Research Institute of Electric Power Industry (CRIEPI),
2: Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF)

K03

Continuous Challenges to Improve Performance, Availability and Operational
Cost of Contaminated Water Treatment Systems for Fukushima Daiichi
Teruki Fukumatsu1, Masaru Nishikubo1, Masahiko Kobayashi1,
Keiji Ishikawa2 and Kei Kobayashi2
1: Toshiba Corporation, 2: Tokyo Electric Power Company Holdings, Inc.

K04

Development of New Radioisotopes Selective Adsorption Media in EBARA
Takashi Sakuma, Makoto Komatsu, Tatsuya Deguchi, Kaoru Kikuchi andTakeshi Izumii
EBARA Corporation

K05

Development of Decontamination Technologies for the Contaminated Water inFukushima Daiichi Nuclear Power Station
Takatoshi Hijikata, Koichi Uozumi, Kenta Inagaki, Takeshi Tsukada andTadafumi Koyama
Central Research Institute of Electric Power Industry (CRIEPI)

K06

Installation of Land-side Impermeable Wall
–Frozen-soil Walls at the Fukushima Daiichi Nuclear Power Station
Hiromitsu Kida, Atsuhiro Fukada, Isao Abe, Taichi Esaki, Shoji Seno andTeru Yoshida
Kajima Corporation

K07

Development of Micro-Polymeric Remover for Uranium Ion
Miki Abe1, Michio Butsugan2, Shukuro Igarashi3 and Yoshitaka Takagai1
1: Fukushima University, 2: Hitachi Chemical Techno Service Co., Ltd.,
3: Ibaraki University

K08

Purification of accumulated waste water and groundwater in Fukushima Dai-ichi
Nuclear Power Station
Yusuke Kitamoto1, Takashi Nishi1, Takashi Asano1, Kenji Noshita1,
Shunsuke Miyake1, Masahiro Ichino1, Yuko Kani2, Toshimasa Ohashi2,
Yutaka Sangu2, Ken Yamaguchi3, Kei Kobayashi3, Yuichi Kurosaki3 andKeiji Ishikawa3
1: Hitachi-GE Nuclear Energy, Ltd., 2: Hitachi, Ltd.,
3: Tokyo Electric Power Company Holdings, Inc.
v

K09

Fabrication of Porous Ceramics Using Natural Minerals for Purification ofRadioactive Water
Shuhei Toike1, Katsumi Yoshida1, Toyohiko Yano1 and Tadahiko Takada2
1: Tokyo Institute of Technology, 2: Silica Material LLC

Session L: Radioactive Waste Management
L01

SIAL®: Geopolymer Solidification Technology Approved by Czech / Slovak Nuclear
Authority
Milena Prazska1, Marcela Blazsekova1 and Hisashi Mikami2
1: Amec Foster Wheeler Nuclear Slovakia s.r.o., 2: Fuji Electric Co., Ltd.

L02

Volume Reduction and Recycling of Contaminated Metal by Melting
Satoshi Karigome1, Masaru Kawamata1, Takahiro Matsuda1,
Toru Kikuchi2 and Hiroki Shimoda2
1: The Japan Atomic Power Company (JAPC),
2: JFE Engineering Corporation (JFEE)

L03

Current Status and Issues about Secondary Waste from Water Treatment
Kosuke Sato
Tokyo Electric Power Company Holdings, Inc.

L04

Examination of Slurry Stabilization Technology
Toshio Nakazawa1, Kohhei Aikawa2
1: Tokai Office, 2: Technical Development Center ATOX CO. Ltd

L05

High-frequency Melting Technology of Radioactive Metal Waste
Takeshi Nishikawa, Nobutake Horiuchi, Naoto Sasaki and
Hirokazu Tanaka
Mitsubishi Materials Corporation

L06

New Radioactive Waste Including Slurry Treatment Systems by Compact Disc
Dryer
Hiroyuki Hayashi, Hirofumi Inagawa, Takashi Otsu, Takashi Sakuma,
and Takeshi Izumii
EBARA Corporation

L07

Research & Development of Radioactive Waste Long-Term Storage Containers
with Passive Autocatalytic Recombiners
Kazuyuki Takase1, Nguyen Huy Hiep1, Yota Suzuki1, Yusei Tanak1,
Isamu Kudo2 and Ryutaro Hino3
1: Nagaoka University of Technology, 2: ADVAN ENG. Co., Ltd.,
3: Japan Atomic Energy Agency (JAEA)

L08

Study on Post-accident Waste Management Scenarios for Fukushima Daiichi
Nuclear Power Plant
Subcommittee on Radioactive Waste Management, Study Committee onviDecommissioning of the Fukushima Daiichi NPP
Atomic Energy Society of Japan (AESJ)L09

A Pyrochemical Treatment Concept of Combinding Selective Fluorination and
Molten Salt Electrolysis for Fuel Debris
Haruaki Matsuura1, Takafumi Uchiyama1, Takahiro Ono1,
Masatoshi Kajimoto1, Yu Sakane1, Tadayoshi Sato1, Tatsuya Emori1,
Atsushi Nezu2 and Nobuaki Sato3
1: Tokyo City University, 2: Tokyo Institute of Technology,
3: Tohoku University

L10

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

Session M: Human Resource Development
M01

Human Resource Development by Creative Robot Contest for Decommissioning
Shigekazu Suzuki
National Institute of Technology, Fukushima College

M02

Advanced Research and Education Program for Nuclear Decommissioning
Toru Obara1, Toyohiko Yano1, Kenji Takeshita1, Yukitaka Kato1,
Hiroshi Akatsuka1, Hiroshige Kikura1, Takehiko Tsukahara1,
Katsumi Yoshita1, Koichiro Takao1, Jun Nishiyama1, Masayuki Harada1,
Nobuyuki Iwatsuki1, Koichi Suzumori1, Gen Endo1, Kenji Kawashima2,
Noriko Asanuma3, Tsuyoshi Arai4 and Naoyuki Takaki5
1: Tokyo Institute of Technology, 2: Tokyo Medical and Dental University,
3: Tokai University, 4: Shibaura Institute of Technology,
5: Tokyo City University

Session N: Several R&D, Technologies and Activities for Decommissioning
N01

Current Status of the JAEA’s R&D Bases for the Decommissioning of the Tokyo
Electric Power Company Holdings, Inc. Fukushima Daiichi Nuclear Power
StationYoshihiro Tsuchida, Makoto Ooka, Toshiro Nakai and Shinichi Nakayama
Japan Atomic Energy Agency (JAEA)

N02

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

N03

Sorting Technology for Efficient and Safe Handling and Storage
Steve Rima1, Jeffrey Lively1, Masaru Noda2 and Haru Hashizume2
1: Amec Foster Wheeler, 2: Obayashi Corporation
vii

N04

Activities for Fuel Cooling at the Fukushima Daiichi Nuclear Power Station
Mitsuyuki Yamazaki
Tokyo Electric Power Company Holdings, Inc.

N05

The Overview of Project of Decommissioning and Contaminated Water
Management
Masaaki Matsumoto, Naoki Kondo, Takaki Ashida and Masaki Kawai
Mitsubishi Research Institute, Inc.

N06

Power Assist Suit (PAS) to Support Heavy Labor in Radiological Environment
Hirohisa Hirai and Yoshihiro Tamura
Mitsubishi Heavy Industries, Ltd.

N07

Veolia Nuclear Solutions Technologies for TEPCO 1F Decommissioning
Mamoru Numata, Syuichi Ueno and Sadaaki Abeta
Kurion Japan K.K.

N08

Contribution to Environmental Remediation in Fukushima
Takao Ikeda and Wataru Kobayashi
JGC Corporation

N09

Works for Destroyed ChNPP Unit 4 Conversion into an Ecologically Safe System
as an Experience in Mitigation of Fukushima Daiichi NPP Accident Aftermath
Viktor Krasnov
Institute for Safety Problems of Nuclear Power Plants National Academyof Sciences of Ukraine (ISP NPP)

Session F: France
F01

Development of Laser Cutting and Dust & Fumes Collection Technologies for Fuel
Debris Retrieval
D. Roulet1, C. Georges2, C. Chagnot2, C. Journeau2 and E. Porcheron3
1: ONET Technologies, 2: CEA, 3: IRSN

F02

Chemical-Free Water Treatment by Hydrodynamic Cavitation (HDC) andElectrocoagulation with Sacrificial Ferrous Anodes
Gilbert Peronnet1, Tristan Cerosky2 and Christian HONNORAT2
1: Optimum Water System, 2: NUVIA France

F03

Hydrodynamic Cavitation (HDC) Application to Existing Reverse Osmosis
Systems
Gilbert Peronnet1, Tristan Cerosky2 and Christian HONNORAT2
1: Optimum Water System, 2: NUVIA France

F04

Froth Flotation Process for Remediation of Contaminated Soil
S. Faure1, A. Ben Said1, A. Grandjean1, F. Charton1, C. Georges1,
T. Prévost2, D. Grange3, J.P. Spanu3 and F. Maneglia4
1: CEA, 2: AREVA NC, 3: VEOLIA, 4: ANADEC
viii

F05

IN CAN MELTER for the Conditioning of High and Medium-level and Long-lived
Waste from Decommissioning Operations
Christophe GIROLD1, Christine GEORGES1, Stéphane CATHERIN2,
Thierry PREVOST3 and Guillaume LECOMTE4
1: CEA, 2: National Radioactive Waste Management Agency (ANDRA),
3: AREVA NC, 4: ECM Technologies

F06

Potentialities Offered by New Binders in Conditioning of Problematic Low- andIntermediate-level Radioactive Waste
David Lambertin, Christine Georges, Fabien Frizon and
Céline Cau dit Coumes
CEA

Session GB： Great Britain
GB01

UK and Japan - Natural Partners in Nuclear Decommissioning
1. UK-Japan Collaboration
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB02

UK and Japan - Natural Partners in Nuclear Decommissioning
2. Sellafield and Fukushima Daiichi - the challenges of stakeholder engagement
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB03

UK and Japan - Natural Partners in Nuclear Decommissioning
3. Waste Retrieval – Magnox Reactor Sites and Dounreay
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,ix10: Amec Foster Wheeler, 11: International Nuclear Services
GB04

UK and Japan - Natural Partners in Nuclear Decommissioning
4. Waste Retrieval – Sellafield
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB05

UK and Japan - Natural Partners in Nuclear Decommissioning
5. Programme Management
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB06

UK and Japan - Natural Partners in Nuclear Decommissioning
6. Decommissioning Technology
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB07

UK and Japan - Natural Partners in Nuclear Decommissioning
7. Needs Based Research and Development
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services

GB08

UK and Japan - Natural Partners in Nuclear Decommissioning
8. Waste ManagementxKeith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5,John Falch6, Matt Mellor7, Chris Carter8, Linda Ashton9, Steve Caley10,
Sarah Groves11
1: British Embassy Tokyo, 2: Sellafield Ltd., 3: Oxford Technologies Ltd,
4: National Nuclear Laboratory, 5: NUVIA, 6: James Fisher Nuclear Ltd,
7: CREATEC, 8: Cavendish Nuclear Ltd, 9: WS Atkins,
10: Amec Foster Wheeler, 11: International Nuclear Services
GB09

Decommissioning and Waste Management
Jorge Luis Weir Vera and Tom Withey
NUVIA Limited

GB10

Decontamination and Decommissioning with Radiation Imaging Solutions
Mike Anderson, Kazushi Watanabe and Victoria Anderson
Innovative Physics Limited

GB11International Case Studies Applicable to the Decommissioning and Waste
Management Challenges at Fukushima Daiichi
Jenna McIntyre, Nick Bold
DBD International

Session R: Russia
R01

Fuel Debris Management System
Main Technologies and Approaches for Fuel Debris Treatment
Oleg Bagryanov
TENEX

R02

Options for Fukushima-Daiichi Liquid RW Management Improvement
Sergei Semin1 and Adamovich Dmitrii2
1: TENEX, 2: NIKIMT

R03

Solid RW Management System Main Technologies and Approaches for Solid RW
Treatment
Oleg Bagryanov
TENEX

Session US: United States of America
US01

Effective Design and Assurance for Permanent Geologic Storage of ContaminatedLiquids: Site Selection, Surveillance, and Safe Operations
Ahmed Abou-Sayed1, JC Rogiers2, Satoshi Mochizuki3 andOmar Abou-Sayed1
1: Advantek Waste Management Services LLC,
2: Oklahoma University, 3: Consultant

US02

Nuclear Decommissioning and Risk Management from 30 Years of ExperiencexiTim Carraway, Stan Hodges, Gavin Winship and Max Ehrhardt
AECOM
US03

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

xii

A01

Measurement of lower part of PCV of Hamaoka NPP Unit 2
by transmission type muon radiographyusing nuclear emulsion

Kunihiro Morishima1, Mitsuhiro Nakamura1, Kenji Tsuji2 and Masataka Ooyama2
1

Nagoya University, 2 CHUBU Electric Power Co., Inc.

Abstract
In order to take images of the bottom of RPV and the lower part of PCV non-destructively by muonradiography, nuclear emulsion detectors, that have many advantages, were set at the lower level than the groundat Hamaoka NPP Unit 2. It enables us to catch the bottom of RPV and the lower part of PCV in the field of view.
1. Introduction
In order to carry out the decommissioning of Fukushima Daiichi NPP (hereinafter referred to “1F site”)
smoothly, it is important to grasp where and how much fuel debris is.
Transmission type muon radiography has been carried out at 1F site, but detectors are installed on the
ground level from the viewpoint of detector size, radiation shield and power supply.
By using nuclear emulsion (Fig.1), three-dimensional trajectories of muons can be detected with theresolution of microns and mrad. Also nuclear emulsion has features such as small size, light weight, no power
supply required and waterproof property, so there is almost no restrictions on installation location. Therefore,
nuclear emulsion enables us to realize transmission type muon radiography to observe the bottom of RPV andlower part of PCV.
2. Observation and model calculation
Nuclear emulation plates were set on the 2nd floor basement of reactor building of Hamaoka NPP Unit2(Phase 1 and 3), and were set in the lower part of sub-drains for groundwater drainage around of the reactor
building (Phase 2) (Fig.2)
Phase 1: 2015 May 22 – Jun. 12, Phase 2: 2016 Jan. 28 to Mar. 10 ,Phase 3: 2016 Aug. 23 - 2017 Feb. 17
In parallel with the observation, model calculations were performed using building data of such as RPV,
in-core components, PCV, biological shielding, etc.
3. Results and Future work
We have confirmed that by
setting nuclear emulsion on the 2nd
floor basement of reactor building
or in the sub-drain around thereactor building, we can observe thebottom of RPV and the lower part of

Fig.1 principle of muon detection

PCV (Fig.3).

Fig.2 place of nuclear
emulsion plates

In the future, we will further
polish up the model calculation andcontinue the study to estimate thelimit resolution how far we cangrasp the bottom zone in detail.
Fig.3 example of results

1

A02

Technology Development to Predict Dose Rate Distribution in PCV
of Fukushima Daiichi Nuclear Power Station
Eka Sapta RIYANA and Keisuke OKUMURA
Collaborative Laboratories for Advanced Decommissioning Science (CLADS),
Japan Atomic Energy Agency (JAEA)

Abstract
A technology to predict the most probable dose rate distribution in the primary containment vessel (PCV)of the Fukushima Daiichi Nuclear Power Station (1F) have been developed by combined use of sophisticatedtheoretical calculation and local measurement results. The technology was applied to the Unit-1 of 1F.

1. Introduction
In order to ensure the safety of 1F-decommissioning workers, it is important to grasp dose rate distributionin PCV. Although local dose rate has been measured by IRID, it is difficult to estimate the overall dose rate
distribution. Therefore, we developed a method to estimate the dose rate distribution by combining theoretical
calculation and measured values.

2. Dose Rate Distributions Calculation Method and Current Result
To determine the radiation source nuclides, we carried out a series of calculations such as: 1) burnup
calculation to obtain fuel compositions at the time of accident, 2) activation calculation for the structural
materials including its impurities, 3) mixing of fuel and structural materials, 4) release of volatile fission
products, 5) decay calculation of radioactive nuclides. The spatial position of fuel debris and Cs contaminationin PCV were determined based on the results of 1F accident progression analyses by IRID.
Using the determined radiation sources, photon transport calculation with a three-dimensional plant model
was performed by PHITS [1] to produces the dose rate distribution
inside PCV and its surroundings. After that, the Cs contaminationon the surface of the wet-well was modified by comparison
between the calculated and measured dose rates to obtain modified
dose rate distribution. Figure 1 shows the finally obtained dose rate
distribution inside PCV of Unit-1 of 1F.

3. Conclusion
A method to predict the most probable dose rate distribution
for 1F has been successfully developed. The reliability of themethod expected will be improved by obtaining more local

Figure 1. Predicted Dose ratemeasurement data in the future.

distribution in PCV of Unit-1

Present study includes the results of “Technology development to evaluate
dose rate distribution in PCV and to search for fuel debris submerged in water” entrusted to Nagaoka University ofTechnology by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
References[1] T. Sato, K. Niita, N. Matsuda, et al., “Particle and Heavy Ion Transport Code System PHITS, Version 2.52,” J.
Nucl. Sci. Technol. 50, pp.913-923 (2013).

2

A03

Investigation inside the Primary Containment Vessel (PCV)
Yoichi Murai1, Satoshi Okada1, 2 and Atsuyoshi Uranishi1, 3

1International Research Institute for Nuclear Decommissioning (IRID),

2

Hitachi GE nuclear Energy, Ltd, 3Toshiba Corporation

Abstract (60word)
As R&D projects for the preparation of fuel debris retrieval of the Fukushima Daiichi NPS, IRID has
developed the technology that enables to investigate inside PCV. In FY 2016 the investigation robots
successfully entered the PCV at Unit 1 and Unit 2. The result of investigation such as images collected by the
robot camera and radiation dose detected are shown.

Crawler

Dosimeter
camera1. Investigation robots
Winch

1-1. For Unit 1 : B2 investigation
100mm

The Robot (Fig.1) named “PMORPH” can passthrough a pipe (Diameter is 100mm), and run roughSensor unitLED(a) “I” shape(b) “U” shapeFig.1 Investigation robot for Unit 1

plane after shape changing. And, the robot lets a sensor
unit descend for investigate downstairs situation.
1-2. For Unit 2 : A2 investigation
The Robot (Fig.2) has been developed to investigate
inside the pedestal of the Unit 2 PCV (Fig.2). After entering
the PCV, the shape of robot changes into a shape like ascorpion raising its tail to improve its field of view. Also thisFig.2 Investigation robot for Unit2

mechanism enables to restore if the robot overturned.
2. Investigation result

Gr ating level
Water level2-1. Unit 1
As a result of B2 investigation, underwater

Falling object

images (Fig.3) and dose rate distribution (Fig.4)
were provided, and the situation of theDistance from bottom floor(m)Fig.3 Bottom image

Fig.4 Measured dose rate

basement floor of Unit 1 gradually became
clear.
2-2. Unit 2
At the first time after the accident, image inside
the pedestal of Unit 2 (Fig.5) were captured. As theresult, the opening and deposition were found on the
structure inside the pedestal.

Fig.5 Image inside the pedestal of Unit 2

3. Conclusion
IRID has developed the investigation technology and applied for the inside PCV of the Fukushima Daiichi
NPS. As results of the investigation, valuable data for the preparation of fuel debris retrieval were obtained.
References[1] TEPCO HP : http://www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170327_01-e.pdf[2] IRID HP : http://irid.or.jp/wp-content/uploads/2017/04/20170330_2.pdf

3

A04

OECD/NEA International Joint Research Project
regarding fuel debris characterization PreADES & TCOFF
Akira Nakayoshi1, Tadahiro Washiya1, Masaki Kurata1, Kimihiko Yano1,
Yuji Nagae1, Masahiko Osaka1, and Hideki Ogino1
1

Japan atomic energy agency

Abstract
JAEA is conducting two projects on OECD/NEA. One is PreADES (Preparatory Study on Analysis of
Fuel Debris) aims at ensuring processes are in place for the debris retrieval and analysis. The other is TCOFF(Thermodynamic Characterization of Fuel Debris and Fission Products based on Scenario Analysis of Severe
Accident Progression at Fukushima-Daiichi NPS(1F)) aims at improving of thermodynamic databases for coreand FPs. Outlines are introduced at the poster.

1. PreADES
The purpose of PreADES (2017-2020) is summarizing of expertise that contribute to understanding debris
characteristics and creating of appropriate methodologies for future safety assessment of the debris retrieval andanalysis. Following actions are ensuring this project; (1)To share and update information and expertise on
debris in nuclear accidents, (2)To jointly compile and evaluate selected topical issues based on updated data andinformation, and identify research gaps and priorities and (3)To prepare future collaborative research anddevelopment plan on analysis of debris. Expected outcomes are; (1) Information

DebrisSampleand data provision（1F study, SA study, etc.）, (2) Methodologies of estimation of

Analysiscurrent properties of fuel debris and future prospect, (3) Methodologies of safety
evaluation on 1F sampling and retrieval and (4) International R&D framework(Fe,Cr,Ni)2Zr

ZrB2

ZrC

on fuel debris analysis. Regarding the cost for the project, Japanese communities
and participants will share the cost arising from activities in the project. The

B4C/SS/Zr Mixture

kick-off meeting will be held in 6-7th of July 2017.
Figure1. Image of debris

2. TCOFFThe purpose of TCOFF is improvement of thermodynamic database (TD) foranalysis and characterizationcore materials and fission products (FP) by analytical and experimental studies byreferring accident progression in 1F. The updated TD is able to contribute to evaluationof fuel degradation progression, FP-behavior, and debris characterization from materialscience viewpoints. Benchmark evaluation using various TDs, prioritization of
improvement of each TD, and evaluation using updated analytical and experimental
knowledge will be performed in collaboration of 15 research institutes. Each participants
contributes to TCOFF as in-kind, which include thermodynamic evaluation, simulation
study, material science analysis and etc. JAEA will act as the technical advisor of theseassessment. Two task forces (TF) are supposed to be established in TCOFF, such as(TF1) corium and fuel and (TF2) FPs. The kick-off meeting will be held in 3rd
Figure.2 Example of phaseand 4th of July 2017.

diagrams planned to be updated
The participation of additional experts are very welcome in both projects.

4in TCOFFA05

Development of Ultrasonic Underwater Imaging Method for
Decommissioning of Fukushima Daiichi Nuclear Power Station
Ryosuke Kobayashi1, Hayato Mori1, Takahiro Nagai1 and Naoyuki Kono1
1

Hitachi Ltd., Research & Development Group

Abstract
For the safe decommissioning of Fukushima Daiichi Nuclear Power Station (1F), survey techniques havebeen expected to provide an understanding of the current plant status. Environmental conditions of survey
targets are high radiation, having underwater areas and so on. This paper describes an ultrasonic imaging
method combined with a convex scan and a divergent beam to visualize environments in turbid water.
1. Introduction
Survey techniques have been demanded to facilitate the planning and the actual decommissioning of 1F.
Underwater imaging techniques are essential for recognizing the surrounding environment.
2. Ultrasonic Underwater Imaging Method
A method that combined a convex array sensor with a defocusing acoustic lens is proposed for wide area
imaging. Ultrasonic beam scanning in the proposed method is shown in Fig.1 (a). Ultrasonic beams are focusedin a direction parallel to the element array direction and defocused in a direction perpendicular to it. Each beam
has a thin fan shape and scanned electrically. Information on the surface shape of an underwater object isobtained as the difference of time-of-flight in each spread fan beam, as shown in Fig.1 (b).

(a) Spread fan beam scanning(b) Imaging method using time-of-flight

Figure 1. Ultrasonic imaging method using convex array sensor and acoustic lens
3. Experiments and Results
A grating panel and a cylindrical weight were measured as aconfirmation

experimentof

imaging

performance.

Locationsofmeasurement objects were about 1 m from the sensor. A photo of themeasurement targets and the measurement results of the ultrasonic imaging
are shown in Fig. 2. From the results, a wide region was visualized with thethree-dimensional effect and spatial resolution was about 20 mm because
the 25-mm spacing of the grating panel was identified.
4. Conclusion
An ultrasonic imaging method combined with a convex scan and adivergent beam was developed to visualize environments in turbid water.
As a result of imaging performance test, it was confirmed that the spatial
resolution was about 20 mm.

5

B01

3-D sensing technologies of radiation and environmental structure at
the Fukushima Daiichi NPS

Yuta Terasaka1, Yuta Tanifuji1, Masaaki Kaburagi1, Yuki Sato1, Hiroshi Usami1,
Kuniaki Kawabata1, Hiroko Nakamura Miyamura1 and Tatsuo Torii1
1
Collaborative Laboratories for Advanced Decommissioning Science, Japan Atomic Energy Agency
Abstract
We confirmed that the compact Compton camera (C3) developed in this study can visualize the radiation
distribution of the hot spot in the coastal region of Fukushima Prefecture. Additionally, we confirmed that thelaser scanner applied in this study can construct the 3-D point cloud data of the mock-up staircase that simulated
the staircase inside Fukushima Daiichi Nuclear Power Station (FDNPS).
1. Introduction
Radiation distribution imaging inside the FDNPS building is urgently required to reduce the external doseof workers and for the efficient decontamination. In this study, radiation distribution imaging using C3 and 3-Dmodelling of mock-up staircase using laser scanner were tentatively conducted for the 3-D sensing of radiation
distribution and environmental structure inside the FDNPS building.
2. Instruments
Compton camera is a radiation imaging device which can estimate the incident direction of gamma-rays
using Compton scattering principle [1]. C3 developed in this study consists of gamma-ray sensor, signal
processing unit, and optical camera. The total weight of C3 is less than 1.0 kg. The gamma-ray sensor employs a
Ce-doped GAGG (Gd3Al2Ga3O12) scintillator coupled with a multi-pixel photon counter. An operation test of C3
was conducted in the outdoor environment of coastal region of Fukushima Prefecture, where the dose rate at
1cm above the ground surface exceeds 20

Sv/h very locally.

