EAFORM2017

6th East Asia Forum on
Radwaste Management Conference

November 27-29, 2017
Hotel Granvia Osaka, 3-1-1 Umeda, Kita-ku,
Osaka, Japan

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General Objective of EAFORM................................................................................... 1
Organization of 6th Conference .................................................................................. 2
Conference Venue and Halls ...................................................................................... 3
Time Table (Program Summary) ................................................................................. 4
Program ...................................................................................................................... 7
Presentation materials for the Plenary Session ( 6 Materials ) ................................. 19
Presentation materials for the Special Session ( 4 Materials ) .................................. 73
Abstracts for the Technical Sessions ...................................................................... 123

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November 27-29, 2017
Hotel Granvia Osaka, 3-1-1 Umeda, Kita-ku, Osaka, 530-0001, JapanG
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Safe, secure, and sustainable solutions to the management of spent nuclear fuels andradioactive wastes are critical to the viability and continuing peaceful use of nuclear power andother advanced nuclear applications. With no exception, current nuclear energy programs in the
East Asian area are all faced with such challenges. Although each of these programs has developed
its own technology base, implementation strategy, and infrastructure, the urgent need for further
technology enhancement on radioactive waste management and implementation of radioactive
waste disposal are becoming more essential to all. This is especially true after the accident at the
Fukushima Daiichi Nuclear Power Plant. Enhancing international collaboration will benefit
participants through the sharing of knowledge and expertise, as well as other intangible benefits
such as public comprehension of safety issues and support for repository establishment.
Recognizing these facts, several East Asian research organizations agreed to hold the East
Asia Forum on Radwaste Management (EAFORM) to enhance future cooperation. EAFORM isprincipally opened to all research organizations and others related to radioactive waste management.
The purpose of EAFORM is to promote information exchange, to share knowledge, and toencourage international collaboration in a bilateral or multilateral mode.
Under this configuration, the “EAFORM Management Committee” was organized todevelop activities aimed at strengthening the above purpose. The 1st EAFORM conference was heldin Taiwan in 2006.

The 6th EAFORM conference will be held in Osaka from November 27 to 29,

2017. Conference participants are encouraged to focus their submissions on the technical topics ofthis conference.

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 Hosted by
Atomic Energy Society of Japan Organized by
EAFORM Management Committee (*)
Division of Nuclear Fuel Cycle and Environment, Atomic Energy Society of Japan* Currently, “EAFORM Management Committee” consists of following members (listed inalphabetical order)- Atomic Energy Society of Japan (Division of Nuclear Fuel Cycle and Environment,
NUCE/AESJ)- China Institute of Atomic Energy (CIAE)- Chung Hwa Nuclear Society (CHNS)- Institute of Nuclear Energy Research (INER)- Korea Radioactive Waste Agency (KORAD)- Sandia National Laboratory (SNL) [Technical supporting member]2

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanC
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 Conference Venue
Hotel Granvia Osaka
3-1-1 Umeda, Kita-ku, Osaka, 530-0001, Japan
http://www.hotelgranviaosaka.jp/maps/maps-transportation.htm(+81)-6-6344-1235, granvia@hgo.co.jp

 Floor Map of the Conference Halls (20th Floor)

Hall-S

Hall-A

Houou

Naniwa-c
Hall-CHall-B

KakujuKujaku3

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

TTiim
mee TTaabbllee ((PPrrooggrraam
mS
Suum
mm
maarryy)) On-site Registration Desk (20th Floor of Hotel Granvia Osaka)
Open between 16:00 to 20:00 on Sunday 26 November 2017
Open from 8:15 on Monday 27 November 2017