Laser scanner is the sensor which can measure distance to the object surface using a principle of TOF(Time of Flight).

In this experiment, we used YVT-X002, Hokuyo Automatic CO. LTD, the sensor that can get

3D point cloud data, and we measured the mock-up staircase that simulated the staircase inside FDNPS and canchange step height, inclination and width of the handrail.
3. Results and future task
Figure 1 shows the result of performance test of C3 in Fukushima Prefecture. Measurement time anddistance from the hotspot were 20 min and 5 m, respectively. C3 visualized the hotspot very clearly, where the
Sv/h. Figure 2 shows the picture of mock-up staircase

dose rate at 1cm above the ground surface exceeds 20

and measured 3-D point cloud data by the laser scanner. We confirmed that we can recognize the shape of thestaircase structure roughly.
In the future, we plan to use the C3 and laser scanner for the 3-D sensing of radiation distribution andenvironmental structure inside FDNPS.

Intensity

1.0

21.7 μSv/h

0.7

Figure 1. Radiation image by C3

Figure 2. Laser scanner data and picture of mock-up staircase

References[1] J. Kataoka et al., Nucl. Instr. and Meth. A, vol. 732, 2013, pp. 403-407.

6

B02

Development of a small and modular UROV for surveying
radioactive materials on the lakebed
Mizuki Tanaki1, Kanno Ryo1, Luis Canete1 and Takayuki Takahashi1
1

Fukushima Univ.

Abstract
An underwater robot for surveying radioactive materials on the lakebed is being developed by the authors.
This underwater robot has a modular structure for easy handling and is capable of undisturbed soil coresampling from many points over a wide area.

This paper describes the soil core sampling and underwatercommunication method. Results of experiments are also presented.

1. Introduction
Surveying of radioactive materials on lakebeds requires soil core sampling without disturbing the soil
layers to investigate sediment chronology. Conventional soil core sampling by divers is manually intensive,
physically challenging, dangerous, and difficult to perform in deep lakes withdepths beyond 40 meters.

Other sampling devices that can be dropped fromboats exist but do not give accurate position data for mapping. These problems
can be solved by the underwater robot being developed [1].

2. Soil core sampling module and inter-module communication
Based on the estimated radioactivity concentration and the sedimentaccumulation rate of the lakebed, the specification of the core sample is Figure 1. Underwater robot
determined to be at least 100g in weight and 100mm in depth. We conducted various experiments and finallyconcluded that a method of intruding a stainless pipe with 40 mm diameter and 0.5 mm thickness yields the best
result [2]. A soil core sampling module that can collect 8 samples per dive is installed to the robot.
Power and communication ports may cause water breaches and so each module has an individual battery
and communicates wirelessly.

However, the reliability of radio-communications in underwater environments

was not enough due to the significant attenuation of radio signals.

To overcome the issue, the authors

developed a patch antenna which is simply attached to the module surface. The radio signal strength was
drastically improved which increased the reliability of inter-module communication.

3. Experiment for soil core sampling
The prototype of the robot that features the developed patch antenna and soil-core sampling module wastested in Lake Inawashiro.

It was shown that inter-module communication was improved and soil-coresampling can be performed successfully by the robot.

References[1] ANZAI K., TAKAHASHI T.

(2015) Development of a small and modular UROV for environmental

surveying-Improvement of an undisturbed soil core sampling module-, ROBOMECH2015, 2A1-E07.
[2]TANAKI M., MANH D., CANETE L., TAKAHASHI T. (2015) Development of a small and modular UROV forenvironmental surveying -Improvement of thruster module and underwater communication method-, The 299th SICE, 299-5.

7

B03 Decommissioning Operations Planning Using Proven Technologies
Yasushi Kikkawa1, Daphne Ogawa1, Bastien de Lazzari1,
1
AREVA ATOX D&D SOLUTIONS Co. Ltd.

Abstract
ANADEC proposes proven technologies and expertise for the decommissioning of the Fukushima-Daiichi(1F) plant based on AREVA’s unique experience as both operator and dismantling entity of its facilities, and
ATOX knowledge of site operation conditions in Japan. Fukushima-Daiichi situation is unique, and requires
both innovative solutions, and robust proven solutions. ANADEC aims to provide dedicated filed engineering
services.
1. Introduction
Dismantling and decommissioning works at the 1F plant has entered into new phases which will require
technologies to:► achieve dose reduction and prevention of contamination dispersion,► perform remote decommissioning to clear access to strategical work area such as PCV, operating floor.
► manage and stabilize radioactive waste.
2. Proven technologies applied by ANADEC for Fukushima-Daiichi D&D
2-1. Dose reduction plan, chemical decontamination and prevention of contamination dispersion
The first step to elaborate a dose reduction plan is to perform a contamination cartography. ANADEC hasdeveloped with a French company IZYmap, remote dosimeter ball sensors that record relative position andradiological data in order to build a 3D mapping of contamination.
Once the hot spot are identified, there are three ways to lower the dose rate: by shielding, bydecontamination of the source, and by removal of the source term. ANADEC is specializing in the second
solution, by providing decontamination operations using gel or foam-based chemical reagents.
Moreover, to prevent contamination dispersion when performing dismantling works, ANADEC with thesupport of French CEA and AREVA develops fixative foams that can be sprayed to cover rubbles to avoid
dissemination of contamination.
2-2. Remote-controlled dismantling operations
Given the high dose rate and the difficult accessibility of 1F operationsites, ANADEC offers solution engineering for intervention using remotely
controlled robot such as the TEΩ600 force-feedback arm developed by
AREVA TEMIS. This robot can be adapted with various tools to carry out
complex operations in hostile environment. By using the force-feedback
feature, remote operations such as cutting, drilling, welding, opening ofvalves, etc. can be performed with more accuracy, more safety.
Figure 1. MC-TEΩ600
3. Field engineering planning by ANADECInterventions onsite 1F requires the elaboration of a thorough planning and operation scenario. Byleveraging on AREVA’s extensive experience in complex dismantling operations, ANADEC performs fieldengineering that makes the best use of pragmatic solutions for both indoor and outdoor works.
► Definition of end state objectives:► Investigation to obtain initial state conditions:► Definition of scenario and corresponding operational sequence:► Selection of technologies to perform the tasks:► Mock-up tests► Equipment procurement, on-site installation, commissioning, and actual operation
4. Conclusion
ANADEC considers D&D is a just-mix of high-tech solutions (e.g. TEΩ600 force-feedback arm) androbust proven solutions. By combining both, the best cost/time/dose compromise can be provided to solve thecustomer issue.
ANADEC benefits of D&D technologies which are already proven by AREVA in France, Germany, UK
and the US. The operational return-on-experience from AREVA and ATOX is an asset of which ANADEC takes
advantage to propose solutions to the particular situation of Fukushima-Daiichi.

8

B04

Radiation Tolerant Camera by Mirion Technologies
The world’s highest radiation tolerant performance camera
Ken Takahashi, Tadashi Sato
Cornes Technologies Limited

Introduction
The Mirion (IST-Rees) brand of radiation tolerant cameras is recognized as the market leader and specialist of
CCTV & Imaging Systems in the nuclear industry, continuing supplying the systems over 30 years world-wide.

1.
1-1.

Radiation tolerant performance camera
R93 High Radiation Camera, R942 High Radiation Zoom Camera
The R93 camera, equipped with a non-browning prime lens and the R942 compact
zoom camera with a non-browning 12 to 72mm zoom lens are designed to provide
high resolution monochrome images in the most extreme radiation environments,
such as class leading radiation tolerance of > 2 MGy for the use of periodical
inspection purposes, as well as for decommissioning reactors in Nuclear Power

Plant. Both cameras are composed of camera, camera controls and cables.
1-2.

R981 Compact System Camera

The R981 compact system camera is a pan-tilt-zoom camera with higher radiation
tolerant to 1MGy. The camera and pan and tilt are combined into an integral unit
with three positions for additional lights and microphone.
1-3.

DOTCAM HR Radiation Resistant Camera, Small Footprint
The Dotcam HR is a miniature color camera designed for use in higher radiation
environments, up to a total absorbed dose of 100krad (1kGy), and being equippedwith an integral white LED lighting ring, 4.0mm f/5.0 lens and is fully immersible
to a depth of 60m.

2.

Real-Time Video Deblurring and Processing System for Live Nuclear Core Inspection

The NOTURB® system was developed following many years close collaboration between Techway and EDF.
The Techway NOTURB® video processing system assists operators during outage inspection by removing theeffects of underwater thermals when viewing hot fuel assemblies. This can reduce the inspection time requiredfor outages and increase operator confidence in the accuracy of the inspection process. Additional standard
features including fuel gap measurement to assist operators in fuel handling
operations and a filter to remove radiation induced video noise from CCD or
CMOS based camera systems.

A) raw video B) processed video

9

B05

3D information technology based support for nuclear
decommissioning
István Szőke

Institute for Energy Technology / OECD Halden Reactor Project, Norway
Abstract
New concepts enabled by advanced information technologies are being employed for improving safety andefficiency of the upcoming extensive nuclear decommissioning work worldwide. The need for such techniques
is especially clear for degrading legacy and accident sites where safety planning is demanding due to the highradiological and other risks. In this work we present how such modern concepts are applied for improving
nuclear decommissioning by enabling effective training and informed safety planning.
1. Introduction
Advanced information technologies like 3D simulation, virtual and augmented reality, advanced userinterfaces, and mobile and wearable computing devices support innovative concepts for improving safety andefficiency of nuclear decommissioning projects. Application of such technologies in combination with 3D
radiological simulation methods enable planning and training for decommissioning work with an effective
perception of risk and readiness for risk mitigation. Hence, such techniques are especially indispensable atdegrading legacy and accident sites where a common awareness of the radiological situation, effective training
for safe execution of complex jobs and preparedness for reposing well to unexpected situations is crucial.
2. Methods
This work describes how 3D radiological simulation methods in combination with advanced 3D
visualization techniques and hardware can be applied for supporting decommissioning of nuclear sites.
At accident sites, for instance, such techniques will enable cost effective training for remote and manual work inenvironments with high radiation levels (Fig 1).
3. Conclusion
Advanced support systems based on 3D simulation are
successfully being applied in the decommissioning of a number ofnuclear installations (e.g. Chernobyl NPP, Andreeva Bay branch ofNorthwest Center for Radioactive Waste Management in NW
Russia, Leningrad NPP, Fugen NPP) [1-4].
Figure 1. Training in Virtual Reality

References[1]

I. Szőke, T. Johnsen, Nuclear Safety and Simulation, 4(3), 219-25, 2013.

[2]

I. Szőke et al. Journal of Radiological Protection; 34(2), 389–416, 2014.

[3]

I. Szőke et al. Journal of Nuclear Science and Technology, 52(3), 371-82, 2014.

[4]

K. Chizhov, M. Sneve, I. Szőke et al. Journal of Radiological Protection, 34(4), 755-73, 2014.

10

B06

Activities of the Fukushima Consortium of
Robotics Research for Decommissioning andDisaster 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 tothedecommissioning, decontamination and disaster response robotics industry, Fukushima Consortium of Robotics
Research for Decommissioning and Disaster Response (hereinafter called “the Consortium”) is working toestablish a network among manufacturing industries in Fukushima Prefecture and related organizations.
1. Overview of the Consortium
The Consortium was established to support the manufacturing companies in Fukushima Prefecture which
intend to participate in the decommissioning and decontamination projects at TEPCO's Fukushima Daiichi NPS,
and to expand their businessto disaster response robotics by utilizing their technologies developed andaccumulated in the decommissioning and decontamination projects. This Consortium organizes technicalseminars and business matching events as well as

a variety of exhibitions and demonstrations of robots andother products for networking and information exchange with related organizations.
2. Main activities of the Consortium
2-1. Technical seminars and technology matching events
The Consortium provides for its members opportunities to exchange
technical information useful in the field. It also offers meetings fordiscussing various use of the robots under development.
2-2. Robots demonstration events and exhibitions
These exhibitions could be a good opportunity for theConsortium members to expose their products and technicalFigure 1. Technology matching events

capabilities, because they can appeal their robots by demonstratingin front of potential customers.
2-3. Other activities
The Consortium members can be supported by expert coordinators.
Field survey and site visit to related companies, and information
gathering from governmental organizations are possible. Besides,
latest project information of new tenders will be provided timely to
Figure 2. Robots demonstration events

our members.
3. Unique Features of the Consortium

Because the Consortium is managed by Fukushima Technology Centre, a public testing research organization
established by Fukushima local government, the members can exchange information with companies involvedin the projects at TEPCO's Fukushima Daiichi NPS, and also can get the latest information about thedecommissioning and decontamination projects headed by the national government. Normally, it is difficult for
local companies alone in Fukushima to get such opportunities, but the Consortium members have various
privileges.

11

B07

Outline of Robots Developed by the Members of
Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response
Yoshiro Owadano1,
1

Fukushima Consortium of Robotics Research for
Decommissioning and Disaster Response

Abstract
Aiming for the entry into the market of the decommissioning, decontamination and disaster response robotics,
member companies of the Fukushima Consortium of Robotics Research for Decommissioning and Disaster
Response have developed robots with unique features. Below are some examples.
1. Overview of robots developed by the Consortium members
1-1. Fire Fighting Robot（Kouyou Co., Ltd.）This fire-fighting robot is configured with no electrical components inthe main body, and all parts are powered by water pressure.
Because of such feature, secondary damage doesn’t occur even if thisrobot would accidentally be broken. It works even in high temperature firesite.
1-2. Small electric-powered disaster response robot「SPIDER」

Figure 1. Fire Fighting Robot（Aizuk, Inc./ The University of Aizu）“SPIDER” is capable of climbing up and down the stairs and climb overdebris by using four sub-crawler that is operational independently.
“SPIDER” applies wireless manipulation using a commonly used gamecontroller and tablet terminal. For this reason, “SPIDER” can search fordisaster victims in the site where the person is hard to enter directly because
it is too narrow to enter, and may occur the secondary disaster such as

Figure 2. SPIDER

landslide and collapse of tunnel.
1-3. Non-contact laser cleaning machine「ELASER」（Tosei EB Tohoku Co., Ltd.）The cleaning laser device "ELASER" is capable of noncontact dry cleaning
without using chemical liquid or abrasive, and no post-processing is necessary.
It can clean impurities on metal surface such as rust, paint, oil content and resin
residue by laser irradiation without damaging base materials.
This is an innovative and environmentally-friendly cleaning machine with highversatility, less work and lower cost.
2. Others

Figure 3. ELASERIn addition to the robots introduced above, many other products developed by
the Consortium members are available in the market.
We are pleased to introduce our member companies capable of helping your new product development. They
can provide proper technical solutions in your business related to the decommissioning and disaster response. So,
if you are interested in such services, please contact the secretariat of the Consortium.

12

C01

Mirion Technologies (Canberra) Experience and Success for
Efficiency and Safety Decontamination and Decommissioning
Atsuo Suzuki, Lou Sai Leong and Helene Lefebvre
Mirion Technologies (Canberra) KK

Abstract
Mirion Technologies (Canberra) Inc. (MTC) has more than 30 year experiences of various radiation
measurement and safety Decontamination and Decommissioning (D&D). Every type of D&D projects has
different requirements and characteristics. MTC proposes a large range of effective products and technologies
for D&D. For example, In-situ Object Counting System (ISOCS), Ultra Portable Gamma-Ray Imaging System(iPIX), Alpha/Beta Portable Air Monitor and Depth Profiling of Contamination Distribution technology are
already widely applied in Nuclear Facilities worldwide. MTC solutions are the key for efficient and costeffective of D&D.
1. Introduction
Every type of D&D projects has different requirements and characteristics. For example, if you aredecommissioning a Nuclear Power Plant, you will predominantly perform gross gamma counting and you willneed well-defined waste fingerprints and streams and re-categorization of waste for storage. D&D activities forfuel cycle facilities and research reactors run the risk of criticality accidents and usually require a full
re-characterization of the facility prior to decommissioning and a high level of expertise using our products.
2. MTC Products and Technologies
2-1. ISOCS The ISOCS (In Situ Object Counting System) Calibration Software brings a new level of capabilities
to gamma sample assay by eliminating the need for traditional calibration sources during the efficiency
calibration process. ISOCS efficiency calibration is usually more accurate than that of standard source calibration
method. With 21 geometry templates available, most objects can easily be modelized. In addition to this, ISOCS
Uncertainty Estimator is very strong tool for evaluation of uncertainty.
2-2. Depth Profiling of Contamination Distribution
In-situ Measurement of Depth Tagged Dust Samples (TruPro Technology)
Drill small holes (few cm) but deep (up to 6 m) with a hollow drill, and collect dust samples at various depths(5cm incremental steps). Place the dust samples in a lead shielding, to directly measure on site the activity byhigh resolution gamma spectroscopy using ISOCS system.
Tools are dedicated drilling system (TruPro) with ISOCS system and iSOLO αβ counting system
Detectors Inserted Inside Drilled Holes
Insert a gamma detector into a drilled hole and shift the detector in several equal or non-equal distance positions.
The results give the distribution of activity along the hole axis. In addition, using a gamma simulation code witha Monte-Carlo code based on this result give a better evaluation of radionuclide distribution and dose rate.
Tools are LaBr, HPGe Sealed probe or Geiger Muller. For Software are MERCURAD, MCNP5 and so on.
3. Successful achievements (Activity depth profiling of building in France)
Objects were decommissioning the building, mapping of the walls, the grounds and the floors without releasing
any dust. By TruPro Technology and ISOCS calibration and analysis, 280 samples were measured and the
project was completed in 21 working days. Our technology was evaluated as follows by EDF. Excellent
detection and reduce project timescales and costs, immediate availability of the data and the accuracy of its
results allowed the team to identify hot spots.

13

C02

Active/passive neutron/gamma assay systems forfuel debris and nuclear waste activity and isotopic
mass determination
Lou Sai Leong, Atsuo Suzuki and Helene Lefebvre
Mirion Technologies (Canberra) KK

Abstract
CANBERRA has more than 30 year experiences of various automated neutron, gamma and combined Non
Destructive Assay (NDA) systems. These systems allow determining the fissile isotopes (Pu/U) quantification,high dose rate activity measurement, in the field of safeguard and safety. This method is widely applied in theworld.

1. Introduction
The neutron system is normally consisted of two types of detectors: active and passive detector. We usedpassive neutron coincidence counter to derive the spontaneous fission. Active counting system is based on

the Differential Die-Away (DDA) method which allows separating the prompt neutrons and delayed
neutrons.
2. Neutron counting system
2-1. Passive coincidence system
The detection system is based on 3He gas for neutron absorption. The passive assay is able to measurespontaneous fission neutrons emitted by the waste. Recorded signals are analyzed using neutron coincidence or
multiplicity electronics, such as shift registers, to derive the spontaneous fission isotopes mass (238Pu, 240Pu and242Pu) of the waste sample.

2-2. Active system
The principle is to use a pulsed neutron generator to produce at a set
frequency of fast neutrons. These neutrons will be slowed down in the
counter moderator and be absorbed in3

He gas counting system.

Quantitative analysis gives the mass of fissile material (235U and 239Pu)
which is calculated by the gate of counts between prompt and delayed
neutrons signals.
2-3. Gamma spectrometry system
Gamma spectroscopy is able to obtain the isotopes ratio using Multi-group
analyzer, especially in Pu/U isotopes mass determination using correlations
between fission product and U/Pu content.
3. Experiences
Several projects of customized NDA systems designs were performed for the challenge D&D application in theworld, such as high dose rate debris removal in UK or feasibility study of NDA system performance inFukushima site. These customized NDA systems depends on the characterization on different project.

http://www.canberra.com/literature/brochures/pdf/Waste_Management_Solutions_C49012.pdf

14

C03

Development of simple and rapid analytical methods for Sr-90 in
the contaminated and treated water in the Fukushima Daiichi
Nuclear Power Station.
Tomomune MATSUNAGA1, Takuya NAKAGAWA1, Takuyo YASUMATSU1
Hirofumi TAZOE2 and Masatoshi YAMADA2
1

Tokyo Power Technology Ltd.,

2

Hirosaki Univ.

Abstract
It is necessary to improve the analytical methods for Sr-90 for assessment of the performance of
decontamination facilities. A simple and rapid method for Sr-90 in the contaminated and treated water in the
Fukushima Daiichi Nuclear Power Station is reported. We developed simple and rapid chemical separation of Sr
method by automatically controlled solid phase extraction system using Sr-Resin column.
1. Introduction
Multiple
Multi-nuclide

decontamination

facilities

Removal(Advanced

Facility

includingLiquidProcessing System = ALPS) treat the contaminated waterin the Fukushima Daiichi Nuclear Power Station. It isnecessary to simplify the analysis of radionuclide, such as
Sr-90, to evaluate the performance of the facilities.
Furthermore, the demand of Sr-90 analysis has been
increased (Fig.1) For the purpose of improvement of thesample throughput and the quality of analysis, wedeveloped simple and rapid chemical separation of Sr byautomatically controlled solid phase extraction system

Fig. 1 The temporal variation of the numberof Sr-90 data. (estimation value)

using Sr-Resin column.
2. Experimental
2-1. The automated Sr separation system
The three-way valve and the peristaltic pump wascontrolled by a sequencer. Operation sequence wasprogrammedforcolumnpreparation(cleaningandconditioning) and sample separation (loading, washing andelution) for Sr.
2-2. Spike and Recovery Test
Automated separation was performed for simulated
sample spiked stable Sr. The elution profile and thechemical yield of Sr in eluent were obtained by ICP-MS

Fig. 2 Schematic layout for the automatic

analysis. Sr was quantitively recovered (98.4%).

Sr separation system.

3. Conclusion
The automated Sr separation system make it possible to simplify and hasten chemical separation of Sr. In
the future, we will carry out the practical application of the developed system using contaminated or treatedwater .

15

C04

ICP Mass-Spectrometric Analysis of Radioactive Strontium
Without Calibration Curve
Mizuki Odashima1, Makoto Furukawa1,2,3, Katz Suzuki4, Yoshitaka Takagai1,5

Fukushima Univ., 2Tokyo Univ., 3Perkin Elmer, 4 JAMSTEC,5IER Fukushima Univ.

1

Abstract

After the accident at Fukushima Daiich Nuclear Power Station, a rapid analysis system for 90Sr,"cascade-type ICP-MS method" was developed.1 In this study, a mass spectrometric isotope dilution
method for the cascade-type ICP-MS is presented. The total run time and high precision can be improved
using this method.
1. Introduction

In 2011, East Japan great earthquake occurred and the accident happened at Fukushima Daiichi Nuclear
Power Station. 90Sr is difficult to analyze due to fundamental interference in instrumental analysis.
Recently, a cascade-type ICP-MS method for 90Sr was developed, and very short time analysis can be
conducted as compared with traditional analytical methods. Not only this method but also otherquantitative analysis of 90Sr has required calibration curve in quantification analysis. We have invented
applying a mass-spectrometric isotope dilution method to the cascade-type ICP-MS. Using the isotope
dilution method, the calibration curve is unnecessary; thus the total analytical run time can be improved.
In this study, a cascade-type ICP mass-spectrometric isotope dilution method for 90Sr was demonstrated.
2. Experiments

Optimized ICP-MS measurement parameters were used and the mass bias in MS was adjusted to natural
isotope abundance ratio (88Sr/86Sr). The sample including 90Sr was added 86Sr-enriched spike solution. The
mix solution was measured using cascade ICP-MS. After measurement, amount of stable Sr was quantified
by isotope dilution method, and then an amount of 90Sr in the sample was calculated by the intensity ratio(88Sr/90Sr).
2-1. Results and Discussion

The proposed method was optimized for the quantification of stable isotopes (86Sr and 88Sr) andradioactive 90Sr. To demonstrate the efficacy of this method, the spike and recovery tests were conductedunder the optimized condition. As these results, stable isotopes were quantified in the concentration
range of environmental level.

In addition, the results between amount of radioactive spiked andrecovered were satisfactorily corresponded. In the application to diluted radioactive contaminated water,
the quantification value was corresponded with value which was obtained by typical radiometric analysis.
3. Conclusion

New quantitative method for 90Sr without calibration curve based on isotope dilution method wasdeveloped and demonstrated. In this method, the addition amount and the measured values weresignificantly corresponded. The total analytical run time of presented method became shorted as comparedwith the previous cascade ICP-MS (original version).
References[1] Y. Takagai, M. Furukawa, Y. Kameo, and K. Suzuki, Anal. Methods, 2014, 6, 355.

The project is supported by the MEXT, Human Resource Development and Research Program for
Decommissioning of Fukushima Daiichi Nuclear Power Station

16

C05

Development of Quantitative Method for Strontium-90
Using Thermal Ionization Mass-Spectrometry

Chihiro Ito1, Takashi Miyazaki2, Shigeyuki Wakaki2, Katz Suzuki2, Yoshitaka Takagai1, 3
1

Fukushima Univ., 2JAMSTEC, 3IER Fukushima Univ.

Abstract
In the decommissioning of Fukushima Dai-ichi Nuclear Power Station (1F), there are problems inradionuclide analysis, in particular, quantification of 90Sr. In the time, the thermal ionization mass-spectrometricquantification method of 90Sr based on the combination system of an isotope dilution method (ID) and a total
evaporation (TE) method is presented.
1. Introduction90

Sr is one of important fission product that is pure β-ray emitter. Typical public analytical method basedon radiometric analysis such as a milking process-low background gas-flow counter consumes time fromtwo-weeks to a month.
90

Recently, an inductively coupled plasma mass-spectrometric quantification method for

Sr analysis was developed; thereby, the rapid and sequence analysis can be achieved. However, in the previous

method, there was problem in analytical sensitivity of 90Sr, when environmental samples contained highly
concentration of natural stable Sr. To overcome this problem, the highly-precision analysis, isotope dilution
thermal ionization mass spectrometry (TIMS), was applied to 90Sr analysis with TE method.
2. Experimental
A trace amount of target Sr nuclide was quantified by TIMS. Quantificational certification was conductedusing stable Sr standard solution. Mass observation was m/z = 90 in secondary electron multiplier (SEM) as
detector, and other stable Sr nuclides were observed using multi-cup faraday counters. To optimize the TIMS
conditions, 1 µL of mixture solution (sample solution containing Sr and 86Sr enriched spike solution (1:1)) was
loaded on a filament, and then measured using ID-TE-TIMS.
3. Result and discussion
In the study, 1 µL of sample solution containing Sr and Ta activator was loaded on Re filament. When thesample was measured in the condition of 100 mA min-1(heat-slope) and 4 V (max-intensity for 88Sr), the typical profile
of TE-TIMS was obtained (Fig. 1). The run time was 30-40 min
which contains ion beam tuning (10 min). As comparison withtypical ID-TIMS, the repeatability (relative standard deviation(RSD)) was 0.147 and 0.109% for ID-TIMS and ID-TE-TIMS,
respectively. As the results, the quantification values were highly
Fig. 1 Signal profile of ID-TE-TIMS

accurately corresponded with additional amount.

for stable 88Sr

4. Conclusion

Precise determination method for Sr using ID-TE-TIMS was proposed. It would be expected to application
to 90Sr analysis in highly concentration of stable Sr samples such as concentrated sea water.

The project is supported by the MEXT, Human Resource Development and Research Program for
Decommissioning of Fukushima Daiichi Nuclear Power Station

17

C06

Simultaneous Quantification Method forRadioactive Strontium and Radioactive Iodine in ICP-MS
Hiroaki Ogata1, Makoto Furukawa1,2,3 and Yoshitaka Takagai1,4
1

Fukushima Univ., 2Tokyo Univ., 3Perkin Elmer, 4 IER Fukushima Univ.

Abstract
In order to sensitively measure β-rays emitter, chemical isolation processes have been required prior to themeasurement using a low background β-ray spectrometer or a mass-spectrometry. However, the identification ofbeta-ray emission nuclides is complicated and consumes long time.
90

analytical method for Sr andIn this study, a simultaneous quantitative

129

I was developed by a split flow path system with cascade ICP-MS1).

1. Introduction
Large quantities of radioactive contaminated water have been generating from the melted-fuels cooling system
in Fukushima Daiichi Nuclear Power Plant. To administrate secure decommission, the development of a rapid
analysis for radionuclide has required. Particularly traditional radiometric analysis of β-ray emitting nuclides,
90

Sr is complicated and consumes two weeks and more. Similarly 129I is also a beta-ray emitting nuclide and is

employed as tracer of environmental dynamics. Recently, a cascade type ICP-MS1) was developed for the
anlysis of 90Sr within 20 minutes; however, multi-nuclides analysis were difficult in that method. In this study, asplit flow path following cascade ICP-MS method presents simultaneous measurement of 90Sr and 129I.
2. Analysis system design
The presented system was constructed basing on a cascade type ICP-MS method1) and a split method2). The
sample solution splits two directions at a volume ratio of 1:1 in the flow path. One was introduced into a solid
phase extraction column for Sr analysis (SPE line). The other flow was directly introduced to ICP-MS for
analysis (split line). After switch of the flow path using automatic valve, each flows were sequentiallyintroduced to ICP-MS and measured.
3. Results and discussion
Preliminarily, simulated sample water containing
stable isotopes (Sr and I) was measured by thismethod. As a result, the system was satisfactorily
operatedas

expected.

Inthesimultaneousquantification, the chromatographic peak area of Sr
and the average signal intensity of I was employed.
Regarding these values of the limit of detection(LOD) can be lower using ultrasonic nebulizer

Fig. 1 Signal profile of 90Sr and 129I

(USN) and coaxial nebulizer for Sr and I, respectively.
From this reason, we have developed new chamber system and it was applied to the addition and recovery testfor radionuclides was conducted. As a result, the LODs were 30.4 and 0.01 Bq/L for 90Sr and 129I, respectively(sample volume: 10 mL) with 3.7 and 6.7% of repeatability (RSD (n=3)) for 90Sr and 129I, respectively.
References
1) Y. Takagai, M. Furukawa, Y. Kameo, and K. Suzuki, Anal. Methods, 2014, 6, 355.
2) M. Furukawa and Y. Takagai, Anal. Chem., 2016, 88, 9397.

The project is supported by the MEXT, Human Resource Development and Research Program for
Decommissioning of Fukushima Daiichi Nuclear Power Station

18

C07

Rapid detection of radioactive Sr using inexpensive reagent
and a simple-mass-spectroscopy
Kaori Chiba1and Masaki Koike2

1

National Institute of Technology, Ibaraki College, 2Hiroshima Institute of Technology

Abstract

We developed a rapid and easy method to detect radioactive 90Sr by a commercial
Matrix Assisted Laser Desorption Ionization (MALDI)-tof-mass spectrometer. As a new
matrix, dipicolinic acid (2,6-pyridine dicarboxylate) was introduced to select Sr ions forionization. Analyte ions are separated two dimensionally both by mass numbers, and byaffinity to the matrix molecule.
1. Introduction

Since the accident in Fukushima No.1 nuclear power plant, analysis of radioactive
substances have been quotidian concern not only for the people inside the disaster area but
also for the ones worldwide. For the gamma-emitting nuclides with shorter half-life such as
131I(8 days) and 134Cs(2 years), the decay of the radiation level is quick and can bemonitored easily. On the other hand for the nuclides with longer half-life, people feel moreanxious about the long-term effect of the radiation. 90Sr is a beta-emitting nuclide, with ahalf-life of 28.8 years. Detection of 90Sr by the conventional quantitative analysis takes longtime (~3 weeks). The necessity of the depthful preparation process to prepare pure Sr fromeach analyte has prevented us from the rapid and on-demand screening.
As we learned from the last disaster, the peak of the demand for the analysis is
supposed to be sudden and tentative. In order to decrease the fear in the mind of the people,
we need an easy and rapid method for the rough screening of 90Sr.
2. Methods and results

For MALDI-tof-mass experiments, Shimadzu AXIMA CFR+ was used. Before theexperiment, a new stainless sample plate was washed with 1% formic acid, methanol, acetone,
and distilled water, following the recommendation in the instrument manual. The way forthe sample preparation was simple. A sample solution containing 88Sr and additional
impurity ions, and a matrix solution including dipicolinic acid was mixed and spotted on the
MALDI-tof-ms sample plate and dried. Measurement was performed in ref ectron-mode.
The results were as follows.
1. Using the dipicolinic acid as a matrix, stable 86Sr,
87Sr and 88Sr peaks were observed depending on its

abundance ratio, even with the other impurity ionsin the sample solution.
2. All the alkali soil metal ions (Sr2+, Ca2+, Mg2+ andBa2+) are bound to the dipicolinic acid and observedin the mass spectra as singly-charged ions.

19

Figure 1. MALDI selection

C08

Laser remote analysis for in-core and in-situ monitoring of debris
in decommissioning of Fukushima Daiichi nuclear power station

Ikuo Wakaida, Hironori Ohba, Katsuaki Akaoka, Masaki Ohoba,
Ayumu Matsumoto, Masabumi Miyabe
Collaborative Laboratories for Advanced Decommissioning Science, Japan Atomic Energy Agency
Abstract
Radiation resistant optical fiber based laser induced breakdown spectroscopy (Fiber LIBS) is developed forin-core and in-situ elemental analysis of debris and its activity is performed under severe environmental
conditions such as high radiation field of about 10kGy/h and under water. Long pulse laser, microwave assisted
LIBS for more high sensitivity and liquid sample application are also introduced.
1. Introduction
For the decommissioning of “Fukushima Daiich Nuclear Power Station” which contained damaged or melt
downed core, development of rapid, easy, onsite and in-situ remote diagnostic/analysis techniques under thesevere environments such as extremely high radioactive condition will be strongly required. In order toaccomplish these requirements, the concept of probing by light and diagnostic by light with radiation resistant
optical fiber will be one of the simple, powerful and applicable choices as the innovative development based on
LIBS technology
2. Optical fiber based laser induced breakdown spectroscopy
Optical Fiber based LIBS
probe (Fig.1) is constructed, andunder water condition, we havesuccessfully observed some specificspectra from the simulated sample
of molten debris made by sinteredoxide of Zr and U, and also
observed the spectrum fromsimulated metal sample underFig.1 Concept of optical fiber based remote analysis
radioactive condition of 10kGy/hand after total dose of 2MGy. For the use of more longer optical fiber, which will be required for the necessary
from on-site or for the human safety, long-pulse laser with the pulse duration about 100 ns will be introduced tocompensate the delivery power loss and/or to enhance the emission signal. In the experiment by a metal sample,
although the emission intensity was not higher than that by conventional laser, the decay time of the emission
came to be longer. And the result, signal enhancement of several times was obtained by the time integration ofemission signal. Detailed characteristics of long-pulse laser LIBS are now under studying.
3. Signal enhancement technology and application for liquid sample
As for the signal enhancement technique, microwave assisted LIBS technique coupled with simple antenna
for the combination use with fiber LIBS will be under developing, and about ten times enhancement wasobserved under the atmosphere condition.
For the application to the liquid phase sample, especially for the analysis of the polluted cooling water, thethin-sheet liquid jets will be also introduced as the convenience and high sensitive monitoring for dissolved
elements of nuclear fuel debris.
Present study includes the results of “Advanced study on remote and in-situ elemental analysis of molten
fuel debris in damaged core by innovative optical spectroscopy” entrusted to Japan Atomic Energy Agency by
the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT)

20

D01

Development of the insoluble neutron absorbers using
self-assembled surfactant aggregates
Noriaki Ushio1, Kouji Koyanagi1 and Hiroyuki Terazaki1
1

Kao Corporation

Abstract
For the purpose of removing fuel debris from the Fukushima-Daiichi Nuclear Power Plants safely andsmoothly, we are developing a re-criticality prevention technology using insoluble neutron absorbers. We havefound out that high viscosity fluid insoluble in the water which can hold B4C particles stably by using the
VISCOTOP (complex of alkyl aryl sulfonate and alkyl ammonium salt). VISCOTOP forms self-assembly
surfactant aggregates known as a worm-like micelle. We also studied improvement of radiation resistance andelution resistance in hot water using VISCOTOP as a key compound, then found a direction of improvement.

1. Introduction
In an aqueous solution, the surfactants can form various self-assembled aggregates. In many self-assembled
surfactant aggregates form, the worm-like micelle can make a fluid with very high viscoelasticity. The
worm-like micelle can hold many B4C particles and can spread on new fracture surface of fuel debris in water.
2. Holding B4C to the worm-like micelleAs the result of having made B4C holding to 55% to VISCOTOP, it turned

out that the insoluble neutron absorbers which has moderate fluidity andthe good spreadability on the fuel debris surface can be created.
3. Radiation resistance of VISCOTOP
If alkyl aryl sulfonate and alkyl ammonium salt were irradiated bystrong γ-ray, decomposition will start and viscoelasticity will disappear

Figure1. Insoluble neutron absorbers
using VISCOTOP

by decay of a worm-like micelle. We looked for the stabilizing agent, although the decomposition rate improved,
it was not enough to maintain of elasticity. When alkyl aryl sulfonate was changed into salicylic acid salt, thedecomposition rate improved and elasticity maintained.
4. Elution resistance against hot water
VISCOTOP was solved in hot water. In reactor containment vessel, since water may become hightemperature, high temperature resistance is required for it. When the alkyl chain length of the alkyl ammonium
salt was changed to long, it became clear that high temperature resistance improves.
5. Conclusion
We developed insoluble neutron absorbers holding many B4C particles to re-criticality prevention of fueldebris by applying the worm-like micelles. We will continue to improvement examination of radiation resistanceand elution resistance against hot water.
References[1] R.Ishibashi, T.Fujita, K.Ishii, Y.Harada,”Development of criticality prevention technology by using insoluble neutron
absorbers for fuel debris removal in the FUKUSHIMA-DAIICHI nuclear power plants”, Proceedings of ICAPP 2017, Fukui
and Kyoto, Apr.24-28
Acknowledgement
This work was performed under the Shared Use Program of QST Facilities.

21

D02

Corrosion inhibition of carbon steel piping by addition of sodium
nitrite and sodium pentaborate
Kazutoshi Fujiwara1, Junichi Tani1, Joji Ohta1, Yuichi Fukaya2 and Toshifumi
Hirasaki2

1

Central Research Institute of Electric Power Industry, 2Tokyo Electric Power Company Holdings

Abstract
The corrosion inhibition effects of sodium nitrite and sodium pentaborate on the carbon steel piping inflowing diluted synthetic seawater at 40 C̊ were evaluated. A clear inhibition effect of sodium nitrate on
corrosion was confirmed at low linear velocity. The corrosion inhibition effect of sodium pentaborate wasconfirmed in the range of flow rate used in the experiment.
1. Introduction
In alternative cooling systems for the spent-fuel pool of the Fukushima Daiichi nuclear power plant, thecorrosion of carbon steel piping induced by flowing diluted seawater should be carefully considered. Although
the water in the present pool has been purified to the level of fresh water[1], further countermeasures to prevent
corrosion should be prepared in advance. In the present study, the corrosion inhibition effects of sodium nitrite
and sodium pentaborate in flowing diluted synthetic seawater at 40 ̊C were evaluated.
2. Experimental Procedure
The diluted synthetic seawater at 40 ºC was fed to the tube
specimens made from carbon steel. The dilution factor was about
200 and the chloride ion concentration was about 99 ppm. Four
specimens were attached to the test section in series. The linear
velocity in the tube specimen was varied from 0.6 to 5.3 m/s.
Sodium nitrite (NaNO2) or sodium pentaborate (Na2B10O16) was
added to the solution and the concentration was increased in a
stepwise manner. The change in the wall thickness of the specimen
was evaluated by the potential drop method[2].
3. Results
A clear inhibition effect of sodium nitrate on corrosion was confirmed at a low linear velocity. The
corrosion rate was decreased to 0.004 mm/y at a linear velocity of 3.2 m/s or less when the sodium nitrite
concentration was increased to 14.7x10-3 mol/dm3 or more. At a relatively high velocity of 5.3 m/s, although thecorrosion inhibition effect of sodium nitrate was unclear, the corrosion rate decreased to about 0.03 mm/y.
The corrosion inhibition effect of sodium pentaborate was confirmed at velocities of 0.6 to 5.3 m/s. The
linear velocity dependence of the corrosion inhibition effect was not confirmed.
References[1] Y. Fukaya, T. Hirasaki, K. Kumagai, T. Tatsuoka, K. Takamori, and S. Suzuki, Proc. 17th International Conference on
Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August, 2015, Ottawa.
[2] D. H. Lister, L. Liu, A. Feicht, M. Khatibi, W. G. Cook, K. Fujiwara, E. Kadoi, T. Ohira, H. Takiguchi, and S. Uchida,
PowerPlant Chemistry, 10(11), p. 659 (2008).

22

D03

Fundamental study for maintaining integrity of coolingwater circulation system: experimental examination ofcarbon steel corrosion under flow
Kazuki Kushida1, Shohei Sato1, Hiroshi Abe1 and Yutaka Watanabe1
1Graduate School of Engineering, Tohoku University

Abstract
In order to proceed the decommissioning of the Fukushima Daiichi Nuclear Power Plant safely, it isimportant to maintain cooling function and piping system for it, most of which is made of carbon steel. This
study was conducted to evaluate corrosion rate of carbon steel under flowing conditions by electrochemical tests
using rotating cylindrical electrode.
1. Introduction
For the present Fukushima Daiichi NPP, the degradation mode we need to pay the most attention is“corrosion” of key components made of metals. In order to maintain cooling systems, a technical package
including understanding rate limiting mechanism of corrosion and effects of key variables on the rate, accurate
prediction of wall thinning rate in the piping systems, and wall thickness monitoring at key locations, is needed.
The key factors would include flow velocity, dissolved oxygen, pH, chloride ion, gamma irradiation, sodium
bicarbonate, sodium pentaborate, and etc. This study aims to understand rate-determining process of carbon
steel corrosion under current and future Fukushima Daiichi NPP conditions as functions of flow conditions anddissolved oxygen level by using electrochemical tests with rotating cylindrical electrodes.
2. The study on the rate determining process based on the characteristics of the polarization curve
Polarization measurement was carried out as functions of flow conditions (rotative velocity of a cylindricalelectrode) and dissolved oxygen level, and the rate-determining process of the corrosion was investigated. The
obtained cathodic polarization curves were classified into four types based on comprehensively judging
presence / absence of diffusion limit current, influence of flow on cathode current density, and corrosion
potential. As an example, Fig. 1 shows the results of polarization measurement under different conditions of the
dissolved oxygen concentration with the rotation speed fixed at 2000 rpm. When the dissolved oxygen
concentration is as low as below 10ppb, since the dominant cathodic reaction is reduction of water, the chargetransfer determines the corrosion rate, so there is no influence of flow on the cathode polarization curve and alinear portion appears (Type 1). When flow rate or oxygen partial pressure was slightly increased, both reduction
reactions of water and dissolved oxygen together
contribute to cathodic current density. In thiscondition, diffusion-limited current density ofoxygen could not be identified (Type2). Whendissolved oxygen concentration is high enough andthe reduction of oxygen is the dominant cathodic
reaction, a diffusion-limited current density appears
in the cathodic polarization curve (Type 3).
Furthermore, when oxygen is sufficiently supplied
to the steel surface, both the charge transfer process
and the diffusion process of the oxygen reduction
reaction determine the reaction rate, so that thecathode polarization curve draws a curve (Type 4).
Also, it was confirmed that the corrosion potential
was lowered by the increase in dissolved oxygen
concentration or flow rate.
Fig.1 Results of polarization measurement
3. Conclusion
In order to understand rate-determining process of carbon steel corrosion as functions of flow conditionand dissolved oxygen level, polarization measurement was carried out using rotating cylindrical electrodes. The
results indicated the rate-determining process changes depending on mass transfer rate of oxygen and provide
database to develop prediction model of corrosion in the coolant circulation systems of Fukushima Daiichi NPP.

Acknowledgment
This study is the result of “Fundamental Research and Human Resources Development Program for Nuclear
Decommissioning related to Integrity Management of Critical Structures including Primary Containment Vessel and Reactor
Building, and Fuel Debris Processing and Radioactive Waste Disposal” carried out under the Center of World Intelligence
Project for Nuclear S&T and Human Resource Development by the Ministry of Education, Culture, Sports, Science andTechnology of Japan.

23

E01

SFP Fuel Removal at Fukushima-Daiichi NPS
Ritsuro Tokumori1,Masashi Fukushima1,and Akimasa Miyamoto1
Tokyo Electric Power Company Holdings,Inc.

1

Abstract
About 3000 fuel assemblies were stored in 4 Units at March 11 2011.We planned that they
were transported to the Common Pool and stored there. However, about 6400 fuels had alreadybeen stored in the Common Pool at that time. Therefore, Temporary Cask Custody Area (TCCA) was
newly built to store fuels that were in the Common Pool in order to ensure free space.
1. Introduction
On March 11, about 3000 fuels were already stored in 4 unit’s SFPs. So removing those from
SFPs and transporting to the Common Pool for safe-storage were intended.
2. Outline of fuel removal
2-1. Introduction
Unit 1, 3 and 4 R/Bs were damaged and FHMs and cranes became out of order. Therefore,
new installation of those after completion of rubble removal and sufficient decontamination was
required. For unit 2, R/B wasn’t broken but a radiation dose was high in the operating floor. So
first of all, needed to implement sufficient dose reduction and removals of walls and ceilings inorder to set up new FHM and crane.
2-2. Preparations of fuel removal
For unit 3 and 4, removals of rubbles and machines of operating floors were completed. For
unit 1, works are still in progress; investigation and removing process are now taken on theoperating floor. For unit 2, walls and ceilings of the operating floor will be removed.
2-3. Fuel removal
For unit 4, a new FHM and a new crane were installed, that equivalent to the previous
machines, and removed all fuels during period of time of Nov. 2013 to Dec. 2014. For unit 3, a new
remote control FHM and a crane are under construction to start removing fuels from 2018.

For

unit 1 and 2, currently we are examining to install new machine.
3. Related Facilities
3-1. Common Pool
About 6400 fuels had already been stored in the Common Pool whose storage capacity was
about 6800 fuels at March 11. After that time, about 1000 of them were removed to TCCA and about
1400 fuels were transported from unit 4 to there.
3-2. Temporary Cask Custody Area
TCCA was built after March 11. It is stored about 400 fuels which were stored in the Cask
Custody Building and about 1000 fuels which were stored in the Common Pool.
4. Conclusions
For unit 3, removal of fuels will be started in 2018, and part of fuels in the Common Pool will
be removed to TCCA accordingly. For unit 1 and 2, preparations of fuel removals will be proceeded surely
and safely.

24

E02

Unit-3 Covering Structure Plan for Spent Fuel Removaland Work Progress

Shinya Okada1, Kenichiro Go1, Takashi Inoue1, Norio Ryoki1, Taimei Hida1, Atushi Kikuchi1,
Kazuhiro Kato1, Miho Miyazaki1, Ippei Matsuo1, Kihei Ogawa1
1

Kajima Corporation

Abstract
Unit-3 covering structure plan for spent fuel removal is outlined: (i) reduction of radiation dose rate forconstructing cover structure, (ii) decontamination and shielding by remote control, (iii) design of cover structure,(iv) practice for fabricating cover structure and (v) constructing cover structure at Fukushima site.
Cover structure is currently under construction at Fukushima site.

1. Reduction of Radiation Dose Rate for Constructing Cover Structure
The rubble removal, decontamination and shielding were performed to reduce radiation exposure in
construction work of the cover structure for fuel removal.

2. Decontamination and Shielding by Remote Control
The decontamination and shielding were performed to further reduce dose rates since they were still higheven after removing the rubble.
1) Large concrete rubble was removed by Bucket attached 600ton-Crawler crane.
2) Small concrete rubble was collected by Dozer and removed by Vacuum.
3) Penetrated contamination was removed by Scabbler.
4) The shielding plates were installed by remote control to further reduce the radiation dose rates.

3. Design of Cover Structure1) The weight of the cover was reduced using dome-shaped steel frame.
2) The vertical and horizontal loads to the reactor building were reduced as much as possible.

4. Practice for Fabricating Cover Structure
We have practiced fabrication of the cover structure at the area with no radiation effects to steadily
construct it without reworking at Fukushima Site.

5. Constructing Cover Structure at Fukushima Site
Since the cover structure will be constructed partially by manned work, the reduction of dose rates are
crucial for the construction work at the site.

Fabrication of Roll Unit

Sliding of Roll Unit
25

Installation of the Last Roll Unit

H01

Study for the Bottom Access Route to remove the Debris from the
Nuclear Reactor Buildings
Atsuo YAMADA1, Hisashi IMAI1, Masanori IMAZU2, and Toshihiko
FUKUDA2

1

Radwest & Environmental Tech. Team, Nuclear Power Department, Hazama Ando Corporation

2

Technological Strategy Group, Nuclear Damage Compensation and Decommissioning Facilitation

Corporation
Abstract
It is important to remove the fuel debris from the primary containment vessel (PCV) and/or the reactor
pressure vessel (RPV) for the decommissioning of the Fukushima Daiichi Nuclear Power Plant. Authors
investigated the possibility of the construction of an access route for removing the fuel debris from thebasements of the nuclear reactor buildings.

1. Access from the bottom of the PCV
As results of investigation of conventional techniques, it is possible to construct the accessroute from the surface to the basements by using civil engineering methods ( a shield tunnelingmethod, an open caisson method, and so on ). It is considered that an all casing method will beable to make an approach hole through the base foundation of the buildings. However the
problems, such as sealing of bearings of boring tools, and so on, are left behind. A clear perspective
is not necessarily provided under the present conditions about such problems.
2. Conclusion※The range that can apply conventional techniques
Access route (Entrance⇒Upward shaft)

As described above, regarding bottom
access retrieval method, even if a route for thebottom-access can be established, a lot of
critical issues are involved in the subsequent
management of contaminated water and fuel

PNEUMATICCONTINUOUSCAISSON
UNDERGROUND WALLSHIELD METHOD

debris retrieval work. Therefore, since its
Entrance

feasibility was evaluated as low for a short or

Frozen
soil
wall
凍土壁

mid-term time period to be developed andverified, this option will not be focused in the

Access route
アクセスルート
立坑
Shaft

studies for the practical application.

アプローチ孔Approach hole

アクセスルート
Accessroute
斜路Inclined
shaft

Accessroute
上向き立坑
Upward shaft

References
アクセスルート
Accessroute
水平坑道
Tunnel[1] M. Imazu, T. Ito, T. Fukuda, A. Yamada: Study
for the access route to remove the debris from the

地下作業空間Basement
（地下室）
(Work
space)

basements of nuclear reactor buildings, proceedingsof the symposium on underground space vol.22,
C1-1, pp.137-142, JSCE, 2017.1.20.