 Day 1: Monday 27 November 2017
Time
08:15-09:00 Registration (Local Space on the 20th floor)
Hall S (Houou)
09:00-09:10 Opening Remarks
Mitsuo Takeuchi, Chair, The 6th EAFORM Organizing Committee
09:10-09:50 Plenary Session 1(1) Katsumoto Yoshimura, Director, Technology Office for RWM and Public Relations
Office for RWM, METI, Japan(2) Yaohiro Inagaki, Prof. (Kyushu Univ.) and Director, Division of Nuclear Fuel Cycle
and Environment, AESJ, Japan
2 speeches will be provided(20min. Each)
09:50-11:10 Plenary Session 2(1) Ching-Tsuen Huang, Prof. (National Tsing Hua Univ.) and Chairman, Academic
Committee on Radwaste Management, CHNS, Taiwan(2) Joo-Wan Park, Vice President of Radioactive Waste Project Division, Korea
Radioactive Waste Agency (Korad), Korea(3) Todd R. Zeitler, WIPP Performance Assessment Technical Team Lead, Sandia
National Laboratories, USA(4) (TBD from China)
4 speeches will be provided(20min. Each)
11:10-11:30 Coffee Break
11:30-12:50 Special Session:
Toward Revitalization of Fukushima
4 speeches will be provided(20min. Each)
12:50-14:00 Lunch (Hall A, Hall B, Hall C)
Time
Hall A (Naniwa-C)
Hall B (Kujaku)
Hall C (Kakuju)
14:00-16:05 Technical Session 1A
Technical Session 1B
Technical Session 1C
Decontamination &Radioactive Waste
Radioactive Wastes
Decommissioning-1
Management Policies,
Treatment
Chair; Hitoshi Owada
Chair; Shinichi Nakayama Regulations and Programs-1
Chair; Motoi Kawanishi
3 Papers
5 Papers
25 min. each including QA
for 5min.

16:05-16:20 Coffee Break
16:20-18:00 Technical Session 2A
SNF/HLW Disposal-1
Chair; Shinzo Ueta
4 Papers
25 min. each including QA
for 5min.

25 min. each including QA
for 5min.

3 Papers
25 min. each including QA for
5min.

Technical Session 2B
Decontamination &Decommissioning-2
Chair; Satoshi Yanagihara

Technical Session 2C
Radioactive Waste
Management Policies,
Regulations and Programs-2
Chair; Kenichi Kaku

3 Papers
25 min. each including QA
for 5min.

18:30-20:30 Welcome Dinner (Hall Naniwa)

4

3 Papers
25 min. each including QA for
5min.

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan Day 2: Tuesday 28 November 2017
TimeHall A (Naniwa-C)
08:30-10:10 Technical Session 3A
SNF/HLW Disposal-2
Chair; Morimasa Naito

Hall B (Kujaku)
Technical Session 3B
LLW/ILW Disposal-1
Chair; Teruyuki Yamada

Hall C (Kakuju)
Technical Session 3C
SNF/HLW Disposal-3
Chair; Yoichi Yamamoto

4 Papers
25 min. each including QA
for 5min.

3 Papers
25 min. each including QA
for 5min.

4 Papers
25 min. each including QA
for 5min.

10:10-10:30 Coffee Break
10:30-11:45 Technical Session 4A
SNF/HLW Disposal-4
Chair; Morimasa Naito
3 Papers
25 min. each including QA
for 5min.

11:45-13:00 Lunch (Hall S)
Time
Hall A (Naniwa-C)
13:00-14:40 Technical Session 5A
SNF/HLW Disposal-6
Chair; Hiroyuki Tsuchi
4 Papers
25 min. each including QA
for 5min.

14:40-14:55 Coffee Break
14:55-16:35 Technical Session 6A
Radioactive Waste
Management Policies,
Regulations andPrograms-3
Chair; Hiroyoshi Ueda

Technical Session 4B
LLW/ILW Disposal-2
Chair; Minoru Emori

Technical Session 4C
SNF/HLW Disposal -5
Chair; Hiromitsu Saegusa

2 Papers
25 min. each including QA
for 5min.

2 Papers
25 min. each including QA
for 5min.

Hall B (Kujaku)
Technical Session 5B
SNF/HLW Disposal-7
Chair; Takamitsu Ishidera

Hall C (Kakuju)
Technical Session 5C
SNF/HLW Disposal-8
Chair; Tetsuo Fujiyama

4 Papers
25 min. each including QA
for 5min.

3 Papers
25 min. each including QA
for 5min.

Technical Session 6B
SNF/HLW Disposal-9
Chair; Masamichi Obata

Technical Session 6C
SNF/HLW Disposal-10
Chair; Soshi Nishimoto

3 Papers
25 min. each including QA
for 5min.

3 Papers
25 min. each including QA
for 5min.

4 Papers
25 min. each including QA
for 5min.

16:35-16:45 Short Break
16:45-17:00 Closing Session
Closing Remarks

56th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan Day 3: Wednesday 29 November 2017
One-day technical tour (optional)◆08:40Gathering & Departure from Hotel Granvia Osaka to;1) HOKUDAN EARTHQUAKE MEMORIAL PARK ( http://www.nojima-danso.co.jp/ )2) The Great Hanshin-Awaji Earthquake memorial museum ( http://www.dri.ne.jp/en )3) Short break at (stop by) KOBE KITANO IJINKAN-GAI ( http://www.kobeijinkan.com/ )◆17:45Return back to Hotel Granvia Osaka (expected time)