PIPE JACKING METHODSHIELD TUNNELING

Figure 1. Outline of the bottom access route

26

H02

A proposal of the New Method to Retrieve Fuel Debris
with Geopolymer
Taichi Sakai, Takayuki Saimu, Shunichi Suzuki and Koji Okamoto(The University of Tokyo)

Abstract
In order to retrieve fuel debris from 1F plants, we need to overcome many difficulties. In this paper, analternative method where a variety of fuel debris are retrieved after grouting with geopolymer is proposed. Thismethod simplifies the control items during fuel debris retrieval and also has advantages on the waste management.
1. BackgroundFor decommissioning of 1F Plants, three methods to retrieve fuel debris such as a submersion method andin-air methods have been discussed [1]. These methods have their own merits and demerits, while the detailed
planning have not been decided yet [1]. In this paper, an alternative method to solve these problems [2] is discussed.
2. Geopolymer Grouting Method
2-1. Outline of Geopolymer Grouting Method
First, the water-stopping technology being developed by IRID isapplied for isolating PCV from S/C. Secondly, the grouting material,
geopolymer is injected on the MCCI products at the bottom of thePCV. Then the radiation-shielding concrete is put on geopolymer toreinforce the radiation shielding ability. After that, a platform isconstructed in the pedestal to retrieve fuel debris (Fig.1).
2-2. Advantages

Fig 1. An Image of the Platform

In this method, various effects are expected including stabilization of the MCCI products. For example,(a)isolation of PCV from the torus, (b)prevention of alpha-nuclides’ dispersion, (c)reduction of operational control
items and (d)long time storage because of less hydrogen generation are expected.
2-3 Research items
There are some issues to be considered in this method. One is the evaluation of the workability of geopolymer
and another is the heat removal of fuel debris buried by geopolymer with or without water cooling. Researches onthese items are currently going.
3. Conclusion
A new method to retrieve fuel debris is proposed, where there are multitudinous effectiveness and someaccompanying issues such as the workability of geopolymer and the heat removal with heat pipes.
References[1] NDF, “Technical Strategic Plan 2015 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo
Electric Power Company”, http://dccc-program.jp/files/20150630_1-2-J.pdf[2] S.Suzuki, “A new concept of the fuel debris retrieval by a hypothesis driven approach”, JAEA CLADS, Fukushima
decommissioning Platform, Nov 2016, https://fukushima.jaea.go.jp/initiatives/cat05/pdf/platform0401.pdf

27

Fiber Laser Decontamination Device
J01 Practical Uses of the CW Single-Mode
Technology
Eisuke J. Minehara,e-mail: minehara@olive.ocn.ne.jp

LDD Corporation (Laser Decontamination and Decommissioning Corporation)
A new continuous- wave (CW) single-mode fiber decontamination device[1,2] has beendeveloped to evaporate instantly with almost no temperature rise and to clean up the radioisotope (RI)
contaminated stainless steel-, other metals- and concrete- made components, devices and major
structural objects of nuclear power station, nuclear fuel reprocessing plants, accelerator facilities andthe related ones.

The laser decontamination device typically delivers the tightly focused laser light ofmulti- GW/cm2 just on the surface to clean up RI contained rust mold on and inside the stresscorrosion cracks and corrosion pitting in the stainless steel-made components, devices and objects.
The decontamination device is newly designed to measure 3-dimensional surface map very quickly,
and successfully to peel the 3-dimensional irregular surface according to the measured map using the 3
dimensional fast scanner, displacement meter and laser system. In the cold tests, the device performed
to peel off the surface of cans and multi-stepped plates very precisely, very uniformly and very
quickly.
The laser cleaning trials over several years were performed to decontaminate the 60Co andothers of RI heavily contaminated stainless steel samples cut from the pipes and plates in the primarycooling loop, and were successfully demonstrated to clean up using the decontamination device to be a
few tens of Bq/kg or less.

The heavily RI contaminated stainless-steel ones of 1MBq/kg were

performed to clean up using the device, and were successfully reached to be not detected in the final
trial year. The laser irradiated surface in the RI contaminated materials has the temperatures extremely
higher than sublimation and boiling points, therefore the irradiated and contaminated ones are blown
away by quick sublimation and instant evaporation, and steam explosion in the stainless steel, othermetal alloys and concrete ones.

Unlike the method using the CW single-mode fiber laser

decontamination device technology, other conventional methods using sand blasting, wet blasting,pulsed laser and laser gouging have no abilities to clean up completely the surface of the RI
contaminants so that some of the RI contaminants remain long time in the surface.
The CW single-mode fiber laser decontamination device technology has been developed overthese 10 years in the Japan Atomic Energy Agency, ATOX Co., Ltd., Nishi-Nihon Create Co., Ltd.,
Wakasa Industrial Technical Engineering Co., Ltd., the Wakasawan Energy Research Center, and LDD
Corporation.

The CW single-mode fiber laser decontamination device will be manufactured andsupplied in the first time in the world by LDD Corporation.
[1]E. J. Minehara, “Laser decontamination device”, US Patent No. US9174304B2, Nov. 3, 2015. EU Patent Under
Processing.
[2] E. J. Minehara, Laser Review, March, 2012, Vol.40, No.3, pp.165-170.

28

J02

Direct decontamination technique by fiber laser
Ryuichiro Yamagishi
The Wakasa Wan Energy Research Center

Abstract
A laser decommissioning technique was developed using focused continuous wave (CW) fiber laser light.
The laser decontamination setup is designed to peel the surface using fast scanner. In cold tests, the surface ofmortar test pieces with a depth of 0.1mm was peeled off at a rate of >1m2/h. The details of the laser
decontamination will be presented in the presentation.

1. Introduction
Laser decommissioning techniques have several advantages such as reduction of the generation of
secondary waste materials and remote controllability, compared with the conventional techniques such as
abrasive water jet, abrasive sand jet and mechanical peeling techniques. Further development of the laser
decommissioning technique is indispensable for removal of contaminated stainless steel and concrete surface.
The present research was undertaken to peel off mortar surface controlling laser power and scanning speeds.

2. Overview of our laser decontamination technique
Characteristics of our technique are irradiating focused laser lightfor surface. By focusing, laser light energy density will exponentially
increases.

This

focused

laser

light

causesvariousinteraction

instantaneously to peel off the surface of mortar. To scanning this laser
light, we can take contactless peeling effect in plane as shown in Fig. 1.
The characteristics of laser light showing contribution for our
decontamination technique are as drawing follows.
•Laser light is non-polluted.
•Easily take highly energy density by focusing.

Fig. 1. Pattern diagram of laser
decontamination•Restrictions by weight reactions and working distance are ignorable.
So our technique can decontaminate most targets in principle by these characteristics. To collecting peeledcontaminants only, the volume of radioactive waste is reduced. Furthermore, our technique can design freely
upon request of decontamination work site like miniaturization for narrow part decontamination and high-power
laser (kW class) usage for high speed decontamination.

3. Conclusion
From our experiment by prototype laser decontamination setup, this technique can peel mortar test piecesurface (as same as concrete surface) over 1 m2/h (0.1mm depth, calculated). In the presentation, we will show
our laser decontamination setup and peeling result by scanning laser.

References

29

J03

Sorting of contaminated soil based on radiological separation-experiences with a high performance mobile belt conveyor systemF. Langer1, Dr. M. Sokcic-Kostic1
1

NUKEM Technologies Engineering Services GmbH

Abstract
The existing, containerized, mobile belt conveyor system offers radiological characterization of soil withpotential contamination. A set of gamma spectrometers is used to sort the material and to direct it torespective waste streams and containers, thus reducing the volume for subsequent treatment. The system also
obtains information and generates documentation to accompany the characterized material in accordancewith regulatory requirements.

1. Purpose of the belt conveyor scanning systemRadiological characterization via gamma spectrometers

Classification of contaminated bulk material

Sorting and filling of the materialVolume reduction for subsequent treatment



Generation of sophisticated documentation per packageComparison with regulatory limits

2. Achieved benefits

Figure 1. Belt conveyor system on site

Volume reduction regarding expensive disposal routes

No secondary waste generated

No need for process media



Detection of Alpha-, Beta- and Gamma- emitters



Evaluation of spatial activity distribution

Proven toughness for construction site applicationFilling in containers according to customer’s requirements

Figure 2. General layout

3. Technical Data

Throughput up to

100 tons per hourBatch size

1000 kg Detection limit< 0.005 mSv/h

3x 40’ transport containers

Installation size

Figure 3. Detection system
3. Conclusion
This mobile belt conveyor system automatically sorts the soil into three different material streams (freerelease, limited release/utilization on site, radioactive waste) and can be optimized to customer’s needs.

References:

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

30

K01

Development of Polymeric Aluminosilicate Agent for Cation Removal
Takenori Masada1,2, Masahiko Matsukata2 and Atsushi Yamazaki2
1

Azmec Inc., 2Waseda Univ.

Abstract
Polymeric aluminosilicate agent for cation removal was developed by applying geo-polymer technique.
-1

Novel

2+

adsorbent had approximately 400 cmol(+) kg of cation exchange capacity, and high selectivity to Sr ion.
1. Introduction
Geopolymer is a hardening body synthesized by reacting calcinated kaolin or fly ash in a highly concentrated
alkali. Geopolymer is basically composed of poly-silicate network structure with SiO4-AlO4 tetrahedra crosslinkedby oxygen. The network structure has nano-size pore system, and contains cation exchange site derived tocompensate charge balance drived from 4-coordinated Al3+. We tried to develop a novel agent by applying
geopolymer to. This agent showed characteristics of ion exchange.
2. Characteristic of developed polymeric aluminosilicate agent
Developed agent was obtained by optimizing a wide variety of synthesis conditions. Generally, geopolymer
should have silica-rich compositions to obtain sufficient mechanical strength. Our novel adsorbent has a wide
range of aluminum-rich composition of SiO2/Al2O3 molar ratio 2.0 to 4.4. In particular,

suitable compositions

for cation removal use was SiO2/Al2O3 ratio < 3.0.
The agent developed showed approximately 400 cmol(+) kg-1 of cation exchange capacity being almost equal tothat of synthetic zeolite, and a high adsorption selectivity to Sr2+ and several minds of heavy metal cation. The
agent can be synthesized under low cost and energy saving conditions, at 20-100oC and atmospheric pressure.
3. Adsorption characteristics of Sr2+
Adsorption test was performed to evaluate of the adsorption characteristics of novel agent. Test method was asfollows: granule adsorbent in the range of 0.5 to 1.0mm was added into a strontium chloride aqueous solution or
the artificial sea water (100%) with the solid liquid ratio of 1: 100 or 1: 500. The test results after 24 h of vigours
strring was evaluated by using ICP-AES. Table 1 lists the test results, showing a high distribution coefficient of Sreven in the sea water containing high concentrations of Mg and Ca ions.
Table 1 Test Results of cation removal
Test solution
Strontiumchloride solution
Artificial sea
water (100%)solid liquid ratio
1:500
1:100

Concentration of Sr ［mg/L］
stock
9.23
After treatment
0.005
stock
7.21
After treatment
0.10

Removal Ratio ［%］

99.9
98.6

Distribution coefficient ［ml/g］

9.2 x 105
7.1 x 103

4. Conclusion
Developed adsorbent showed high cation exchange capacity and a high distribution coefficient of Sr2+ in solution,
resulting in a high selectivity to Sr2+..
References[1] “Ion Exchange Characteristic of Aluminosilicate adsorbent”, Takenori Masada, Masahiko Matsukata, Atsushi Yamazaki,
SCEJ 80th Annual Meeting, March 2017

31

K02
CRIEPI*NDF **

Environmental Risk Analysis for Radionuclides Transported in
Groundwater of Fukushima Daiichi NPS
T. Koyama, Y. Tanaka, D. Tsumune, T. Hijikata, Y. Tateda, T. Hattori
M. Imazu, M. Kamoshida, A. Hasunuma, K. Hida, A. Ono, T. Fukuda* Central Research Institute of Electric Power Industry**Nuclear Damage Compensation and Decommissioning Facilitation Corporation

Preliminary radiological risk analysis was carried out for the radionuclides assumed to betransported through groundwater in Fukushima Daiichi NPS. The result shows radiation dose will
be kept in very low, less than 1E-5 mSv/year, until 2111 even when contaminated PCV water isassumed to be leaked during debris retrieval.

Introduction
In many countries having contaminatednuclear facilities [1], “radiological risk
assessment” has been developed as a key
measure to decide actions for cleanup of
Fig. 1 Assumed pathway of contaminated watercontaminated sites and control of emissions
to the environment. As a preliminary risk
assessment, radiological risk analysis was
carried out for the radionuclides to betransported through ground water.
Analytical model
Most of the current countermeasures tomanage contaminated water were assumed
to be continued during decommissioning.
Source terms were the contaminated groundwater currently existing around reactor
buildings and the PCV water contaminatedwith cutting powder of fuel debris. Fig.1
illustrates the countermeasures and theFig.2 Behavior of radionuclides in ground water
assumed pathway.
(left) reference case (right) risk case
Results & discussion
Fig. 2 (left) shows the future of Sr-90 distribution in groundwater around reactor buildings.
As the water levels of drains around buildings are assumed to be kept below sea level
during decommissioning, Sr-90 plumes move from sea side to buildings. Then the plumes
move back to sea side because most of the countermeasures are assumed to be ineffective
after decommissioning. Public radiation dose through ingestion of sea foods contaminatedwith radioactive nuclides permeated through impermeable sea wall was calculated to be
kept very low, <1E-5 mSv/year. Because leaked radionuclides are collected by drains
within few years as shown in Fig. 2 (right), similar low radiation dose is expected whencontaminated PCV water leaked into groundwater.
References[1] A. Clark, “A risk-informed approach to decommissioning and remediation”, The 1st Int’l Forum on theDecom. of Fukushima Daiichi NPS.

32

K03

Continuous Challenges to Improve Performance, Availability andOperational Cost of Contaminated Water Treatment Systems for
Fukushima Daiichi

Teruki Fukumatsu1, Masaru Nishikubo1, Masahiko Kobayashi1, Keiji Ishikawa2, and Kei Kobayashi2
1

Toshiba Corporation, 2Tokyo Electric Power Company Holdings, Incorporated

Abstract
TEPCO and Toshiba have developed water treatment systems in Fukushima Daiichi to manage problemsof contaminated water produced by fuel debris cooling and ground water intrusion to buildings. Water
treatment systems consist of Cesium Reducing Device to utilize contaminated water as coolant of fuel debris,and Multi-nuclide Removal Facilities to remove radioactive nuclides except Tritium. Those devices andfacilities have been operated to manage contaminated water, and also have been improved their adsorbing
materials and operation condition to fit to changes of characterization of contaminated water.
Especially, the Second Cesium/Strontium Reducing Device (SARRYTM) developed by Toshiba treatedmore than 80% of contaminated water to be treated by Cesium/Strontium Reducing Devices, and theimproved SARRYTM is now planned to be installed as the Third Cesium/Strontium Reducing Device(SARRYTM -II). Also, the Multi-nuclide Removal Facility (MRRSTM) and Improved MRRSTM developed by
EnergySolutions and Toshiba have treated more than 90% of contaminated water to be treated byMulti-nuclide Removal Facilities. SARRYTM and MRRSTM use adsorbing process to simplify their process
and operation. TEPCO and Toshiba continue to develop the materials, adsorbents for the systems and theirprocesses in order to improve performance and availability, reduce the operational cost of the systems.
1. Outline of SARRYTMSARRYTM was developed and installed by Toshiba and it has been in
operation from June 2011, which followed the Cesium Reducing Device
installed by Toshiba, has treated about 170Million m3 of contaminated waterusing less than 170 m3 of adsorbent. It means the SARRYTM causes secondary
waste of less than 1/1,000,000 of treated water in volume. SARRYTM has been improved to fit to changes ofcharacteristics of contaminated water, by developing Cs/Sr simultaneous adsorbents, and by optimizing
schedule of replacing columns, etc. Cs/Sr simultaneous adsorbents, also reduces dose rate on the border ofFukushima Daiichi due to water treatment systems. Improved SARRYTM is planned to be installed by Toshibaand in operation in 2017, to reduce contaminated water of turbine buildings.
2. Outline of MRRSTM
MRRSTM has been developed and installed by Toshiba and it has been in
operation from March 2013. MRRSTM has treated more than 600,000 m3 ofcontaminated water and reduced the volume to less than 6000m3 in 2,400 of High
Integrity Containers. MRRSTM is a system consists of multiple adsorbents andeffective pretreatment process to retrieve multiple nuclides. We have been developing and improving thematerials, adsorbents and its processes.
3. Conclusion
SARRYTM and MRRSTM significantly contribute to manage water issues of Fukushima Daiichi. We will also
try to improve performance and availabilities, minimize the operational cost of the systems, amount of
secondary waste from waste treatment systems.

33

K04

Development of New Radioisotopes Selective
Adsorption Media in EBARA
Takashi Sakuma1, Makoto Komatsu1, Tatsuya Deguchi1,
Kaoru Kikuchi1, and Takeshi Izumii1
1

EBARA Corporation

Abstract
EBARA has been developing radioisotopes selective adsorption media for treatment of contaminated water
since 2011. During this period, 800 kinds of adsorption media were supplied from some manufacturers andthe radioisotope removal efficiency was tested. These media have been successfully used at NPS in Japanto process radioactive contaminated water and contribute the settlement of severe accident occurred in
March 2011. This paper introduces the media which has high removal efficiency of radioisotopes.

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

Figure 1 Cs/Sr Removal efficiency underseawater condition

Figure 2 Cs/Sr Removal efficiency under diluted
seawater condition

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

34

K05

Development of Decontamination Technologies for the Contaminated Water
in Fukushima Daiichi Nuclear Power Station

Takatoshi Hijikata, Koichi Uozumi, Kenta Inagaki, Takeshi Tsukada and Tadafumi Koyama
Central Research Institute of Electric Power Industry (CRIEPI)Abstract
CRIEPI has supported the operation of the decontamination system for high radioactively contaminatedwater in Fukushima Daiichi Power Station since the accident. Results of several tests and simulations clearly
explained the several phenomena that occurred on the actual columns of the KURION system. Moreover, thepermeable reactive barrier (PRB) of H4 tank area was installed using our recommended zeolite.

1. Introduction
CRIEPI has been studied the removal technologies of radioactive nuclides in molten salt by using zeolite.
After large earthquake in 2011, these technologies were applied to the various decontamination of the highradioactive contaminated water in Fukushima Daiichi Power Station.
2. Decontamination system for the high radioactively contaminated water
The equilibrium capacity of KURION media to absorb Cs wasdetermined by batch tests using mixtures of sea water and oils [1].
Adsorption kinetics in columns was verified by experiments of asmall-scale zeolite column [2]. The code was developed to simulate
the actual column system [3]. The kinetic parameters (film masstransfer coefficient and effective intra-particle diffusivity) in theactual column were estimated by combination of these column tests
and simulation code. The DF (Decontamination factor) of the actual

Fig.1

Simulation DF of Cs.

KURION system was examined by the simulation using the estimated kinetic parameter. As shown in Fig. 1, thecalculated DF value of Cs agreed well with the measured value.
3. Permeable Reactive Barrier for the contaminated ground water (H4 Tank area)
PRB has been introduced in Hanford and West valley sites in U.S.A. to remove radioactive nuclides in thecontaminated groundwater. CRIEPI revealed that the performance of Sradsorption on natural zeolite was better than that on apatite as shown in
Fig.2. The permeable reactive barrier (PRB) of H4 tank area wasrecommended the natural zeolite [4].
4. Conclusion
CRIEPI has supported to solve the various difficulties of thecontaminated water in Fukushima Daiichi Power Station.

1-gl・ 101
o
m
m
/
100
e
t
li
o
e
z
10-1
in
r
S
f
o 10-2niotart 10-3
n
e
c
n
o -4
c 10

Simulated groundwater
Na48,K22,Mg21,Ca28ppm

Natural zeolite (V/S=1000)
Apatite(V/S=100)

Gravel(V/S=10)muirbil 10-5
i
u
10-5
10-4
10-3
10-2
10-1
10-7
10-6
q
E
-3
Equilibrium concentration of Sr in solution / mmol・cm

References[1] T. Tsukada, et al., J. Nucl. Sci. Technol 51, 7–8, 886–893 (2014).

Fig.2 Absorption isotherm of Sr
in simulated groundwater.

[2] T. Hijikata et al., ibid ,51, 7-8, 894-905 (2014).
[3] K. Inagaki et al., ibid, 51 7-8, 906-915 (2014)[4] Tokyo Electric Power Company Ltd., http://www.tepco.co.jp/nu/fukushima-np/roadmap/images/d140627_07-j.pdf.

35

K06

Installation of Land-side Impermeable Wall
–Frozen-soil Walls at the Fukushima Daiichi Nuclear Power Station

Hiromitsu Kida1, Atsuhiro Fukada1, Isao Abe1, Taichi Esaki1, Shoji Seno1 and Teru Yoshida1
1Kajima Corporation

Abstract
Artificial frozen soil walls have been installed, as a measure against the increase of contaminated water,
surrounding four damaged nuclear reactor buildings at the Fukushima Daiichi Nuclear Power Station, Tokyo
Electric Power Company (TEPCO). Purpose of the Frozen-soil Walls is to cut off the groundwater inflow tothe buildings, which would become highly contaminated inside the buildings.

1. BackgroundAt the Fukushima Daiichi Nuclear Power Station, approximately 600m3 of groundwater has been pumped
up every day and stocked in tanks by TEPCO. Almost half of the above (300m3) is estimated to seep into thedamaged reactor buildings. To cope with this issue, the Committee on Countermeasures for Contaminated
Water Treatment, organized by the Ministry of Economy, Trade and Industry (METI), decided a basic policy inMay, 2013, i.e. the installation of the Frozen-soil Walls.

2. Construction
To form seamless Frozen-soil Wall, approximately 30m long vertical freezing pipes were installed into theground surrounding the buildings with an interval of 1m. As the freezing operation, brine solution, cooled
down to -30 degree Celsius, has been circulated through the freezing pipes, since March, 2015. Grouting
method was also applied partially to reduce the groundwater flow across the frozen wall being formed.
Total horizontal length of the completed wall is about 1,500m, which is definitely one of the largest
artificial frozen soil body ever made in the world.

3. Monitoring

Groundwater flow

To grasp the freezing behavior, groundtemperatures

have

been

monitored.

The

temperature sensors (optical fibers) were installedinside vertical tubes, every 5m along the Frozen-soil
Wall.

In addition, discontinuity of groundwater

Frozen-soil Walls

level along the Frozen-soil Wall has also beenchecked, using observational wells, to confirm thecompletion of the wall.

Figure 1. Concept of the Frozen-soil Walls

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

36

K07

Development of Micro-Polymeric Removerfor Uranium Ion
Miki Abe1, Michio Butsugan2, Shukuro Igarashi3, Yoshitaka Takagai1(Fukushima Univ.1, Hitachi Chemical Techno Service Co., LTD.2, Ibaraki Univ.3)
Abstract
Shiderophore chelate desferrioxamine B(DFB)-modified acrylic micro-polymer (MP) resins weresynthesized and uranium (Ⅵ) ion adsorption properties on the resin were determined. The polymericparticles which have different functional groups and three size regions of the micro particle (less 7, 7-12
and 45-90 μm) were used as mother material. The synthesized resin showed good adsorption ability ofuranium (Ⅵ) ion in weak alkaline solution. In addition, the investigation of absorbability of 61-elementson this resin was conducted. The result showed that it can collect some metallic elements of them.

Introduction
Since the Fukushima Daiichi nuclear accident, the contaminated water has been producing every day due
to the melted-fuel refrigerating. The contaminated water contains high quantities of uranium ion which iseluted from fuel.

Therefore, uranium remover for contaminated water has been required.

In this study,

we developed a micropolymers immobilizing desferrioxamine B(DFB) which has some hydroxamic acidgroups in molecular structure.

In addition, we investigated adsorption and desorption behavior of

uranium ion for the polymeric resin.

Experiments
Five milligrams of DFB-MP were added to plastic centrifuge tubes, and then 1 ppb metal ions solution(i.e. sample solution) was added.

The sample solution was adjusted to proper pH with buffer in advanceof the use. After the addition of sample solution, it was shaken using shaker under 210 rpm for 10 min.
Then, these were filtered using syringe filter and nitric acid was added. The solutions (filtrate) were
measured by ICP-MS.

Conclusion
DFB-EG80-2(S) which is a series of small particle size (7 ～ 12 µm) showed sufficient U(VI)absorbability(96%). Optimum adsorption was under pH 9 and temp from 40 to 70⁰C.

Uranium(VI) was

efficiently adsorbed compared with U(IV) in whole pH condition. In addition, highly adsorptions wereshown regarding V, Co, Ni and Ga. The Na2B4O7 buffer solution exhibited good adsorption efficiency ascompared with acetate. Regarding the capacity factor, we are going on the progress; however we foundthat 533 µg and more.

Desorption was successful using HCl solution (100%).

The project is supported by the MEXT, Human Resource Development and Research Program for
Decommissioning of Fukushima Daiichi Nuclear Power Station

37

K08

Purification of accumulated waste water and groundwaterin Fukushima Dai-ichi Nuclear Power Station
Yusuke Kitamoto1, Takashi Nishi1, Takashi Asano1
Kenji Noshita1, Shunsuke Miyake1, Masahiro Ichino1
Yuko Kani2, Toshimasa Ohashi2, Yutaka Sangu2
Ken Yamaguchi3, Kei Kobayashi3, Yuichi Kurosaki3, Keiji Ishikawa3
1
Hitachi-GE Nuclear Energy, Ltd.
2
3
Hitachi, Ltd. Tokyo Electric Power Company Holdings, Inc.