1st Meeting place at 08:40 is “Meeting place (see following floor map)” on the Grandfloor of Hotel Granvia Osaka[Floor Map of the Grand Floor (1st floor) of Hotel Granvia Osaka]

To the central entranceof JR Osaka station

Café Restaurant Ripple

Meeting place
at 8:40

Hotel Granvia Osaka

6

Program

76th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan Day 1: Monday 27 November 2017
Plenary Session 1 9:10-9:50 [Hall S] Chair, Reiko Nunome (NUMO)
Status of Radioactive Waste management in Japan

Chair, Reiko Nunome (NUMO)(1) Nationwide Map of Scientific Features for Geological Disposal of High-Level

Radioactive Waste in Japan
Katsumoto Yoshimura (Director, Technology Office for RWM and Public Relations Office for RWM,
Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry, Japan)(2) How to Integrate Various Research Fields for Reasonable and Reliable Radwaste
Management
Yaohiro Inagaki (Prof. (Kyushu Univ.) and Director, Division of Nuclear Fuel Cycle and Environment,
AESJ, Japan)

Plenary Session 2 9:50-11:10 [Hall S]
Status of Radioactive Waste management in Taiwan, Korea, China and USA

Chair, Reiko Nunome (NUMO)(1) Current Situation of Radioactive Wastes Management in Taiwan
Ching-Tsuen Huang (Prof. (National Tsing Hua Univ.) and Chairman, Academic Committee on
Radwaste Management, CHNS, Taiwan)(2) Status of High-level Radioactive Waste Management Program in Korea
Joo-Wan Park, Vice President of Radioactive Waste Project Division, Korea Radioactive Waste
Agency (Korad), Korea(3) Recertification of WIPP: From Submittal of the CRA-2014 to an EPA Recertification Decision
Todd R. Zeitler ( WIPP Performance Assessment Technical Team Lead, Sandia National
Laboratories, USA)(4) (TBD from China)

Special Session 11:30-12:50 [Hall S]
Toward Revitalization of Fukushima
Chair : Motoi Kawanishi (ASANO TAISEIKISO ENGINEERING Co.)
(1) Technical Strategy for Decommissioning of the Fukushima Daiichi NPS

Kazuyuki Kato (Managing Director, Technological Strategy Group, Nuclear Damage
Compensation and Decommissioning Facilitation Corporation)

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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan(2) Study on Post-Accident Waste Management Scenarios for Fukushima Daiichi Nuclear

Power Station
Satoshi Yanagihara (Prof. (Univ. of Fukui) and Executive committee member, Review
Committee on Decommissioning of the Fukushima Daiichi Nuclear Power Station, AESJ)(3) JAEA's R&D for Decommissioning of Fukushima Daiichi Nuclear Power Station
Shinichi Nakayama (Director General, Fukushima Research Institute, JAEA)(4) Distribution of Radionuclides Nearby the Fukushima Daiichi Nuclear Power Station
Terumi Dohi (Assistant Principal Engineer, Fukushima Environmental Safety Center,
Fukushima Research Institute, JAEA)

Technical Session 1A 14:00-16:05 [Hall A]
Radioactive Wastes Treatment: Source Term & Waste Treatment
Chair ; Hitoshi Owada (RWMC)
1A-1

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

1A-2

C-14 Release Behavior and Thermal Decomposition Characteristics of Crud Particles
Collected from the Coolant Filter of Commercial PWR
Shintaro Tsuji, Toshikazu Waki, Akira Sakashita and Kenichiro Kino

1A-3

Evaluation of Carbon-14 Release from Irradiated Zircaloy Fuel Cladding through aLong-term Static Leaching Test
Hiroyoshi Ueda, Tomofumi Sakuragi, Naoki Fujii and Hitoshi Owada

Technical Session 1B 14:00-16:05 [Hall B]
Decontamination & Decommissioning: Decontamination-1: Decontamination
Chair ; Shinichi Nakayama (JAEA)
1B-1