Abstract
We developed three types of water treatment system and are contributing to reduce theradionuclides in contaminated water and ground water in Fukushima Dai-ichi Nuclear Power
Plant Station. Here we developed the simulation code for Sr-90 behavior prediction that assists
stable operation of the systems. Also, we established the technology to remove Sr-90 in RO-brinefrom the order of 108Bq/L to 1Bq/L under the existence of interfering substance.
We developed the facility “Sub-Drain Treatment System” (SDTS) to treat groundwater aroundthe buildings called “Sub-Drain”. The groundwater contains radionuclides such as Cs-137 andSr-90 with relatively low concentration (100Bq/L). SDTS utilizes adsorption technology to reduces
the level of Cs-137 and Sr-90 less than the control limit for ocean release. It is featured by the
special adsorption media to remove both Cs and Sr at the same time. It has already treated andreleased 32,000m3 of groundwater, and is still in operation.
Furthermore, Hitachi-GE applied this technology to purify the accumulated waste water(RO-brine generated from Reverse Osmosis (RO) system treating building water) stored in theyard of Fukushima Dai-ichi, which contains much higher Cs-137 (105Bq/cm3), Sr-90(108Bq/L) andother nuclides. One is “Modified Sub-Drain Treatment System for RO brine treatment” (Sub-RO),
which removes 99% of Sr-90 for the dose reduction of stored water. The other is “High performance
multi-nuclide removal system” called “HERO”, which removes Sr-90 almost completely into thelevel of 1Bq/L. These systems contributed the risk reduction of Fukushima Daiichi. In this report,
we introduce the feature of those systems.
Also, we introduce the simulation code developed for the operation planning of SDTS and
Sub-RO. It enables to predict the behavior of Sr-90 concentration at outlet of adsorption vessel. It
shows operator when to exchange the vessel in advance, and contributes stable operation of the
facilities.
In addition to those achievement, we also introduce the typical difficult feature of Sr-90, which
was found in the operation of HERO. It was found that small content (0.1%) of Sr-90 exists as thechemical form not removed simply by adsorption media. Hitachi-GE researched the cause of thisissue and found that Sr-90 forms some complex with the reagent used in RO system. Therefore,
Hitachi-GE developed the way to dissolve the complex and remove the reagent so that Sr-90 stays
the form of Sr2+ that can be removed by adsorption media. By the use of this technology, HERO
achieved the reduction of Sr-90 less than 1Bq/L and contributed to reduce the risk accompanied by
the storage of highly contaminated water in the yard of Fukushima-Daiichi.

38

K09

Fabrication of Porous Ceramics using Natural Minerals
for Purification of Radioactive Water
Shuhei Toike1, Katsumi Yoshida1, Toyohiko Yano1, Tadahiko Takada2
1

Tokyo Institute of Technology, 2 Silica Material LLC

Abstract
After the accident of Tokyo Electric Power Fukushima Daiichi Nuclear Power Plant caused byearthquake on March 11, 2011, the treatment of radioactive water in the power station port has been a serious
problem. We fabricated porous ceramic adsorbents using natural minerals for purification of radioactive water,
and adsorption of Cs with the raw powders and the porous ceramics was evaluated. Porous ceramic adsorbents
prepared in this study showed good adsorption performance for Cs.
1. Introduction
To converge the accident of Tokyo Electric Power Fukushima Daiichi Nuclear Power Plant, thetreatment of radioactive water in the power station port has been a serious problem [1]. In this study, porous
ceramic adsorbents for radioactive Cs and Sr were fabricated using natural minerals, which show higher
adsorption for Cs and Sr, to improve handling ability during treatment of radioactive water and to simplify theimmobilization process after treatment. In addition, effects of the raw materials used for the porous ceramic
adsorbents on their adsorption ability for Cs were evaluated.
2. Experiment
Green compacts (20 mm, height: 10 mm)
Samplewere formed by uniaxial pressing using natural porous
silica (hereinafter referred to as A-WPS), naturalzeolite (NKZ) and diatomaceous earth (TPS) with 10
A-WPS
wt% natural clay (NC). The compacts were
NKZ
pressureless-sintered at 500℃ and 700℃ for 2 hours
TPS
in air. Surface microstructure was observed with aNCscanning electron microscope (SEM). Specific surface
area and pore diameter distribution of the raw powders andsintered bodies were measured by nitrogen gas adsorption
amount measurement method. Adsorption batch test of
powders and sintered bodies for Cs was carried out using
simulated radioactive water prepared by adding CsCl in purewater. Adsorption of Cs with the raw powders and the sinteredbodies (porous ceramics) was evaluated.

Table1. Specific surface areaof raw powders and sintered bodies
Specific surface area(m²/g)
Raw powder

Sintered body

Sintered body(at 500℃)(at 700℃)

108.1

112.8

106.3

19.8

27.5

19.8

39.0

39.7

34.5

39.9

-

-

Fig1. Relation between adsorption time and Cs
adsorption of porous ceramics sintered at 700℃.

3. Results

Table 1 shows the specific surface area of raw
powders and sintered bodies obtained from BET plots. The
specific surface area of A-WPS and porous ceramics with A-WPS showed the highest values. Specific surface
area of the porous ceramics decreased with increasing firing temperature.
Figure 1 shows the relation between adsorption time and Cs adsorption of porous ceramics sintered at
700℃. In this test, pH was adjusted to be 7, liquid-solid ratio was 10, and initial concentration of Cs was
adjusted to be 1 ppm. All samples showed Cs adsorption of 80% or more after 2 hours. These results indicatedthat the porous ceramic adsorbent fabricated in this study showed good adsorption performance for Cs.
References[1] Tokyo Electric Power Company Press Release, November 19, 2015 "Current Status and Future Measures of TEPCO
Fukushima Daiichi Nuclear Power Station"

39

L01

SIAL® :Geopolymer solidification technology approved by
Czech / Slovak Nuclear Authority
Milena Prazska1, Marcela Blazsekova1 and Hisashi Mikami2
1

Amec Foster Wheeler Nuclear Slovakia s.r.o., 2Fuji Electric Co.,Ltd.

Abstract
We present the features and performance record of the geopolymer solidification technology, SIAL®,
licensed for use by both the Slovakian (ÚJD SR) and Czech Nuclear (SUJB) regulators. An overview of its
performance at Bohunice in Slovakia, where it has been used to successfully treat resins and sludge wasteforms
will be provided along with references to other plants where it has been used. More recently, geopolymers havebeen noted as an immobilization technology which shows potential for resins produced by treatment ofradioactive contaminated water at the Fukushima Daiichi Accident Plant. Reference will also be made to someof the activities being undertaken in Japan to demonstrate its performance.
1. Introduction
The Nuclear Power Plant A-1 located in Jaslovske Bohunice, completed in 1972 operated for 5 years until
two accidents in 1976 and 1977.

After the second accident in February 1977 (INES level 4) the Nuclear Power

Plant was permanently shut down for decommissioning. Fuel assemblies and fuel cladding damaged in theaccidents led to significant strontium-90, caesium-137 and transuranic contamination. As a consequence of thelong-term corrosion of the barrier’s material, highly contaminated sludge accumulated which could not beeffectively immobilized using conventional methods such as cementation or bitumen treatment due to adverse
physical-chemical properties and high specific activity of the waste (caesium-137).
Foster Wheeler developing and licensing the geopolymer, SIAL®.

This challenge led to Amec

Today, SIAL® is one of the most provenwidely used geopolymer for on-site solidification of radioactive materials such as sludge, resins, sorbents andorganic liquids.
2. Features of Geopolymer solidification material
SIAL® can provide efficient and practical on-site
treatment of radioactive waste streams at room
temperature and can incorporate on average four times
as much waste as a cement matrix equivalent.

Theequipment used to deploy SIAL® is also modular,
flexible and versatile.

It can encapsulate waste

quicker than cementation and can even set under water.

Figure 1

solidification process

In addition, it is characterized by excellent mechanical andphysical properties, compared with the earlier generation
techniques. This includes higher mechanical strength, lower
leachability, low volatility, posing a low fire risk and excellent
physical stability in the presence of frost and water (nodistortion or cracking).
3. Performance record exmaple
About 750 m3 of radioactive waste (spent resins,

Figure 2

sludge and borates) stored in 14 tanks situated in auxiliary building of

SIAL® matrix with resins

the V1 NPP in Jaslovske Bohunice. The

waste was characterized into two types of waste, namely resins and crystalline sediment and sludge. This
comprehensive scope of work started with the licensing process, solidification and then was followed by thedecontamination and cleaning of the workplace post clean up and transport of all equipment off site.
This is directly applicable to the conditions at Fukushima Daiichi (as well as other Japanese NPPs) where thewaste streams are not well understood.

40

L02

Volume reduction and recycling of contaminated metal by melting
Satoshi Karigome1, Masaru Kawamata1, Takahiro Matsuda1,
Toru Kikuchi2 and Hiroki Shimoda2
1

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

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

Radiation Shield

Bench

Block

Intrusion Assessment of contaminated wastes fromdecommissioned NPPs for clearance andrecycling. (Figure 2)

Figure 1. Dismantling of Fuelprevention block
handling Machine Figure 2. Recycled products

2. Background of JFEE Construction and operation of temporary
incinerators for disaster waste, including
radioactive substances, in Fukushima Pref.
(Figure 3)
 Design and supply of ash melting furnaces forgeneral waste incineration ash.
3. Melting of radioactive contaminated metal Volume

reduction

of

scrap

metal

contaminated with radioactive cesium (Cs) bymelting, and separation of radioactive Cs
Figure 3. Incineration plants for radioactive
disaster wastes in Fukushima Pref.

from the metal.
 Assessment of decontaminated metal for
clearance and recycling.






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

Figure 4. Process flow of radioactive Cs separation

41



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

Figure 5. Conceptual design ofthe equipment

L03

Current status and issues about secondary waste from water
treatment
Kosuke Sato1
1

Tokyo Electric Power Company Holdings, Inc.

Abstract
There are many kinds of secondary waste from water treatment and they should be stabilized to reduce
potential risks. It is necessary to solve technical problems for each waste but some of them have yet to beundertaken. In this poster, current storage management status and issues for stabilization treatment aresummarized.

1. Recognition of risks
・Waste from water treatment is sometimes stored temporarily with its moisture content, and therefore, there isan potential risk of leakage of radioactive materials due to corrosion of the container or generation of hydrogenetc.
・It is necessary to keep risks low enough during long term storage in 1F until the period of disposal.

2. Current status
・Waste from water treatment is temporarily stored outdoors.
・The approximate amount of waste that needs to be treated is shown in Fig.2.
・Many kinds of media have been generated due to changes of media for improving Decontamination Factor.

Fig.1 Overview of water treatment system

Fig.2 Quantity to be treated (quantity indicates rounded
off figures for end of 2016, only major substances)

3. Issues
It is difficult to develop an optimal treatment strategy from the following reasons.
・Stabilization such as dehydration, drying and solidification is necessary but the applicable technologies aredifferent depending on their radioactivity level, physical properties and chemical components.
・For rational treatment, it is necessary to think about a method of treating all waste with as few equipment as
possible.
・The properties of secondary waste from water treatment are often unclear because sampling is difficult.
・Remote handling is necessary because of impossibility to approach high-dose secondary waste.
4. Introduction of effort
Our efforts to address these issues will be introduced on the day of the poster session.

42

L04

Examination of slurry stabilization technology
Toshio Nakazawa1, Kohhei Aikawa2

1Tokai Office,2Technical Development Center

ATOX CO.Ltd
Rotate１．Abstract
The Fukushima Daiichi Nuclear Power Plant treats the water
including radionuclides by Advanced Liquid Processing
Systems(ALPS).ALPS discharges the secondly waste called“slurry” and it includes a lot of water. Therefore , there are risks

Discharge port ofof leaking and pollution of the environment. This study

Dewatered slurry

Feed Pipe of Slurry

researched and examined a dewater technology as a stabilization

Scraperof the secondary waste.
Photo１: CD (Compact disk）Dryer

２．Introduction
１） Technical investigation
As basic study , simulated slurry was prepaired and dewater
processes using it was evaluated in laboratory scale. The slurry
particle size and moisture contents were evaluated after dewater
tests.

Figure1：Operational Overview（CD Dryer）

２） Selection requirements
The requirements of treatment processes were examinedbased on the characteristics of the slurry and the restrictions such

Filter clothas throughput, treatment result, processing properties,
dewatering performance, radiation reduction, scattering prevention,
secondary waste generation, etc.

Press Filter
Discharge port of
Dewatered slurry

Photo2： Filter Press

３）Machine selection
Two type dewater machines in general industrious filed wereselected from investigation results.

３．Results and Discussion
As a dewater device of the general industry by the real scale cold
test, CD (Compact disk） dryer and Filter press (Photo2) were foundto be suitable.

Figure2：Operational Overview（Filter Press）

43

L05

High-frequency melting technology of radioactive metal waste
Takeshi Nishikawa1, Nobutake Horiuchi1, Naoto Sasaki1, and Hirokazu Tanaka1
1

Mitsubishi Materials Corporation

Abstract
Mitsubishi Materials Corporation has been developing technology concerning the high-frequency meltingof radioactive metal waste, for the purpose of volume reduction, decontamination and recycling of metal waste
contaminated with radioactive materials from NPP and nuclear facilities.
1. Introduction
In the melting of metal waste, a high volume reduction effect can be obtained. And radionuclides such as
cesium (Cs), strontium (Sr) and α nuclides can be decontaminated from metals by melting using a highfrequency induction furnace. In addition, the melting process is also suitable for the recycling of metal waste,because it can be homogenized and cast into any shape. We focused on these advantages of melting processes
and have been developing technologies. The development results are shown below.
2. Development results
2-1. Volume reduction
By the melting, it is possible to make metal waste into avoid free ingot, and it is confirmed that the volume of metal
waste can be reduced to about 1/5 to 1/3.
2-2. Decontamination
The boiling point of Cs is lower than the iron melting
temperature (1600deg-C), so iron-based metal waste can be
decontaminated because Cs volatilizes during metal melting. Sr,uranium (U) and plutonium (Pu) are elements that are morestable as oxide than iron (Fig. 1). Therefore, in theelectromagnetic field of the high-frequency induction furnace,
the oxide (including nuclides) is ejected outward without theFig.1 Oxide formation free energy ofaction of electromagnetic force, and can be decontaminated, while
main metals
metal is moved to the center of the furnace by electromagnetic force※The Japan Society of Thermometry(Fig. 2). In addition, the metal has good homogeneity because
Calculated by thermodynamic data calculation
software "MALT 2"
molten metal is stirred by electromagnetic force.
Melting decontamination test using real metal waste contaminated withradioactive Cs caused by nuclear accident was carried out using our small
Moltenmetal
induction furnace (1 kg / batch). As a result, it was confirmed that real metal
Electrophoretic[1]
waste can be decontaminated to less than the detection limit (0.2 Bq / kg) .
force
Oxide

r
re
nt
e
C

r
e
ut
O

In addition, melting decontamination tests using simulated
nuclides (Cs [1] and Sr) were carried out using the small induction
furnace and our demonstration scale induction furnace (1 ton /batch, Fig. 3). We confirmed the migration behavior of these simulatedElectromagnetic
force

nuclides, and confirmed that decontamination is possible even on theFig.2 Behavior of oxide in highdemonstration scale. Furthermore, it was confirmed that U was
frequency induction furnace[2][3]
decontaminable by melting decontamination tests using U.
2-3. Recycling
We have been developing technology to manufacture metal waste
containers by centrifugal casting method, because this method has few
secondary waste, is easy to manufacture, and is suitable for remote operation.
By using the centrifugal casting method, it is possible to manufacture acylindrical container without a core mold. It was adopted for advanced
volume reduction facilities of JAEA.
If metal waste is made into ingots, they can be processed into various
products later. Melting technology may be effective as a means to reduce andrecycle radioactive metal waste generated at TEPCO’s Fukushima Daiichi
NPP.
3. Acknowledgments
The development of melting technology of radioactive metal waste
Fig.3 Implementation status ofdemonstration scale test
introduced here was carried out under the guidance and cooperation ofmany people, demonstration project of MOE, Basic research programs for the next generation vitrification
technology of METI, JNC, RWMC, effective use project of MITI, and so on, in addition to our own technology
development. Thank you deeply.
References[1] 2015, MOE, FY2010 decontamination / volume reduction technology selection and evaluation work.
[2] 2002, JNC, JNC technical report No.14.
[3] 2004, RWMC, Uranium waste decommission technology investigation.

44

L06

New Radioactive Waste Including Slurry Treatment Systems
by Compact Disc Dryer
Hiroyuki Hayashi1, Hirofumi Inagawa1, Takashi Otsu1,
Takashi Sakuma1, and Takeshi Izumii1
1

EBARA Corporation

Abstract
EBARA has been developing treatment System for radioactive waste by CD DRYERS. (Nishimura Work
Co., Ltd.)With that system, radioactive waste including slurry is reduced in volume and separated into

solid matter and water. CD DRYERS belong to the category of conductive heating dryers which indirectly
heat and dry materials through conduction.

1. Construction

Figure 1

Construction

2. Flow of drying operation
2-1 The material in material tank is supplied(1)

to the circulation tank. (1)(2)

2-2 The material in the circulation tank is sprayedover the disc surface by the circulation pumpthrough the header. (2)(3)

2-3 The material deposited on the disc surface
is scraped off

by means of a scraper

as it is dried while the disc is rotating. (3)3. Features
3-1. High performance
3-2. Compact
3-3. Energy saving / High efficiency
3-4. Less thermal decomposition of material
3-5. Curtailment of process

4. Application
Radioactive waste, Waste resist, High concentrated salt water, Waste water-soluble oil and so on.

45

L07

Research&Development of Radioactive Waste Long-Term Storage
Containers with Passive Autocatalytic Recombiners
Kazuyuki Takase1, Nguyen Huy Hiep1, Yota Suzuki1, Yusei Tanak1,
Isamu Kudo2 and Ryutaro Hino3
1

Nagaoka Univ. of Technology, 2ADVAN ENG. Co., Ltd., 3Japan Atomic Energy Agency

Abstract
In the decommissioning of Fukushima Daiichi nuclear power plant (1F), the long-term management of fueldebris is necessary. In this process, hydrogen which is the flammable gas is generated in storage containers bydecomposition of water due to radiation. Then, in order to reduce the hydrogen concentration in a long-term
waste storage container, use of passive autocatalytic recombiners (PAR) was considered.

1. Introduction
The objective of this study is to confirm experimentally that PAR functions effectively, even if thecontainer which stores radioactive waste, such as the fuel debris, for a long period of time includes lots of water.

2. Experimental Plans
A high water concentration experiment was planned. In this experiment, the condition that much water ismixed with radioactive waste into a container was assumed. Two kinds of experimental devices with different
size were set up in consideration of the TMI-2 results [1]. One is a large cylindrical container with a diameter of
35 cm and length of 200 cm, and another is a small cylindrical container with a diameter of 16 cm and 50 cm.
Each container has simulated radioactive material and PAR in the inside as can be seen in Figure 1. Figure 2
shows appearance of the small cylindrical container. The inside can be confirmed in visible from the windows
installed in the side wall. The procedure of the experiment is as follows: on the condition that water is enclosed
in a container, hydrogen gas flows into the container from the bottom of that or the side; the oxygen in thecontainer and the hydrogen gas changes to vapor by catalyzed reaction due to PAR; and, the experiment isrepeated changing the amount of the
16 cm

accumulated water. That is, it isconfirmed

that

PAR

Vapor

functionsPAReffectively, without depending onlevel in the container.

Water
level

50 cm

the amount of water and the water

Hydrogen

3. Conclusion

Windows

The present study is scheduled

Simulatedradioactive
material

to be begun from the summer of
2017. The results will be reflected tothe decommissioning of 1F.

Hydrogen gas

References[1] J. Henrie, et al., GEND 051, 1985.

Figure 1. PAR and simulated
radioactive material

46

Figure 2. Small cylindricalcontainer

L08

Study on Post-accident Waste Management Scenarios for Fukushima
Daiichi Nuclear Power Plant
Subcommittee on Radioactive Waste Management
Study Committee on Decommissioning of the Fukushima Daiichi NPP
Atomic Energy Society of Japan

Abstract
In the decommissioning of the Fukushima Daiichi NPP, the radioactive waste management plan needs to
be integrated with the plans for the series of activities from post-accident management to decommissioning andsite remediation. Optimization of radioactive waste management (WM) scenarios is essential in preparing thedecommissioning/remediation plans. This study aims to semi-quantitatively evaluate potential WM scenarios for
multiple potential interim and final end-states of decommissioning/remediation. The outline of the study ispresented in this poster.
1. Introduction
Atomic Energy Society of Japan has organized Study Committee on Decommissioning of the Fukushima
Daiichi NPP consisting of several subcommittees, each focusing on a different area of expertise. One of thesubcommittees, started in 2016, considers the management of radioactive waste generated from a series ofdecommissioning and site remediation activities. The objective of the subcommittee is to identify therequirements and constraints on decommissioning and remediation of Fukushima Daiichi NPP from the WM
perspective.
2. Strategic Considerations
One of the first tasks is to review the reports published by international organizations on the experience of
decommissioning and environmental remediation after nuclear accidents (e.g., [1] which emphasizes theimportance of long-term strategies). Following the lessons from the past accidents, potential interim and final
end-states in Fukushima Daiichi Decommissioning Roadmap [2] are discussed, as a basis to construct WM
scenarios.
3. Scenario Analysis
The interim and final end-states define the goals for each step of post-accident clean-up, stabilization,
decommissioning, and site remediation, and will facilitate the estimation of the quantity of waste generated in
the process. Up to 2016 when post-accident clean-up and stabilization activities are in progress, more than
400,000 m3 of wastes, including contaminated soil, trees, and rubbles, have been generated. A large amount ofradioactive waste will be produced by dismantling the reactor facilities. There are numerous potential
alternatives to manage this waste (e.g., treatment technologies, locations of storage and disposal sites). Choicesof WM alternatives may influence the progress of decommissioning/remediation activities because, for example,
storing waste on-site or constructing waste treatment facilities may limit the work area for decommissioning.
The WM scenarios are being formulated as a combination of the possible alternatives.
4. Future study
The end states as well as the scenarios will be discussed and evaluated thoroughly and carefully in terms offactors such as feasibility, radiation safety, technology availability, consistency with the national framework.
References:[1] IAEA, Decommissioning and Remediation after a Nuclear Accident, 28 January–1 February 2013, Vienna, Austria.
[2] Inter-Ministerial Council for Contaminated Water and Decommissioning Issues, June 12, 2015.

47

L09

A pyrochemical treatment concept of combinding selective
fluorination and molten salt electrolysis for fuel debris

Haruaki Matsuura1, Takafumi Uchiyama1, Takahiro Ono1, Masatoshi Kajimoto1, Yu Sakane1,
Tadayoshi Sato1, Tatsuya Emori1, Atsushi Nezu2 and Nobuaki Sato3
1

Tokyo City Univ., 2Tokyo Inst. Technol., 3Tohoku Univ.

Abstract
An on-site pyrochemical treatment process for fuel debris has been proposed. The procedure proposed is:1) selective fluorination, 2) selective dissolution into molten salt, and 3) molten salt electrolysis. By thethermodynamic studies on HF fluorination of uranium - zirconium oxide mixtures and other structural materials,
depending on composition and surrounding environment, fluorination condition should be selected.

1. Introduction
In the damaged reactors in the FDNPP, fuel debris is expected to exist in the variety of chemical form.
After the struggling of taking out fuel debris from damaged cores, partitioning and stabilizing process would be
required. We have focused on pyrochemical technology as one of the best solutions, because of its compact
concept. The poster describes the concept in three steps and fundamental R&D studies performed for each step.

2. Concept of pyrochemical process in three steps
2-1. Selective fluorination
As the first step, fluorination treatment by dry hydrogen fluoride gas has been proposed, since uranium andtrans uranium elements should not be vaporized as hexavalent form, but kept as tetravalent form. In this step, thebest condition of selective fluorination of uranium must be found. According to thermodynamic studies on
various composition of uranium – zirconium oxide under the treatment of dry HF gas, pre-treament by hydrogengas reduction is considered to be very effective to selective fluorination of uranium. However, if the fuel debris
contains some metallic compounds, to avoid the aggressive reaction with HF gas, pre-oxidation treatment
should be added.
2-2. Selective dissolution into molten salt
Before the molten salt electrolysis, partially fluorinated fuel debris is introduced into molten salt, anddissolved selectively. In this process, selective dissolution of fluorinated uranium would be ideal. By using
zirconium, dissolution behavior of its fluoride and oxide mixture into LiF-NaF-KF eutectic has been
investigated. It is considered that coexisting oxide restricts the rate of dissolution of fluoride into molten salt.
2-3. Molten salt electrolysis
In the final step, dissolved uranium can be reduced electrochemically in the molten salt bath. According to
the electrochemical studies on zirconium in various chloride, fluoride, and oxide systems, control ofconcentration of each anions is found to be the most important factor to deposit selectively targetted material.

3. Conclusion
These studies have been launched since 2013. We are planning to continue the studies using uranium in
molten salt near future. This study has been performed by the JSPS Kakenhi JP 25630429 and 15K06665.

48

L10

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

KOBELCO STUDSVIK Co., Ltd. was formed in Japan on July 1st, 2016, jointly by Kobe Steel Ltd.
and Studsvik AB, focusing on delivering design, engineering and implementation to provide
innovative radioactive waste management solutions to the Japanese nuclear industry. KOBELCOSTUDSVIK will provide technology and processes to safely reduce radioactive waste andeffectively recycle metals for the decommissioning of the Fukushima Daiichi Nuclear Power
Station.

1.THOR treatment technology

THOR (THermal Organic Reduction) is
proven thermal treatment technology forprocessing of LLW and ILW, with pyrolysis
and steam reforming mechanism. In Fig. 1,
THOR process flow is shown.
(1) Wastes
Ion
exchangeresins,rubbers,
activated carbon, plastics etc. areapplicable wastes.
(2) Reformed Residue
Radioactive wastes are converted into
dry and stable solid residue product(Reformed residue).
The reformed residue in whichradioactive nuclides are captured isinsoluble and leach-resistant.
(3) Volume reduction
When spent ion exchange resin istreated, its volume is reduced up to 1/15 ,depending on waste type and final waste
form.

Fig. 1 THOR treatment process flow

Fig. 2 Metal Recycle system flow

2. Metal Recycling

In Sweden, where the effective reuse ofresources has socially taken root, metalmaterials of nuclear power equipment aresafely recycled. The Studsvik model, which has
been experienced in compliance withinternational rules and regulations since theFig. 3 Photos of metal recycling process
mid-1980ies, will be able to respond to theincreasing demand for metal recycling in Japan . In Fig. 2, metal recycle system flow and, in Fig. 3,
photos of metal recycling system are shown respectively.

49

M01

Human resource development by creative robot contest fordecommissioning
Shigekazu Suzuki 1
1

National Institute of Technology, Fukushima College

Abstract
For human resource development related to Fukushima Daiichi decommissioning, it was held to plan thefirst time the Creative Robot Contest for Decommissioning.
and ability to identify challenges type.

It was held as robot contest of problem-solving

Naraha Summer School was held in September 2016 and the students

visited Fukushima Daiichi and JAEA Naraha center.

Human resource development through the robot contest isreported

1. Overview of Creative Robot Contest for Decommissioning
In order to get the young generation interested in the decommissioning, education throughthe creations of robots is believed to be effective.

In addition, a variety of practices, such as PBL

education and active learning lesson are offered in each technical college.
also carried out in PBL and active learning.

This Robot contest is

If the main purpose is to interest students indecommissioning through the manufacturing of robot it also aims at cultivating the students andcontributing to their "creativity education", "problem-solving ability" and "ability to identify
challenges".

Overview of the competition challenges are as follows.

1) Field
To select the playing field in each team from the two fields of the following assumes the reactor building.
a) Mock-up stairsb) Standard step field

2) Field environment
Mock-up stairs and standard test field both have the following
environment features:a) It is no darkness lightingb) Not be able to face up to the body to operate the robot by remotec) Radio wave does not reach because there is a thick wall of concreteFig. 1 Mock up stairs

3) Challenges of robota) Mock-up stairs- Carry a 5 kg luggage up to the second floor from the first floor andreturn to the original location by placing the luggage.
- Examine the thing that is put on the second floor.

It should be

noted that the location is undefined.
b) Standard step field- Examine the field of shape (such as area and irregularities).

Fig. 2 Standard step field- Determine objects located in the field. It should be noted that their locations are undefined.
There were submitted from all over the country in the national and public technical colleges of the 13
colleges 15 team in this contest.

Results of document screening, was observed participation of 15 teams.

50

M02

Advanced Research and Education Program for Nuclear
Decommissioning

Toru Obara1, Toyohiko Yano1, Kenji Takeshita1, Yukitaka Kato1, Hiroshi Akatsuka1, Hiroshige Kikura1,
Takehiko Tsukahara1, Katsumi Yoshita1, Koichiro Takao1, Jun Nishiyama1, Masayuki Harada1,
Nobuyuki Iwatsuki1, Koichi Suzumori1, Gen Endo1, Kenji Kawashima2, Noriko Asanuma3, Tsuyoshi
Arai4, Naoyuki Takaki5
1

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

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

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

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

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

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

51

N01

Current Status of the JAEA’s R&D Bases for the Decommissioningof the Tokyo Electric Power Company Holdings, Inc. Fukushima
Daiichi Nuclear Power Station
Yoshihiro Tsuchida1, Makoto Ooka1, Toshiro Nakai1 and Shinichi Nakayama1
1

Japan Atomic Energy Agency

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

1. Introduction
Restoration of the environment and decommissioning of the TEPCO Fukushima Daiichi NPS is positioned
as Japan’s national issues, and response to the accident is a prioritized mission of JAEA. In Fukushima
Prefecture, JAEA has acting at the Collaborative Laboratories for Advanced Decommissioning Science(CLADS) in Tomioka Town and the Naraha Remote Technology Development Center in Naraha Town, andconstructing the Radioactive Material Analysis and Research Facility in Okuma Town (Okuma Analysis andResearch Center).

2. JAEA’s R&D bases for the decommissioning of the Fukushima Daiichi NPS
2-1. Collaborative Laboratories for Advanced Decommissioning Science
CLADS is the core of JAEA’s R&D on this research area, strongly linked with two infrastructural bases ofNaraha Center and Okuma Center. CLADS began the operation in April 2017 of the Main Building in Tomioka
Town, where the evacuation order was released in March 2017. CLADS is expected to be as both aninternational research hub and a symbol of the shared international interest in safely decommissioning the
Fukushima Daiichi NPS.

2-2. Naraha Remote Technology Development Center
The Naraha Remote Technology Development Center, which began operation in April 2016, wasestablished to conduct development/demonstration tests of remote controlled equipment and devices. The center
provides test equipment for remote control devices such as Robot Testing Water Tank, Mock-up Stairs, andMotion Capture for external users.

2-3. Okuma Analysis and Research Center
The purpose of the Okuma Analysis and Research Center is to analyze and characterize the radioactive
wastes and fuel debris for contribute to the establishment of treatment/disposal methods. The center consists of 3
buildings: an Administrative building, Laboratory-1 and Laboratory-2. The center is planning to start operationof the facility sequentially starting with the Administrative building in 2018.

52

N02

Overview of IRID R&D Projects
1

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

Abstract
Towards the decommissioning of the Fukushima Daiichi NPS, four entities cooperate closely to work together as one team.
IRID is a complex entity consisted of eighteen organizations that play the leading role in R&D for the decommissioning of
the Fukushima Daiichi NPS. IRID is conducting main three R&D projects; “Preparation of Fuel Debris Retrieval”,“Treatment and Disposal of Radioactive Waste”, and “Fuel Removal from Spent Fuel Pool”. In preparation for determinationof fuel debris retrieval policy in 2017, IRID continues to develop technologies for; 1. Intensive investigation on the existenceof fuel debris and damaged conditions inside the reactor, 2. Potential risk management and verification for nuclear safety, 3.
Reliable remote operation under high radiation environments.
1. Progress of Research and Development
As R&D projects for the preparation of fuel debris retrieval, IRID is undertaking research based on three elements.
Firstly, detection technology that enables to directly retrieve fuel debris in the PCV has been developed. In April 2015, a
robot successfully entered the PCV at Unit 1. In FY 2016, the preparation of fuel debris investigation outside of the Pedestal
is ongoing. At the same time, fuel debris investigation robots inside of the Pedestal, and remotely operated drilling device for
the PCV penetration have been developed to reduce the exposure of workers at Unit 2. Additionally, investigation of fueldebris inside the Pedestal, an underwater swimming robot has been developed for Unit 3. When these robots remotely accessfuel debris, it will enable to investigate a distribution and compositions of fuel debris. Severe Accident Analysis Code isupgraded to identify fuel debris inside the reactor, and investigations through the cosmic-ray Muon are ongoing. The
distribution of fuel debris in the reactor is investigated from outside of the reactor building through the Muon at Unit 1. It
resulted that the possibility is small that a large amount of fuel remains in the reactor core. The Muon investigation has
conducted at Unit 2 since March 2016. Essential technologies such as access to fuel debris in the RPV or the PCV are
ongoing. It is necessary to secure nuclear safety risks. Full-scale mock-up test facility of the lower part of the PCV at 1/8
sector was constructed for water leak blockage test at Naraha Remote Technology Development Center Test Building.

2. Future Development
With an aim to advance a very tough task of the decommissioning, IRID intends to proceed with R&D, collecting
knowledge and expertise from around the world. Specifically, not only hardware support from oversea organizations isrequired for removal and storage of damaged fuel, but also experiences of safety management and process. We continue tobuild and maintain close cooperation between us for R&D on safe and steady decommissioning.

53

N03

Sorting Technology for Efficient and Safe Handling and Storage
Steve Rima1, Jeffrey Lively1, Masaru Noda2, and Haru Hashizume2
Amec Foster Wheeler1, Obayashi Corporation2

Abstract
Obayashi and Amec Foster Wheeler (AFW) have been working on the application of sorting technology forradioactively contaminated soil after Fukushima incident. That includes the adaptation and upgrade ofwell-proven system in the US to the scopes and specifications for Fukushima offsite remediation. Its
adaptability is further expected to extend into onsite work demands in any challenges. This paper was
submitted at ANS Joint Topical Meeting on Remote Control and Robotics in Pittsburgh in August 2016.
1. Introduction
Japan’s national government commissioned the construction of waste consolidation facilities called Interim
Storage Facility (ISF) which functions include waste consolidation, assay, segregation, volume reduction andengineered temporary storage. ISF must be efficient, technically & scientifically competent, and able to handle
and segregate a variety of materials. Waste volume reduction is considered to be a key factor in ISF design.
2. Description of the Actual Work
Accelerating the processes of waste receiving, assay, segregation, and waste volume reduction are key factors
not only for timely completion of the whole ISF project but for protecting workers from radioactive exposure.
Obayashi has developed on ISF design that incorporates key measuring and sorting technologies to make the
ISF process more efficient, productive, competent, and reliable. One of the technologies that is being integrated
into the overall ISF process design is AFW’s ScanSortSM technology. To integrate the ScanSortSM technology
into the broader scoped and larger scale process train that Obayashi designed for ISF projects, it has become
necessary to reengineer many of the electro-mechanical components of the ScanSortSM system such that theywork collaboratively and seamlessly with Obayashi’s overarching ISF design.
3. Conceptual ISF Process Train Design
An appropriately scaled process train needed to be engineered with the overall electro-mechanical function sothat the ScanSortSM system does not function as “stand-alone”. Obayashi designed for the ISF process train
considering a number of factors in order to arriveat the design specification. Some of theconstraining parameters included the limitations
for material handling and feeding of 1 cubic
meter bags, the desire for built in redundancy andflexibility of scale, as well as the expectedvolume of material and the proposed time line for
ISF operations. Figure on the left shows the flowof fully integrated and automated soil processing
Figure: Conceptual Design of ISF Process Train
train. The process train incorporates a cubic
meter bag loading system, a container-bag shredder, a combustible-material separator, and the ScanSortSM
radiological measurement and sorting device supported with an engineering process to centralize, unify andintegrate mechanical function, power supply and distribution, electro-mechanical control, and instrumentationand control systems into a seamless and fully automated process train with a single point of command andcontrol.
4. Generation II ScanSortSM Design
AFW’s reengineering for an endpoint of command and control falls into two major categories: Decoupling of the sensor systems from the electrical power distribution system, the I&C system, and theelectro-mechanical control components, and Redesign of the ScanSortSM operating system software to support the decoupling of the I&C system and theredistribution of command and control functions to non-dedicated electro-mechanical modules
In addition, the redesigned software will provide for the remote display of real-time system information to acentralized command and control station. Obayashi and AFW collaborated on the specifications for theengineering effort required to produce new standardized interfaces for electrical power, I&C signals, andmodular mechanical attachment to the mechanical systems (conveyors and motors) sourced completely in Japan.
5. Benefits of the Generation II ScanSortSM Design
The benefits of this design are many, including: 1) State of the art radioactive measurement using large volume
scanning spectroscopy detectors and software algorithms, 2) Proven and reliable sorting technology, which will
confidently segregate materials based on their radiological content thereby reducing the volume of materials, 3)High material throughput rates, 4) Built in redundancy using a multi-line process train configuration, 5) A
higher degree of modularity than what exists in the current ScanSort SM system, allowing for increased flexibility,
redundancy, and reliability.

54

N04

Activities for fuel cooling at the Fukushima Daiichi Nuclear Power
StationMitsuyuki Yamazaki1

1

Tokyo Electric Power Company Holdings, Inc.

Abstract
TEPCO is trying to review fuel cooling operation at Fukushima Daiichi Nuclear Power Station, because
decay heat of fuel in reactor and spent fuel pool is declining gradually. The volume of water injected to cool thereactors was reduced from 4.5 to 3.0m3/h and moreover reduction is planning. At spent fuel pool Unit 1, it isconfirmed that decay heat is released by only natural heat removal.
1. Reduction in volume of water injected into the Unit 1-3 reactor pressure vessel (RPV)
Water injected to cool the Unit 1-3 reactor falls into
basement floors of building. The fallen injected water istreated by the contaminated water treatment facility with
underground water flowed in buildings and accumulated
water in buildings.

4.5 m3/h 4.0m3/h

3.0m3/h

3.5m3/hRPV bottom
temperature

Water injection
temperature

Air temperature

Reduction of the water injection volume is effective
to mitigate the contamination facility load. The surplus

Water injection volume (total)

capacity of the contaminated water treatment facility canbe utilized to accelerate treatment of accumulated water

Figure 1. RPV Temperatures in Unit 2

in the buildings. Furthermore, the distribution of melted fuel would be estimated by comparing behavior of
temperature measurements in RPV when the water injection volume is changed.
TEPCO reduced the water injection volume to Unit 1-3 reactors steadily by 0.5m3/h from 4.5 to 3.0m3/h
from December, 2016 to March, 2017.When reducing the water injection volume, monitoring parameters wereconfirmed, including the temperatures of the RPV bottom and inside the PCV, and PCV gas management facility
dust monitors. Upon carrying out this activity, real time disclosure of plant parameters was started from February,
2017. For moreover reduction, cut down from two inlets to one inlet to RPV is in preparation.
2. Check of spent fuel pool (SFP) temperature while bypassing the Unit 1 SFP cooling facility
Decay heat in SFP had declined sufficiently, and less than

Evaluation at 1.0m/s wind speed

200kW (calculated value) at Unit 1-3 for each one. TEPCO
made no heat removal condition by SFP cooling facility andinvestigated natural heat removal effect (such as evaporation,
convection from surface of the water) at Unit 1 during theat 2.0m/s
SFP temperature

at 1.5m/s

Air temperature(Average 12.3℃)

period April 5-26, 2017.
The results confirmed that the SFP water temperature was

SFP cooling facility bypass operation

stable at around 30℃ and the trend was almost equivalent to
the evaluation model including natural heat removal with a

Figure 2. SFP Temperature in Unit 1

wind speed of 1.5m/s. It is anticipated that the SFP water temperature will remain stable at 35-40℃, even if theambient temperature increases to 27℃(max. average monthly temperature over the past decade).
Continuously, natural heat removal effect will be investigated by similar bypass operation at Unit 2-3.

N05

The Overview of Project of Decommissioning andContaminated Water Management
Masaaki Matsumoto1, Naoki Kondo1, Takaki Ashida1, Masaki Kawai1
1

Mitsubishi Research Institute, Inc.

(Management Office for the Project of Decommissioning and Contaminated Water Management)
Abstract
For making decommissioning at Fukushima Daiichi NPS proceed in a safe and consistent manner, it isimportant for us to gather domestic and overseas wisdom for its research and development. Therefore, METI has
set up the foundation in the FY2013 for carrying out the “Project of Decommissioning and Contaminated Water
Management” which supports the R&D with the high level of technology.

1. Purpose
The purpose of this project is to advance the measures for decommissioning and contaminated water at
Fukushima Daiichi NPS consistently, based on “Mid-and-Long-Term Roadmap towards the Decommissioningof TEPCO's Fukushima Daiichi NPS.”

2. Overview of the ProjectNot to mention the technologies for retrieving fuel debris, we are also working on the research anddevelopment of the internal investigation technology for Primary Containment Vessel (“PCV”)/Reactor Pressure
Vessel, the repair and water stoppage technologies for PCV and the criticality control technology which can
improve the working environment.
In addition to the retrieval of fuel debris, the research and development of the long-term soundness
evaluation on fuel assemblies taken out of the spent fuel pool, the processing and disposal technologies for solid
waste, and the measures for the contaminated water have been implemented.
Owing to the past research and development,
various technologies are highly developed.
Therefore, the in-core state is gradually revealed
and the information for making fuel debris
retrieval policies is being gathered. Those
achievements in the research and development
are released on the website of the Project

For storage of SFP fuel:○Long-term soundness evaluation
on SFP fuelFor lowering of radiation dose inreactor building○Remote decontamination technology

For addressing the contaminated water:○Tritium separation technology○Seawater purification technologies○Capturing radioactive substancesfrom soil○Decontamination of contaminatedwater tank
○ Unmanned boring technology

For processing and disposing ofsolid waste:○Processing and disposal
technologies

Management Office. ( http://dccc-program.jp/ )For grasping state inside reactor:○Fuel debris detection technology by using
muon○Grasping state inside reactor by analysis○Investigation technology of inside of
PCV/RPV○Grasping properties of fuel debris○Sampling technologies for fuel debris
For ensuring the stability:○Corrosion inhibition technology for
PCV/RPV○Technology for evaluating seismic
performance of PCV/RPV○Criticality control technology

For stopping leakage from PCV:
○Repair technology for leakage sections
and water stoppage technology○ Full-scale test

For storage of fuel debris:○Storage/transportation/
repository technologies

For retrieving fuel debris:○Fundamental technology○Advancement of method and system

Figure 1. Main R&D Projects
3. Role of the Project Management Office
From the beginning of the subsidized project, Mitsubishi Research Institute, Inc. has been serving as the
Project Management Office by gathering experts of nuclear energy, project management, etc. The ProjectManagement Office is practicing a coherent project management which includes the solicitation of domestic andoverseas subsidized project operating entities, the evaluation of proposal/presentation and the adoption, the work
related to the subsidy grant, the management of project’s progress, the settlement of subsidy amounts, making
payment, etc.

56

N06

Power Assist Suit (PAS) to support heavy laborin radiological environment
Hirohisa Hirai, Yoshihiro Tamura
Mitsubishi Heavy Industries, Ltd.

Abstract
Mitsubishi Heavy Industries, LTD(MHI),in collaboration with The Japan Atomic
Power Company (JAPC), developed a wearable robot ,Power Assist Suit
prototype(PAS), to support heavy tasks in radiological environment(Figure 1). ThisPAS is expected the synergistic effect of the flexible correspondence by human and
Figure 1 Image of PAS usage

the high power by a robot. This presents such technical summary.

2. Development
In development of PAS, it was considered as the specification with which the following concepts are filled.
1.An outer size of PAS is restricted to about 700 mm or less supposing passing theairtight door which is the minimum narrow part inside of a nuclear plant.
2.In order to make the calibration for any workers unnecessary, the sensors have to beable to gauge the physical quantity but not a biological signal.
3.The rate of a radiation shield which is considered as equivalent 40 to 50% to theconventional one, and the mechanical structure which can be assisted by PAS are
adopted.

After the basic functional test, to verify the performance at actual environment, PAS

Figure 2 Tasks with PAS

was tested using the mockup equipment.

Passing through the door

Walking through the stairs

Accessing to the instrument

Figure 3 the test using mockup equipment
3. Conclusion
The PAS system has been developed and verified for the purpose of supporting heavy labor in radiologicalenvironment, and is improving to apply to actual fields continuously.
The PAS is expected to accelerate decommissioning tasks which are continuing in the Fukushima Daiichi
nuclear power plant. Furthermore the PAS shall become a utility tool which can be used for various heavy laborremained in normal maintenance of operating plants.

References[1] 1http://www.mhi.com/news/story/1512011943.htm (December 01.2015 No.1943)

57

N07

Veolia Nuclear Solutions Technologies for TEPCO 1F Decommissioning
Mamoru Numata, Syuichi Ueno, and Sadaaki Abeta
Kurion Japan K.K.

Abstract
Veolia Nuclear Solutions (VNS) has provided water treatment systems and remote access technologies forcleanup and decommissioning of TEPCO 1F.

VNS has also proposed vitrification system named GeoMelt®

for processing and solidifying radioactive wastes in the 1F site.

Technologies provided and proposed by VNS

for the 1F decommissioning are outlined briefly.

1. Introduction
A variety of technologies are required to respond to the decommissioning measures of TEPCO 1F site.
VNS has provided water treatment technologies for selectively removing radionuclide in water as well as remote
access technologies for inspection under high dose rate.

In addition, VNS is studying application of

vitrification system for reducing waste volume and stabilizing radionuclide in wastes.

2. Outline of Technologies
2-1. Water Treatment Technologies
VNS introduced selective adsorption systems called ISMS and KMPS at the 1F site to remove Cs and Sr
from contaminated water.

ISMS can process the contaminated water with a process rating of 1200 m3/day andwith decontamination factors of Cs for 105 and Sr for 3.5x103.

KMPS is the 5 skid-mounted assemblies used

for at-tank mobile isotope removal to treat radioactive water stored in hundreds of water storage tanks on-site at
1F.

The system is primarily used to remove Sr from the water with 300 m3/day.

2-2. West Stabilization Technologies
VNS proposes the GeoMelt® vitriﬁcation process for immobilization of wastes generated from the 1F sites.
The process has signiﬁcant advantages below:• High tolerance for debris such as concrete, metal, soil, etc.
• Commercially proven at large scale• High degree of technical maturity at U.S. DOE sites for treatment ofradioactive waste
2-3. Remote Access technologies
VNS has remote access technologies to support various initiatives
including reactor inspection, repair and fuel debris removal.

3. Conclusion
VNS offers unique value to customers, with a broad array ofcapabilities and oﬀerings in nuclear clean up.

58

Figure 1. Fukushima inspection
Manipulator(FIM)

N08

Contribution to Environmental Remediation in Fukushima
Takao Ikeda, Wataru Kobayashi
JGC Corporation

Abstract
Incineration ash and removed soil from decontamination work will be stored in interim storage followed by
final disposal outside Fukushima Prefecture in 30 years. To reduce the volume of the final disposal material, wedeveloped and have been demonstrating a new pyro-treatment technology for removal of persistently adhering
radioactive cesium from the object materials (incineration ash and removed soil). This demonstration project
was entrusted to us by the Ministry of the Environment, and we are implementing it at a demonstration facility
in Iitate village, Fukushima Prefecture, collecting object material from Iitate and its vicinity.

1. Dry Cesium Removal Technology
1-1. Treatment Process
The key technology of this treatment is to vaporize and remove radioactive cesium in the object materials
using some reaction accelerant in a rotary furnace at high temperatures around 1,350°C. Through this treatment,
the cesium content of the object materials is reduced to levels that allow unconditional recycling. The process
flow diagram is shown in Figure1.
1-2. Treatment Facility
After some operational tuning, actual operations began in April, 2016. A perspective of the facility isshown in Figure 2.

Figure 1 Process flow diagram of dry pyro-treatment process

Figure 2 Demonstration heat treatment facility

2. Conclusions


The incineration ash and the removed soil containing tens of thousands Bq/kg of radioactive cesium weretreated with the demonstration pyro-treatment furnace.

The radioactive cesium contents of the treated materials were reduced to the IAEA clearance levels (<100Bq/kg ), which levels allow unconditional recycling use including road bed material, backfilling, banking
or aggregate of concrete products.



The pyro-process properly scaled up is expected to reduce the amount for final disposal of incineration ash
and removed soil, thus contributing to the environmental remediation in Fukushima.

59

N09

Works for destroyed ChNPP Unit 4 conversion into an ecologicallysafe system as an experience in mitigation of Fukushima Daiichi
NPP accident aftermath
Viktor Krasnov1
1

Institute for Safety Problems of Nuclear Power Plants National Academyof Sciences of Ukraine (ISP NPP)

Abstract
In Chernobyl, New Safe Conferment (NSC) was designed and constructed by International Shelter Projectand it was successfully installed in the design position in November 2016. The shelter implementation plan
(SIP) was established in 1997 as internationally and the stabilization plans are implementing now.

1. Complex «New Safe Confinement - Shelter Object»
The NSC will bring to new quality complex – «New Safe
Confinement-Shelter Object», which prevent temperature and moisturechanges, and provide high radiation safety. The analysis of processes
currently occurring in “Shelter” object (SO) has demonstrated that there is ahigh probability for appearance of factors capable destabilizing the existing
conditions of nuclear and radiation hazard.

Figure 1. FCM location in NSC
and SO

2. Shelter Implementation Plan (SIP)The SIP shall be achieved through implementation of 3 stages. Stage1 (Done) is “Stabilization” which stabilize
the existing facility, enhance its operational reliability and longevity of structures. Stage 2 (Done) is“Preparation for Conversion” which create additional protective barriers, primarily the NSC, and ensure thepersonnel, public and environment safety. Stage3 (Progessing) is “Conversion” which remove the fuelcontaining materials and long-lived Radio Active Waste (RAW) from the SO, perform its conditioning with thefollowing storage and disposal in the RAW storage facilities in accordance with the effective standards;
complete the SO decommissioning. The SIP has a mandatory of “Early deconstruction”, which shall be
implemented by the end of 2023, taking into account design life time of the SO reinforcement structures.

Table 1. Shelter Implementation Plan (SIP)

60

F01

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

1

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

Abstract
Among the numerous challenges towards the decommissioning of Fukushima Dai-ichi reactors, removing fueldebris is for sure on of the tougher ones. A French consortium “ONET Technologies, CEA and IRSN” isdeveloping laser cutting technologies for fuel debris with the implementation experience in UP1 French
reprocessing plant. In addition to laser cutting development in air and underwater, key aspects of fuel debris
cutting are the representativity of fuel debris simulants to reflect the quantities of aerosols, gas and dusts released
during the debris laser cutting and their collection.

1. Cutting Technologies
The variety of fuel debris location, material and shape willneed various cutting techniques to be developed. Laser cutting
technologies are capable to overcome many configurations
with proven performance, and great benefit with regards toremote application. Research for application to Fukushimaenvironment is on-going with promising results.
Cutting trials have been conducted on melted fuel debris
simulant (inactive but heterogeneous with representative
simulant of all fuel debris elements, except Uranium andRuthenium).

2. Knowledge of dust and fumes generated during cutting
Laser cutting being a thermal cutting method, aerosols, gases and particles are emitted during the process andthe characterization of such secondary outlets is of paramount importance. The conducted trials have enabled thefirst estimation of released radioactive elements, aerosols size and gases, which are an important input data forcollection and treatment systems definition and estimation of release to environment.

3. Collection Technologies
In parallel to cutting technologies, collection systems are being
developed to capture the dust and fumes generated during the cutting.
Such technologies are to be deployed remotely, either embedded with
cutting system either by separate means.

Conclusions
Laser cutting technologies associated with collection technologies are key processes for fuel debris retrieval, but
could be also used for other systems cutting as part of Fukushima-Daiichi decommissioning as CRD or RPVinternal structures.

61

F02

Chemical-free Water treatment by Hydrodynamic Cavitation (HDC) andElectrocoagulation with sacrificial Ferrous anodes.
Gilbert Peronnet1, Tristan Cerosky2 and Christian HONNORAT2
1

Optimum Water System, a company of ISB Water., 2NUVIA France

Introduction
Ferrous Anodes Galvano-in Electro-Precipitation HDC Reactors can be applied as pretreatment before the
SARRY’s and ALPS systems. In this solution, ferrous iron hydroxides are electrically produced directly in thebulk fluid, and cavitation enhances the adsorption capacity of the produced iron hydroxide for a better removalof radioactive pollutants before polishing by existing columns, resulting in reduction of waste volume.

Principle

After injection of air microbubbles, the water to betreated is passed through a reactor in which an electric
current flows through sacrificial electrodes. This
results in electrochemical reactions taking place
between them. The sacrificial electrodes dissolve,
releasing their metal ions, and metal hydroxide flocs
are produced in the process. These generated metal

Figure 1; P&ID of typical systemhydroxide flocs have a high adsorption capability andcan thus bind the finely dispersed radioactive particles. The ferrous ion is specifically added to the water to betreated, and in the formation of hydroxide flocs there are also co-precipitation and occlusion precipitation
reactions, in which undesirable substances are precipitated. The co-precipitated substances can then be separated
mechanically by sedimentation or by filtration.

The water passing through the reactor is also subject to hydrodynamic cavitation. When a liquid flow passes
through the constriction, the velocity of the liquid increases and the local pressure is consequently reduced. In
HDC reactors, throttling is sufficient to cause the pressure around the point of lowest local pressure to fall below
the threshold pressure for cavitation which is usually around the vapor pressure of the medium at the operating
temperature, cavities are generated. Subsequently, as the liquid jet expands, the pressure recovers and this resultsin the collapse of the cavities. This phenomenon of generation, growth
and rapid collapse of the cavities has been described as hydrodynamic
cavitation. When pressure increases, the cavities cannot sustain thesurrounding pressure, and consequently collapse, creating numerous
localized points of extreme high pressure and temperature. The intense
mixing forces around the collapsing bubble will lead to more complete
adsorption of contaminants inside the floc material.
Figure 2: Floc with Contaminant Embedment

62

F03

Hydrodynamic Cavitation (HDC) application to existing Reverse
Osmosis Systems.
Gilbert Peronnet1, Tristan Cerosky2 and Christian HONNORAT2
1

Optimum Water System, a company of ISB Water., 2NUVIA France

Introduction
Zinc Anodes Galvano-Chemical-Precipitation HDC Reactor units applied for scale mitigation in front of
membrane filtration units. The "brutality of collapse" of the cavitation bubbles maximizes the supersaturationsof ions in the vicinity of the bubbles, and create bulk precipitation of carbonates, sulfates, phosphates or othersalts present in the water to be treated. RO Membranes are protected with better output.

Principle of treatment
The water passing through the reactor issubjecttohydrodynamic

cavitation.

Hydrodynamic cavitation can simply begenerated by the passage of the liquidthrough a constriction such as orifice plates,
nozzles, or throttling valves. When a liquidflow passes through the constriction, theFigure 1 P&ID of typical systemvelocity of the liquid increases and the local
pressure is consequently reduced. If the throttling is sufficient to cause the pressure around the point of lowest
local pressure to fall below the threshold pressure for cavitation which is usually around the vapor pressure of
the medium at the operating temperature, cavities are generated. Subsequently, as the liquid jet expands, thepressure recovers and this results in the collapse of the cavities. This phenomenon of generation, growth andrapid collapse of the cavities has been described as cavitation and in this case, ‘’ hydrodynamic ‘’ cavitation.

When pressure decreases to low values, cavities are formed in the liquid. When pressure increases, the cavities
cannotofsustainthe surrounding

pressure, andconsequently collapse, creating numerous localized points

extreme high pressure and temperature. The bubble will eventually collapse to a minute fraction of its

original size, at which point the gas within dissipates into the surrounding liquid via a rather violent mechanism,
which releases a significant amount of energy. At the point of total collapse, the temperature of the vapor within
the bubble may be several thousand Kelvin, and the pressure several hundred atmospheres. This is resulting in
numerous chemical reactions settling in, and release CO2 and other dissolved gases from the solution. When
properly designed and run, this technology can lead to very high sur-saturations levels in the surrounding of thecollapsing bubbles and to the precipitation of calcium carbonate and other salts.

This early precipitation nuclei are also the seed for further precipitation to takeplace on the “concentrate side” of the RO membrane and thus protecting it.

Figure 2: Non-Clogging Solid crystals
63

F04

Froth flotation process for remediation of contaminated soil
S. Faure 1, A. Ben Said 1, A. Grandjean 1, F. Charton 1, C. Georges 1,
T. Prévost 2, D. Grange 3, J.P. Spanu 3, F. Maneglia 4
1
CEA, French Alternative Energies and Atomic Energy Commission, Nuclear Energy
Division, 2 AREVA NC, Business Unit Recycling, 3 VEOLIA, 4 ANADEC

Abstract
Soils with large clay fraction are difficult to decontaminate because of clay affinity with Cesium 137 in a nearly irreversible
way, resistant to washing.
Froth flotation process demonstrated its ability to extract contaminated clay, without destroying its organic part, allowing to
reduce the volume of the contaminated waste soil to store and to reuse part of the decontaminated soil.
1. Introduction
Despite the numerous existing soil decontamination technologies such as soil washing, magnetic separation, electro kinetic
processing, and phytoremediation methods [1, 2, and 3], there is a need for environmentally friendly process limiting the
production of secondary waste and able to treat high volumes of contaminated soil.
Based on the fact that the finest particles <20 µm contain 60 to 90% of the cesium radioactivity, solutions were searched toseparate them with the objectives:- To minimize the amount of radioactive wastes to be disposed in the intermediate storage centers (>8000 Bq/kg)- To provide an eco-friendly decontamination process accepted by the population- To provide a complete and transportable compact solution compatible with the disposal of radioactive waste- To use emerging proven technologies in partnership with Japanese entities
Micro-filtration being un-adapted because of different limits (problems of cake resistance, clogging of the membrane, limited
flowrate, variable cut-off, and cost), CEA developed a solid-liquid process based on froth flotation , in partnership with AREVA
and VEOLIA in the frame of a project supported by the French Research Agency (ANR), called DEMETERRES [4].
2. Process Description
Contaminated soil is continuously introduced in a column where it
generates a ground suspension in water and forms a flotation froth of the clay
particles under the effect of a dispersed air bubbling and addition of asurfactant.
The foam volume evacuated by vacuum at the top of the column is transferred
to two storage containers. The slurry is discharged continuously from thebottom of the column into a decanter. Sedimented particles ofdecontaminated soil are isolated in plastic containers. 50 to 80% of water isrecycled in the process.

Contaminated soil(>8,000 Bq/kg)

Air
Decontaminated soil(<8,000 Bq/kg)

Figure 1 Diagram and photo of the heart of process(particulate flottation)

3. Results within project Demeterre
Several soils, presenting various textural characteristics, were floated and compared. The effects of processing parameters,
including surfactant/soil ratio, froth residence time, suspension concentration and air flow rate, upon fine particles extraction
performance to determine the most significant variable that affect flotation and obtain the operating conditions which led tothe best performance. Finally, the selectivity of the process on the fine particles was investigated and scale-up studies were
performed at pilot scale. Flotation results showed that fine particles extraction from soil is feasible thanks to attachment to air
bubbles and exhibits excellent floatability and selectivity. The results obtained at lab and pilot scales evidenced good agreement
which provides valuable evidence to support the technical viability of the soil flotation process for
future exploitation at industrial scale.
The flotation column is the heart of the process. Additional independent skid which includes thesoil preparation functions, recovery of the treated soil and recycling and recirculation of water isconnected to the flotation column. Integrated pilot allows an overall process scheme "Preparationprocessing-recycling" suspensions of soil.
This pilot unit was successfully tested in France and in Japan over hundred hours on 5 different
soils.
4. Conclusion
Froth flotation is a proven technology, already widely used in industrial plants e.g. for the extraction /purification of minerals. It
tends to be an innovative and cost-effective alternative to decontaminate radio-contaminated soils by selectively extracting thefine clay particles containing the majority of contamination and thus minimize volumes of waste to be stored (15-30% of initial
weight). In addition a new demonstration campaign should occur soon, with main objectives to asses and consolidate the foam
flotation process performances on radioactive cesium 137 contaminated clay soils from Fukushima area
References[1] B.C. Bostick, M. A. Vairavamurthy, K.G. Karthikeyan and J. Chorover, Environmental Science Technology, 2002, 36, 2670.
[2] Y. Yuki, K. Sekiya, N. Nomura, F. Mishima, Y. Akiyama and S. Nishijima, IEEE Transactions on Applied Superconductivity, 2014, 25, 1.
[3] S. Sharma, B. Singh and V.K. Manchanda, Environmental Science and Pollution Research, 2015, 22, 946.
[4] S. Faure and M. Messalier, Method for the radioactive decontamination of soil by dispersed air flotation foam and said foam, Patent
WO2013 167728 A1

64

F05

IN CAN MELTER for the conditioning of high and medium-level andlong-lived waste from decommissioning operations
Christophe GIROLD 1, Christine GEORGES 1, Stéphane CATHERIN 2,
Thierry PREVOST 3, Guillaume LECOMTE 4
1
CEA, French organization for nuclear and alternative energies, Nuclear Energy
Division, 2 National Radioactive Waste Management Agency (ANDRA),
3
AREVA NC, Business Unit Recycling, 4 ECM Technologies

Abstract”IN CAN MELTER” process is the solution developed by CEA for its own waste coming from D&D operations, including legacy
waste management, when “glass-like” encapsulation is required for further disposal but when cost reduction is an aim.
A specific melting tool was designed to be implemented In Situ, close to D&D workshops, and to match the requirements and constraints
of D&D projects.
1. Introduction
Some waste streams from decommissioning operations are identified as an issue in terms of handling, transportation, adapted
conditioning matrix or evacuation outlet within the existing regulations. The numerous constraints associated with the management ofsuch waste led CEA to investigate the implementation of an in situ conditioning treatment able to produce a waste package complying
with existing evacuation routes and/or on-site storage facilities.
2. Specification
In this context, glasslike matrix brings a satisfactory answer as the glass material and the vitrification process demonstrated long ago tobe well adapted to high level waste and to be flexible enough to accommodate to variable waste streams.
Specification for a vitrification process adapted to D&D was then:- to be adapted to high or medium level waste,- to be flexible enough to accommodate uncertainties on waste composition and however produce a confinement with quality
description, with source term calculation methodology for long term storage,- to be adequately designed and compact enough to be installed and operated inside a decommissioned cell,- to be designed for a short duration of use as a decommissioning tool aimed to be dismantled just after the treatment operation,- to produce a small amount of secondary waste,- to generate minimum investment and operation costs.
2. Developments
To rapidly deliver a high TRL and provide all the basic data necessary to start an industrial
application, “IN CAN MELTER” process took as a basis a previous development alreadyfully qualified for specific CEA needs in facilities with alpha contaminants [1, 2].
In this process, the container is used as a crucible ("in can"), renewed with each batch,which greatly simplifies the equipment and makes it very robust: there is indeed no needto drain the glass and corrosion problems are limited by the batch limited time.
Secondary waste are limited to the heater as no components with molten glass remains
after use. Thanks to its small size, the crucible can be easily heated by a simple and robust
resistance heated furnace, inexpensive, easy to maintain, easy to implement and offering
a great homogeneity to the melted mixture. This heating mean allows to have a metallic
fraction in the canister and allow an excellent control of the treatment temperature, with
low temperature steps if needed,
The process shown considerable flexibility as it will for example for the CEA application
be operated only with two shift teams and no operator during night shifts,
The compactness and basic technology of the IN CAN MELTER makes it arelatively low investment.
3. Conclusion“IN CAN MELTER” with low temperature glass (900 °C) is presently under qualification for the vitrification of high active deposits
and sludge from French shut down facilities, containing fission products and alpha components.
Another field where low temperature solidification without additives would be of great interest is the encapsulation of mixed effluent
waste such as zeolite, co precipitation sludge or powder of fuel debris, etc.
References[1] « Qualification of In-Can Melting Process Applied to Vitrification of High Activity Waste Solutions (HAWS) at the CEA’s Valduc Center:
Description of Process and Equipment, Methodology and Initial Results”, P. Gruber, S. Lemonnier, J. Lacombe, A. Ledoux, Y. Papin, I. Hugon,
J.L. Dussossoy, B. Batifol, L. Pescayre, WM 2010 Conference[2] Development of In-Can Melting Process Applied to Vitrification of High Activity Waste Solutions (HAWS): Glass characterizations and process
tests results”, P. Gruber, S. Lemonnier, J. Lacombe, Y. Papin, I. Hugon, B. Batifol, L. Pescayre, WM 2012 Conference

65

F06

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

Abstract
It has long been common practice to solidify and stabilize low- and intermediate-level radioactive waste with calcium silicate based
cements (Ordinary Portland Cement). However, some wastes produced by decommissioning operations may chemically react withcement phases or mixing water, affecting hydration and eventually resulting in the deterioration of the waste form in storage or disposal,
for instance by swelling and cracking. In order to avoid pretreatment or increased volumes of waste, three kinds of alternative inorganic
binders offering better chemical compatibility with these problematic wastes are of great interest: (i) calcium aluminate and sulfoaluminate cements, (ii) magnesium and calcium phosphate cements, and (iii) alkali-activated binders and more specificallygeopolymers.
1. Calcium Sulphoaluminate Cements (CSAC)
CSAC clinker are produced by firing mixtures of limestone, gypsum and bauxite.
Foreseen applications of CSAC for waste conditioning ranges from Cementation of borate or heavy metals such as lead, cadmium, zinc,
and trivalent chromium or to immobilization of Aluminum.
Due to the different cement chemistries, the rate of hydration of CSAC may be less affected by strong retarders of OPC such as heavy
metals or borate ions and its mineralogical structure is favorable for waste immobilization, offering possibilities of ion substitutions.
2. Magnesium Phosphate Cements (MKP)
Phosphate cements are the main representatives of acid-base cements. Waste immobilization results from two processes: precipitationof many contaminants, particularly actinides, as phosphates with very low solubility, and physical encapsulation in a dense phosphatematrix. Good results have been reported for several types of wastes, e.g. low-level debris wastes contaminated by 137Cs, highly saline
effluents or metallic aluminum [1].
3. Geopolymer: an alkali-activated binder
Alkali-activated binders are made by mixing solid
aluminosilicates, such as fly ash, metakaolin, various
clays usually activated by heat or blast furnace slag, withan activating solution comprising high concentrations ofalkali hydroxide (NaOH, KOH) and / or polysilicate(Na2O.nSiO2, K2O.nSiO2). The reaction product,
formed according to a dissolution/polycondensation
process, exhibits an X-ray amorphous network structure
and is usually called geopolymers [2].
By selecting appropriate starting materials and byvarying the conditions of processing and curing, it ispossible to vary the properties of the produced alkaliactivated binders over a wide range, and to tailor them tospecific requirements like high compressive strength,
low permeability, low shrinkage.
Geopolymers offer great potential to stabilize/solidify
hazardous wastes like heavy metals, alkali-earth, alkaliions and magnesium alloy encapsulation (Figure 1).
Geopolymers present a new way for direct organic liquid
waste encapsulation by stabilizing emulsion of oil in an
alkali silicate solution and being directly solidified byaddition of metakaolin (Figure 2) [3].

Figure 1: Encapsulation of Mg
cladding in geopolymer

Figure 2 : Oil/ geopolymer composite

3. Conclusion
New binders offer great opportunities for the management of problematic waste besides, understanding the chemistry of cement – waste
interactions, and their consequences on the physical properties of the solidified waste forms, including their long-term evolution, will
have to be further considered for the acceptance of these alternative binders in nuclear waste conditioning. Particular attention will also
have to be paid to their possible interactions with the near-field environment.
References[1] Cau dit Coumes C., Lambertin D., Antonucci P., Charlot M., Liant et son utilisation pour le conditionnement des déchets contenant de l’aluminium
métallique, Patent FR 13/52233, 13/03/2013.
[2] Davidovits J (1991) Geopolymers: inorganic polymeric new materials. J. Therm. Anal. 37:1633-1656[3]. Lambertin D., Rooses A., Frizon F., Procédé pour préparer un matériau composite à partir d'un liquide organique et matériau ainsi obtenu, FR
12/58880

66

GB01

UK and Japan - Natural Partners in Nuclear Decommissioning
1. UK-Japan Collaboration

Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
The home of the World’s first full scale Nuclear Power station at Calder Hall, the UK nuclear industry iswell over 60 years old, and has developed a mature and capable decommissioning sector. This industry has ahistory of successful collaboration and this series of posters demonstrates the benefits of this collaboration, andhow it has been used to support the work at Fukushima Dai-ichi.
Figure 1. UK-Japan Collaboration

NDF/NDA Agreement

TEPCO/Sellafield Agreement

Japan/UK Government MoC

1. UK-Japan Collaboration
The UK and Japan have recognised the benefits of sharing experience in decommissioning through a series ofcollaboration agreements (Figure 1).
Figure 2. NDA Decommissioning Sites
2-1. The NDA
Decommissioning in the UK is the responsibility of the
Nuclear Decommissioning Authority (NDA). This allows
sharing of information and learning across all sites, leading tocost savings and more efficient decommissioning.
2-2. Technical Challenges
The technical challenges faced at UK decommissioning sites
such as Sellafield, Magnox and Dounreay, have similarities tothose at Fukushima Dai-ichi. This series of 8 posters show
how this experience is proving valuable in Japan.
3. Conclusion
Collaboration, whether between companies or between nations is the key to successful decommissioning.

67

GB02

UK and Japan - Natural Partners in Nuclear Decommissioning

2. Sellafield and Fukushima Daiichi - the challenges of
stakeholder engagement
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
Sellafield has been a key part of the of the UK’s nuclear industry
since the 1940’s. It deals with many challenging nuclear wastes,and over the years has learned important lessons about waste
management and decommissioning complex facilities. The site
has learned the benefits of working with local, national andinternational stakeholders, and is now sharing this experience
with TEPCO.

1. Sellafield-TEPCO Agreement
Since 2014 Sellafield Ltd and TEPCO have been supporting each other’s decommissioning programmes

2. Fukushima-West Cumbria Study on Stakeholder Engagement
TEPCO and Sellafield have a joint project looking at how eachcompany engages with the local community surrounding their sites.
This has included TEPCO staff attending the local stakeholder
meetings in West Cumbria, which is where Sellafield is situated.

3. Conclusion
Sellafield and Fukushima Dai-ichi are sites with similar issues which cannot be tackled without the support ofthe local community. Sharing how best to engagement with local stakeholders is an important part of ensuring
successful decommissioning at both sites.

68

GB03

UK and Japan - Natural Partners in Nuclear Decommissioning

3. Waste Retrieval – Magnox Reactor Sites and Dounreay
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
The final UK Magnox reactor completed power generation in December 2015, the last of 11 Magnox reactor
sites which started operation in 1956.

In addition, the breeder reactors at the Dounreay site in Scotland are also

shut down and the site is being decommissioned.

An on-going decommissioning programme in the UK acrossmultiple Magnox sites and at the Dounreay site is addressing many varied decommissioning challenges.
Significant progress is being made across all of these decommissioning sites with lessons being learned that are
applicable to decommissioning and waste management projects at Fukushima Dai-ichi.
1.

Retrieval and Management of Radioactive Wastes

Retrieving highly radioactive waste that is difficult to access, and storing it safely, is a challenge currently facing
Fukushima Dai-ichi, as well as at the Dounreay and the UK Magnox sites.

Technologies and methods havebeen developed in the UK decommissioning programme for retrieval of radioactive wastes and the subsequent
processing, packaging and storage of these wastes.

Figure 1. Retrieval of waste at Figure 2. Interim storage facility at
Hunterston A
2.

Berkeley

Figure 3. Retrieval of wastes at
Dounreay shaft

Conclusion

Innovative work is taking place at Magnox stations where operational waste that was in concrete bunkers above
or below ground, is being retrieved and placed into interim storage. Work is also taking place at Dounreay whereoperational wastes are to be recovered from a 64m deep shaft. These technologies and methods are relevant tothe challenges at Fukushima and are being applied to the development of the fuel debris retrieval methodologies.

69

GB04

UK and Japan - Natural Partners in Nuclear Decommissioning

4. Waste Retrieval – Sellafield
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
Material stored in ponds at Sellafield for many decades have given Sellafield a waste retrieval and disposal
challenge which is in many ways similar to the challenges facing Fukushima Daiichi.
1. Waste Retrieval
Water, radioactive sludge and swarf must all bemonitored, categorised, retrieved and packaged for future
disposal. Remote handling both in and out of water are
key to the effective and efficient delivery of this work.
2. Regulator Involvement
Close engagement with the UK regulator is key to a
successful project. The UK has developed methods todeal with challenging retrievals without the need to
Figure 1. Removal of waste from a Sellafield Pond

vastly expensive and time-consuming new facilities.
3. Innovative Technology
Innovativetechnologies

have

been

developed and are UK organisations areworking with Japanese partners to apply
these to the challenges at Fukushima
Dai-ichi and elsewhere.
Figure 2. Fuel Debris at Sellafield

Figure 3. Metal Fuel removal

4. Final Disposal

from a Sellafield pond

All retrievals consider the final disposal option for the material.

Wherever possible the waste is packaged withthis in mind which saves cost further down the line.

5. Conclusion
Working together Sellafield and Fukushima Dai-ichi experience can accelerate the decommissioning process
and reduce the hazards on both sites.

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GB05

UK and Japan - Natural Partners in Nuclear Decommissioning

5. Programme Management
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
The UK’s Nuclear Decommissioning Authority (NDA) utilises a programmatic approach to effectively delivery
best value for the UK tax payer. The structured approach is deployed throughout the NDA’s 17 sites enabling
consistent decision making, delivery andreporting.

Projects,

programmes

and

portfolios become common terminology.. The
UK

has

many

organisations

skilledinmanaging complex decommissioning projects.
They are working with Japanese organisations
to ensure the lessons learned from the UK
programme can benefit Japan.

Figure 1

shows the NDA’s strategic themes.
Figure 1. NDA’s Strategic Themes

1. Principles
Effective and efficient delivery across a range of missions within a complex multi facility and multi-site estate
requires a structured framework. By utilising established portfolio, programme and project structures the UK
nuclear decommissioning sector lead by the NDA has been able to give a higher degree of clarity to UK
Government. The framework enables relative priorities to be understood, assessed, prioritised and delivered.
UK industry has been able to work with these arrangements as a site licence company, a Management andOperations (M&O) organisation or as a supplier.

2. Conclusion
NDA has established a programme management framework that has delivered benefit to the UK and they havebeen very pleased to be able to share this with NDF.

UK companies have been working with TEPCO to share

the approaches used, their strengths and the successes this brings as well as the areas where the UK has gained
significant learning.

71

GB06

UK and Japan - Natural Partners in Nuclear Decommissioning

6. Decommissioning Technology
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
Technology and innovation are key to delivering safe andaffordable energy at each stage of the nuclear lifecycle.

Particularly

in decommissioning where legacy challenges require ‘first of a kind’solutions. However, entirely new technology is often not required, andUK companies have often been able to adapt technologies from otherindustries to provide ‘fit-for purpose’ solutions.
1.

Fig 1

Examples of Technology

Figure 1.

‘RISER’, a3D Laser scanning and radiation mapping

mounted on an Unmanned Air Vehicle (UAV) developed in the UK by
Createc and BlueBear now deployed at Fukushima Dai-ichi.

Figure 2. Advanced robotic equipment developed by Oxford
Technologies for retrieving complex waste from Dounreay facilities.

Fig 2

Figure 3. Underwater remotely operated vehicles with specialist toolingdeveloped by James Fisher Nuclear are tested before deployment at Sellafield.

Figure 4. The GROUNDHOG Evolution
2 and Synergy developed in the UK by
Fig 3

Nuvia for real time detection of particles
to carry out beach monitoring surveys
both at Sellafield and Dounreay.

2.

Fig 4

Conclusion

In addition to new innovations, Nuclear Decommissioning technologies can often be adapted from otherindustries. UK firms have successfully applied this methodology to solve decommissioning challenges at manysites in the UK and also at Fukushima Dai-ichi.

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GB07

UK and Japan - Natural Partners in Nuclear Decommissioning

7. Needs Based Research and Development
Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
Nuclear decommissioning at sites such as Fukushima Dai-ichi and Sellafield can throw up a series of uniquechallenges. In order to solve these challenges it is important that research matches the needs of thedecommissioning industry.

Figure 1: Research Hierarchy
1.

Needs based research

The philosophy of needs based research has spawned a number ofinnovations to address decommissioning challenges in the UK. Thismethodology is being shared with Japan through collaborations withorganizations such as CLADS.

2.

Examples of Research Successes



Detailed studies of hydrogen evolution has enabled a significant reduction in cost and schedule at
Sellafield.

Development of waste forms has allowed difficult wastes at Sellafield to be immobilised for long term
storage and disposal.

Studies to understand long term waste behaviour has allowed increased waste loadings in containers saving
time and cost.
Figure 2: Nuvia development rig

3.

Figure 3: Windscale fuel characterisation

Conclusion

The UK has many years of applying needs based research to accelerate or de-risk decommissioning plans. By
UK and Japanese organisations working together, there is the opportunity to find solutions to the challenges
faced by both countries through coordinated needs based R&D.

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GB08

UK and Japan - Natural Partners in Nuclear Decommissioning
8. Waste Management

Keith Franklin1, Roger Cowton2, Mark Sharpe3, Kat Lennox4, Jorge Weir5, John Falch6, Matt Mellor7,
Chris Carter8, Linda Ashton9, Steve Caley10, Sarah Groves11
1

4

British Embassy Tokyo, 2Sellafield Ltd., 3Oxford Technologies Ltd, National Nuclear Laboratory,

5

NUVIA, 6James Fisher Nuclear Ltd, 7CREATEC, 8 Cavendish Nuclear Ltd, 9 WS Atkins, 10Amec
Foster Wheeler, 11International Nuclear Services

Abstract
Successful Waste Management is the key to any decommissioning project, and is a large proportion of the cost.
The UK has a national integrated waste management strategy, which is designed to minimise both the amount ofwaste and the cost of treating it. It provides broad categories for waste, and defined disposal routes, reducing thecosts and uncertainties in decommissioning projects.
Figure 1. The Waste Management Hierarchy
1. Waste Hierarchy
By using the waste hierarchy, the amount of waste which
has to be treated and stored is significantly reduced. Thisis achieved through careful management of the activities
which produce waste, and treatment processes whichallow waste to be disposed of safely or reused.
Figure 2. Dounreay Store

2-1. Waste Storage
A range of waste storage facilities have been built in the UK,
after full consultation with local communities.
2-2. Waste Treatment
Treatment processes have been designed to minimise
radioactive wastes
3. Conclusion
Using an integrated waste management strategy, combined with proven treatment processes and appropriate
storage facilities the UK has been able to provide solutions to many of the waste issues, and this learning isbeing shared with Japan in order to tackle the task of managing the wastes produced as part of the Fukushima
Dai-ichi clean-up process.

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GB09

Decommissioning and Waste Management
Jorge Luis Weir Vera1 (Jorge.Weir@nuvia.co.uk);
Tom Withey1 (tom.withey@nuvia.co.uk);
1
Nuvia Limited (www.nuvia.co.uk)

Abstract
Nuvia has over 40 years’ experience in the decommissioning and waste management of Nuclear power
plants in UK and other parts of the world from planning to demolition and site remediation. Nuvia’s leadingedge consultancy services, engineering services and technologies continue to be used throughout thedecommissioning lifecycle and waste management of NPSs in UK and globally, providing a unique opportunity
to deal with the challenges faced with decommissioning and waste management at the Fukushima Daiichi
Nuclear Power Stations (NPS).
1. Introduction
This paper summarises features requiring consideration for design of decommissioning, some of the key
challenges and how Nuvia’s experience can be utilised.
2. Lifecycle of Decommissioning and Waste Management
In line with IAEA guidance [Ref 1], Nuvia’s waste management &decommissioning projects proactively aim to follow a set of design
features relevant to international best practice:Fig -1
Characterisation and minimisation of activated products;
Contamination control and decontamination facilities;
Segregation of contaminated items;
Ability to segment large items in-situ;
Minimisation of operational and decommissioning waste;
Adequate rad waste packaging and storage facilities;
Management of records for decommissioning.
Fig -2
3. Dealing with key Fukushima challenges
In order to deal with Fukushima effectively, prior knowledge andexperience of decommissioning and waste management is essential.
Below is a numbered list of some of Nuvia’s projects (in line with
Figures 1 to 6) each one followed by a challenge faced at Fukushima[Ref 2]. The aim is to show how Nuvia’s experience can support theFig -3management to deal with some of the Fukushima challenges.
Fig-1. B204 Contaminated Stack Decommissioning and design of
Stack Climbing Platform (SCP).
o Fukushima Daiichi Unit 1&2 Stack Dismantling.
Fig-2. Decommissioning of the Steam Generating Heavy Water
Reactor (SGHWR).
Fig -4o Fukushima Daiichi decommissioning and waste management.
Fig-3. External Active Sludge Tank (EAST) Treatment Plant (WETP).
o Fukushima Daiichi radioactive waste characterisation,
treatment, reduction, storage and disposal measures.
Fig-4. Modular Active Effluent Treatment Plant (MAETP)
o Fukushima Daiichi contaminated ground water processing.
Fig-5. Nuclear Safety & Design Support Chernobyl’s New Safe
Fig -5
Confinement (NSC) Dome, Ukraine.
o Fukushima Daiichi radioactive waste storage facilities.
Fig-6. Radiation monitoring using Groundhog & Hi-Ram.
o Fukushima Daiichi seawater and land monitoring.
Fig -6
4. Conclusion
This paper presented an overview of Nuvia’s unique experience throughout the lifecycle of decommissioning
and waste management. A number of decommissioning project examples were presented highlighting how thisexperience would be useful in dealing with key Fukushima challenges.
References[Ref 1] IAEA, “Design Lessons Drawn from Decommissioning of Nuclear Facilities, “IAEA TECDOC No. 1657, 2011.
[Ref 2] TEPCO CUUSOO Website, “An open innovation platform for energy related technologies and ideas,” Feb. 1st, 2016.

75

GB10

Decontamination and Decommissioning with Radiation
Imaging Solutions
Mike Anderson, Kazushi Watanabe and Victoria Anderson
Innovative Physics Limited

Abstract
Effective decontamination and decommissioning of radio activity for post event clean-up
operations and dismantling of Nuclear Power Plants. Innovative Physics (IPL) have developed arange of radiation imaging solutions for Fukushima Prefecture following the Dai-ichi Power Plant
incident in 2011. Rapid visual representation of radioactive areas makes these radiation cameras
a key tool for decontamination and decommissioning teams worldwide.

76

GB11

International case studies applicable to thedecommissioning and waste management challenges at
Fukushima Daiichi
Jenna McIntyre1, Nick Bold1
1

DBD International

Abstract
Whilst the Fukushima Daiichi (1F) site is unique, there are a number of relevant worldwide decommissioning
projects and related learning from experience that can be utilised to benefit the site. DBD are well positioned toapply this knowledge to the issues faced on 1F in order to accelerate a reduction in hazard levels on site.

1. Introduction
This poster summarises some international case studies that are relevant to the range of technical and strategic
issues that face the Fukushima Daiichi site. The poster focusses on four prominent case studies which highlightthe range of projects DBD has previously been involved in and how this knowledge and learning fromexperience could be applied to further the work and assist strategic decommissioning on the 1F site.

2. Poster and Case Study Summary
2-1. Support to the 1F site (see the top photo in Figure 1)
This section highlights DBD’s work on the Fukushima site where DBD’s D2O
optioneering process was used to assist with current issues facing the 1F site.
2-2. WRPS (see the middle photo in Figure 1)
This example summarises work DBD carried out to assist a US client inestablishing a waste form development program to address performance needs andunderstand waste form characteristics of Solid Secondary Wastes.
2-3. MSSS (see the bottom photo in Figure 1)
DBD has been part of an integrated client team, providing a range of strategic andtechnical support, this section highlights key support DBD has provided to MSSS.
2-4. Strategic Modelling

Figure 1 – Fukushimasite, Hanford site, MSSS
- UK (top, middle &bottom, respectively) [1]

DBD produced a new effluent treatment strategy model for a highly complex chemical facility, providing theclient with a tool that predicted multiple scenarios for reducing risk on site.
2-5. DBD’s Core capabilities
The centrepiece of the poster is DBD’s
five core capabilities (Figure 2). Eachcase study refers to these key capabilities.

Figure 2 - DBD's Core Capabilities

3. Conclusion
This poster represents a portion of the wide range of work that DBD carries out internationally and theapplicability of this work to problems currently faced on the 1F site.

References[1] Publicly available images from online search engine – Google, sourced June 2017

77

R01

Fuel debris management system
Main technologies and approaches for fuel debris treatment.
Oleg Bagryanov, Senior Project manager, TENEX

Abstract
To detect the localization of fuel debris inside the PCV, to implement the sampling of fuel debris inside the
PCV and also to develop recommendations for removing of fuel debris inside the PCV and RPV.
1. Introduction
Rosatom develops and applies modern control system neutron-physical characteristics of reactor units oftype VVER, RBMK and BN, as well as points of SNF storage that allows to create neutron detectors of small
dimensions for the detection of fuel debris within any geometry. Rosatom is also developing a remote controlled
robotic systems for repairs and sampling of metalwork inside the RPV that allows to create sampling system for
implementation of sampling inside the PCV of Fukushima Daiichi NPP.
2. Main steps in fuel debris treatment
2-1. Detection of fuel debris
Detection of fuel debris plays one of the most important role in developing of debris removal technology.
Rosatom’s technologies (by using of small-size neutron detectors with automatic tuning) allow to accurately
determine the localization of fuel debris.
2-2. Developing of sampling technology (inside the PCV). Implementation of sampling (inside the PCV).
Studying of the characteristics of the samples
Debris sampling will allow to obtain data on the properties of the debris and also allow to formulate theinitial data for development of equipment for the debris removal.
Based on the experience in the development of robotic tools for conducting manufacturing operations at
the RBMK type reactors (without fuel unloading) the enterprises of Rosatom are ready to develop and deliver arobotic system for sampling (inside the PCV) of fuel debris.
2-3. Developing of safety justification of debris removal (inside the PCV)
technology. Developing of debris removal technology (inside the PCV).
Developing of safety justification of debris removal (inside theRPV )technology. Developing of debris removal technology inside the RPV
Safety justification will allow to pick up equipment for dust suppression
when performing cutting. At the enterprises of Rosatom has extensive
experience in the development of safety justification and perform cutting of
SNF in a variety of environments (under water, in air).
2-4. Packaging of fuel debris
Packaging for fuel debris samples plays an important role in ensuring thesafe and efficient preparation of waste for interim storage or disposal.
3. Conclusion
Fuel debris sampling, removal and packing should be highly technical, effective
and safe. Rosatom enterprises at each step ready to provide own, tested andunique technologies and solutions to overcome the challenges of different
complexity.

78

R02

Options for Fukushima-Daiichi Liquid RW Management
Improvement
Sergei Semin, Sector Head, TENEX
Adamovich Dmitrii, PhD, Head of Department, NIKIMT

Abstract
To improve current Fukishima-Daiichi liquid RW management, to reduce secondary waste generation, toreduce operative cost for liquid RW treatment.
1. Introduction
Well known, that liquid RW (LRW) treatment contains not only water decontamination phase but
secondary waste conditioning phase too. And the main problem is how to prepare these secondary waste for
disposal requirements. In this case, you there are several points of costs: capital, operating and service costs.
During decades, ROSATOM develops and applies effective liquid RW treatment solutions in the different
cases like a legacy liquid RW treatment or new generated liquid RW treatment. ROSATOM has the modern
cheap and effective solutions.
2. Current status
The main disadvantages of this scheme are the large volume of secondary waste generation andconditioning issues. ROSATOM offers to improve some part of basic scheme, which contain exclusion numberof retrogradation steps.
3. Proposed improvements
The following points showed proposed scheme:
1.

Special reagents added with highly dispersed flock generation for Sr capturing;2.

Flock separation on ceramic and ultrafiltration membrane;

3.

Cs decontamination by finely divided sorbent and extra Sr sorbtion on flock;4.

Special containers with ferrocyanides sorbents disposap;

5.

Sludge and slurry solidification into cement matrix or calcination;

6.

Sludge and slurry solidification into cement matrix or calcination/

ROSATOM proposed improvements base on selective Sr isotopes concentration with inorganic collector
and Cs isotopes sorbtion on finely divided sorbent. Sr and Cs isotopes codeposition occurs when special
chemical reagents A and B are applying. Moreover, there is Sr-isotopes retrogtadation on highly dispersed flock.
We note that maximum effect is achieved with the optimum ration of A and B reagents. And next step is extra
Cs-isotopes decontamination.
Varying reagents volume and number of decontamination steps, you can achieve your own
decontamination requirements with minimal secondary waste volume.
For Cs isotopes decontamination, finely divided (50-200 μm) inorganic sorbent is added. This sorbent
captures Sr-isotops too. Only 5% of initial LRW volume is a secondary waste (sludge).
Thus, this proposed scheme allows reduce of Sr and Cs activity 725 and 7800 times respectively.
For extra Cs isotopes decontamination we offer to use ferrocyanides sorbents contain into the special
containers, which could be disposal immediately. The column volume of this sorbent is 7000-8000 for seawater.
5. Conclusions
Offering scheme allows reduce secondary waste volume and reduce treatment cost by using cheap reagentsand sorbents.

79

R03

Solid RW management system
Main technologies and approaches for solid RW treatment.
Oleg Bagryanov, Senior project manager, TENEX

Abstract
To reduce the release of radioactive and other toxic materials into the environment requires safety andeffective treatment methods. Treatment of solid radioactive waste involves the application of several steps:
decontamination, fragmentation, conditioning, rehabilitation and packaging. Each step must meet not only thesafety criteria but also the effectiveness criteria.
1. Introduction
Rosatom applies progressive approaches to overcome the challenges associated with decommissioning,radioactive waste management, territory rehabilitation and elimination of accident consequences. During
Russian nuclear industry development to overcome these challenges were designed various technologies thatwere tested in real conditions.
2. Main steps in solid RW treatment
2-1. Decontamination
Radioactive contamination of the working areas and equipment may result from normal operations or from
accidental release. Decontamination, therefore, form an essential part of a radiation protection, serving to ensureclean and safe working conditions.
2-2. Fragmentation
For implantation of work in high radiation fields to minimize dose to personnel and to reduce the potential
impact on the environment for distance fragmentation remotely controlled equipment should be used.
2-3. Conditioning
Before packaging radioactive wastes should be conditioned or to be brought in a convenient form forstorage. Volume of radioactive waste arising from decommissioning process or from accident also should be
reduced.
2-4. Rehabilitation
To significantly reduce a huge amount of radioactive soil
and rehabilitate the territory arising from decommissioning
process or from accident, it is necessary preliminarily
segregate a waste stream by activity and separate by size
particles.
2-5. Packaging
Packaging for radioactive wastes plays an important rolein ensuring the safe and efficient preparation of waste forinterim storage or disposal.
3. Conclusion
Radioactive waste treatment arising from regular decommissioning process or accident should be highly
technical, effective and safe. Rosatom enterprises at each step ready to provide own, tested and unique
technologies and solutions to overcome the challenges of different complexity.

80

"Effective design and assurance for permanent geologic storage ofUS01 contaminatedliquids: site selection, surveillance, and safe operations"
Ahmed Abou-Sayed1, JC Rogiers2, Satoshi Mochizuki3, Omar Abou-Sayed1
1

Advantek Waste Management Services LLC, 2Oklahoma University, 3Consultant

Abstract (60 words)
The proposed solution of subsurface injection has been reviewed (June 2016) by the Tritiated Water Task
Force. Their concerns were: no suitable geological structure, lack of monitoring technology and regulatory
hurdles. However, our well-established technology today in oil and gas industry would address these issues .

1. Introduction
The present efforts and technologies currently deployed to manage contamination at Fukushima site, workers
are exposed to radiation up to 8 times higher than the acceptable limit. It is also unavoidable to consider theradioactive water discharge to ocean during emergency. The accidental and planned diluted radioactive water
discharges continue to contaminate the ocean and land. The concentration level increases through the entire foodchains over the foreseeable future (500 to 1000 years). The challenge is to find a sustainable solution thatspans many generations of mankind for such a long-term risk issue. We believe that our proven technology ofgeologic disposal is, so far, the only technology that meets the challenge. We have more than 30 years ofexperience of such application and would like to share our proven practice in oil, gas and chemical industries.
Specifically, we refer to how we select the site, monitor, and operate safely through modeling, testing and realtime surveillance for process integrity assurance.

2. Technology
The site selection involves the detailed study of geological structural column at the selected location. Thisrequires seismic images, logging, core samples and modeling fluid and slurries flow in subsurface layers toguarantee containment and prevent any potential fault activation (breach or earthquakes), and to achieve
controlled fracture propagation, if any.

Pre-injection tests (fall-off test and step rate test) are performed tocalibrate simulation models. We have established this workflow over 30 years for many fields around the world.
Injection fluids and slurries (viscosity, density and particle size).are designed to optimize the injection and safety
The monitoring is done at the surface and near the injection point (fiber optics pressure gauges). The
pressure response from the injection is continually analyzed to ensure the safe operation. Our industry-leading
proprietary modeling technology plays the key role in interpreting data and taking actions.
We have worked with many governmental agencies on the legal framework around the world to apply thistechnology. We believe our experience would help formulate regulations in Japan.
3. Conclusion
The cesium contamination is spread to nearly entire Pacific Ocean. The dilution to legal limit is not theideal solution because it concentrates through the entire food chain. Our technology is considered “risky” amongsome experts, but the damage that the current solution provides is extensive, and not sustainable for thousandsof years.

Our technology can provide much safer, economical and sustainable long-term solution.

81

US02

Nuclear Decommissioning and Risk Managementfrom 30 Years of Experience
Tim Carraway, Stan Hodges, Gavin Winship, and Max Ehrhardt
AECOM

Abstract
This poster illustrates AECOM’s decommissioning and dismantlement (D&D) program and approach. Thisapproach includes all necessary elements to estimate and execute a complex D&D project. AECOM’s provenapproach is based on 30+ years of experience at commercial reactor and high-hazard decommissioning sites.
1. Introduction
In addition to selecting the appropriate D&D technology, an accurate cost estimate and schedule is a crucial
factor in completing large, complex D&D projects. AECOM’s database of experience guided by a provenproject management approach with an integrated ISO 31000-compliant Risk Management program produces
accurate estimates and allows efficient execution throughout a D&D project.
2. Decommissioning Project Management - Elements of an effective project execution plan2-1. Baseline Development
By analyzing past performance and lessons learned, critical quantities and characteristics (labor, materials,
equipment, D&D production rates for various hazard levels, etc.) are estimated. Key personnel within AECOM
and the Site Owner with an essential base of knowledge are identified as the project management team.
2-2. Cost Estimating
Costs are estimated based on the Work Breakdown Structure identified during the baseline analysis and thenintegrated into customized D&D estimating software. The estimate is supported by input from D&D Subject
Matter Experts and validated by comparison with prior estimates in AECOM’s D&D Database of Knowledge.
2-3. Scheduling
Combining AECOM’s Database of Knowledge with commercially available scheduling software, the entire
work scope is logically sequenced. The activities and cost estimates are shared between the schedule and cost
software, creating a fully integrated cost andresource-loaded schedule.
2-4. Risk Management
Experienced Subject Matter Experts identify
risks in the performance baseline and determine
their likelihood and impact to the D&D project.
Methods to mitigate or avoid each risk aredeveloped. Using a Monte Carlo analysis of therisks, an appropriate amount of reserve funding(contingency) is calculated for emergent work.

Figure 1. Monte Carlo analysis for contingency funding
3. AECOM employs this project management approach on every large D&D project3-1. AECOM’s D&D Database of Knowledge is the foundation of AECOM’s D&D program
With 30+ years of accumulated knowledge in procurement, design, fabrication, and operations in all stagesof D&D projects at both Commercial Nuclear Power Plants and DOE/NDA facilities, AECOM incorporates
actual results from completed projects to develop each element of the D&D execution plans.
3-2. AECOM’s D&D Approach is integrated into
the ETTP and San Onofre projects
AECOM’s project management system (via its
affiliate UCOR) was used to construct a Level 9 WBS
with over 2000 activities at the East Tennessee
Technology Park (ETTP) site. This approach has been
proven in executing the Oak Ridge Reservation
projects on-time and under budget. It is also used as
the baseline for the SONGS D&D project.
Figure 2. Contract performance at ETTP 2011-2015
4. Conclusion
AECOM’s project management approach along with access to several decades of nuclear D&D experiencecan produce a cost estimate and schedule that properly accounts for and minimizes project risk. AECOM’sstrong proven resume and confidence to share risks with the site owners is a key feature that distinguishes
AECOM from other D&D companies.

82

US03

EnergySolutions Nuclear Power Plant Decommissioning Experience
Colin Austin1 and Makoto Kikuchi2
1

EnergySolutions, 2 MKC Consulting

Abstract
EnergySolutions (ES) is an international leader in commercial nuclear power plant (NPP) decommissioning.
It has developed this experience over a period of 30 years performing NPP decommissioning and related nuclear
waste management services in the USA and Europe. This presentation will provide some background on ES,
some specific examples of its decommissioning experience in the USA, and identify benefits that this experience
brings to NPP decommissioning projects.

Scope of Work and Experience
ES was formed 10 years ago and is now made up of over 10 different companies that have been integrated
to provide the complete value chain of services necessary to successfully perform NPP decommissioning
projects. The companies that make up ES have a rich history in the industry going back as far as 60 years. The
scope of work that ES has been involved in includes everything necessary to successfully perform NPP
decommissioning including:•Regulator compliant estimates for decommissioning funding,•

Detailed decommissioning planning and license application,•

Spent fuel management, handling and storage,•

Decontamination and dismantling,•

Equipment, facility and site survey and characterization,•

Waste packaging, transportation and logistics,•

Waste treatment and disposal, and,•NPP site and waste management facility operating license holder.

This experience allows ES to understand and optimize the NPP decommissioning strategy to ensure
projects are performed safely, in full compliance with all regulatory and permitting requirements and at thelowest cost. The company has developed different technical, contracting, business and licensing models tominimize both owner and contractor risk. Examples include:•

Transfer of the nuclear licenses at Zion and La Crosse from the nuclear utility to ES the D&D
contractor,•

The first and only successful dismantlement of two NPP at the same time,•

Largest and fasted spent fuel transfer campaign, and•First segmentation of large NPP reactor vessel in the USA, completed in 1 month.

Many of the lessons involve cultural change from that of a nuclear operator or an engineering projectcompany to that of a nuclear facility dismantling and waste management project. The key is to recognize that a
NPP decommissioning project is primarily a nuclear waste management project. The ES approach significantly
reduces project risk and ensures safe and compliant performance for the lowest cost.
ES hopes to work with JAPC and the Fukui Prefecture to ensure safe compliant and effective execution of
the Tsuruga-1 project, and to make available its experience to ensure the effectiveness of the NPP
decommissioning program in Japan.

83

List of Organizations
Organization

ID No.

ADVAN ENG. Co., Ltd.

L07

Advantek Waste Management Services LLC

US01

AECOM

US02

Amec Foster Wheeler

N03, GB01,
GB02, GB03,
GB04, GB05,
GB06, GB07,
GB08

Amec Foster Wheeler Nuclear Slovakia s.r.o.

L01

ANADEC

F04

AREVA ATOX D&D SOLUTIONS Co. Ltd.

B03

AREVA NC

F04, F05

Atomic Energy Society of Japan (AESJ)

L08

ATOX Co. Ltd

L04

Azmec Inc.

K01

British Embassy Tokyo

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

Cavendish Nuclear Ltd

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

CEA (The French Alternative Energies and Atomic Energy Commission) F01, F04, F05,
F06
Central Research Institute of Electric Power Industry (CRIEPI)

D02, K02, K05

CHUBU Electric Power Co., Inc.

A01

Cornes Technologies Limited

B04

CREATEC

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

DBD International

GB11

EBARA Corporation

K04, L06

84

ECM Technologies

F05

EnergySolutions

US03

Fukushima Consortium of Robotics Research for Decommissioning and Disaster
Response

B06, B07

Fuji Electric Co., Ltd.

L01

Fukushima University

B02, C04, C05,
C06, K07

Hazama Ando Corporation

H01

Hirosaki University

C03

Hiroshima Institute of Technology

C07

Hitachi Ltd.

A05, K08

Hitachi Chemical Techno Service Co., Ltd.

K07

Hitachi GE Nuclear Energy, Ltd

A03, K08

Ibaraki University

K07

IER Fukushima University

C04, C05, C06

Innovative Physics Limited

GB10

Institut de Radioprotection et de Sûreté Nucléaire (IRSN)

F01

Institute for Energy Technology / OECD Halden Reactor Project

B05

Institute for Safety Problems of Nuclear Power Plants National Academy of Sciences of
Ukraine (ISP NPP)

N09

International Nuclear Services

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

International Research Institute for Nuclear Decommissioning (IRID)

A03, N02

James Fisher Nuclear Ltd

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

Japan Agency for Marine-earth Science and Technology (JAMSTEC)

C04, C05

Japan Atomic Energy Agency (JAEA)

A02, A04, B01,
C08, L07, N01

JFE Engineering Corporation (JFEE)L02

JGC Corporation

N08

Kajima Corporation

E02, K06

Kao Corporation

D01

KOBELCO STUDSVIK Co., Ltd.

L10

Kurion Japan K.K.

N07

Laser Decontamination and Decommissioning (LDD) Corporation

J01

85

Mirion Technologies (Canberra) KK

C01, C02

Mitsubishi Heavy Industries, Ltd.

N06

Mitsubishi Materials Corporation

L05

Mitsubishi Research Institute, Inc.

N05

MKC Consulting

US03

Nagaoka University of Technology

L07

Nagoya University

A01

National Institute of Technology, Fukushima College

M01

National Institute of Technology, Ibaraki College

C07

National Nuclear Laboratory (NNL)

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

National Radioactive Waste Management Agency (ANDRA)

F05

NIKIMT

R02

Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF)
H01, K02
NUKEM Technologies Engineering Services GmbHJ03

NUVIA France

F02, F03

NUVIA

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08,
GB09

Obayashi Corporation

N03

Oklahoma University

US01

ONET Technologies

F01

Optimum Water System

F02, F03

Oxford Technologies Ltd

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

Perkin Elmer

C04, C06

Sellafield Ltd.

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB08

Shibaura Institute of Technology

M02

Silica Material LLC

K09

86

TENEX

R01, R02, R03

The Japan Atomic Power Company (JAPC)L02

The University of Tokyo

C04, C06, H02

The Wakasa Wan Energy Research Center

J02

Tohoku University

D03, L09

Tokai University

M02

Tokyo City University

L09, M02

Tokyo Electric Power Company Holdings

D02, E01, K03,
K08, L03, N04

Tokyo Institute of Technology

K09, L09, M02

Tokyo Medical and Dental University

M02

Tokyo Power Technology Ltd.

C03

Toshiba Corporation

A03, K03

VEOLIA

F04

Waseda University

K01

WS Atkins

GB01, GB02,
GB03, GB04,
GB05, GB06,
GB07, GB0887