The Assessments and Comparisons of RASCAL and EPZDose on Atmospheric
Dispersion and Dose Consequences in Radioactive Material Release Accidents
Shaohsuan Chen, Jongrong Wang, Chunkuan Shih, YuTing Ku, Yu Chiang, ShaoWen Chen
and WenSheng Hsu

1B-2

Confirmation of Appropriate Operation Condition with Blasting Device
Yuuki Tsuhara, Hirotaka Izuka, Shinro Hirano, Yuji Matsunaga, Yasuhiro Sugahara
and Yoji Kanamori

1B-3

Development of In-situ Radioactivity Inspection System for Radioactive Waste andDecontamination System Using Microalgae
Sujung Min, Dohyung Kim, Woyoung Kim, Sheunghyun Ha, San Chae, Jongman Lee
and Unjang Lee

106th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

1B-4

Rapid Removal of Cesium-137 from Urban Area After the Fukushima Dai-ichi Nuclear
Power Plant Accident
Kazuya Yoshimura

1B-5

A Study on the Performance of Flocculating Agent for Radioactively Contaminated Soilby Soil Washing Process
Jong Soon Song and Sun Il Kim

Technical Session 1C 14:00-16:05 [Hall C]
Radioactive Waste Policies, regulations and Programs-1
Chair : Motoi Kawanishi (ASANO TAISEIKISO ENGINEERING Co.)
1C-1

Spent Nuclear Fuel Final Disposal Management in Taiwan
Allan Lee

1C-2

Scientific Basis for Nationwide Screening of Geological Disposal Sites in Japan
Takehiro Matsumoto, Hiromitsu Saegusa, Hideaki Hyodo, Akira Deguchi and Hiroyuki Umeki

1C-3

Sensitivity of WIPP Performance Assessment Results to Regulator-prescribed Changes
Todd R. Zeitler and Brad A. Day

Technical Session 2A 16:20-18:00 [Hall A]
SNF/HLW Disposal-1: Modelling (THMC, etc.)
Chair : Shinzo Ueta (MITSUBISHI MATERIALS Co.)
2A-1

Geochemical Modeling of Water-rock Interactions in Granitic Rocks, Eastern Taiwan
Yu-Te Chang

2A-2

Centrifuge Model Test to Gain Reliability of the Future Prediction in Terms of Long Term
THM Processes in Deep Geological Repository
Soshi Nishimoto

2A-3

Thermo-hydro-chemical Processes Influence on Buffer Material Degradation in HighLevel Radioactive Waste Disposal
Wen-Sheng Lin, Suu-Yan Liang and Chen-Wuing Liu

2A-4

Coupled THMC Processes in Radionuclide Waste Management
Gour-Tsyh (George) Yeh and Chia-Hsing (Peter) Tsai

Technical Session 2B 16:20-18:00 [Hall B]
Decontamination & Decommissioning: Decommissioning-2: Decommissioning
Chair : Satoshi Yanagihara (Univ. of Fukui)
2B-1

Efforts toward Safe, Steady and Efficient Decommissioning

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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Toyoaki Yamauchi, Satoshi Karigome and Hitoshi Ohata
2B-2

Decommissioning Planning for Unit 1 and 2 at Mihama Nuclear Power Plant
Shinya Kato, Koichi Kamahori, Hiromu Isaka and Shinro Hirano

2B-3

Radiation Survey and Waste Inventory Estimation for Decommissioning of Taiwan NPP
Chun-Cheng Lin, Wei-Sin Chen, Jyi-Lan Wuu and Shu-Jun Chang

Technical Session 2C 16:20-18:00 [Hall C]
Radioactive Waste Management Policies, Regulations and Programs-2:
Chair : Kenichi Kaku (NUMO)
2C-1

Status of Radioactive Waste and Spent Nuclear Fuel Management in Finland
Jari Tuunanen

2C-2

Communication Activities through Dialogue in Japanese Geological Disposal Project ofHigh-level Radioactive Waste
Ayako Araki, Saki Ikeda, Kumiko Ezaki and Kenichi Kaku

2C-3

Efforts in the Field of Education for Japanese Geological Disposal of High-level
Radioactive Waste
Kumiko Ezaki, Kenichi Kaku and Shunsuke Suzuki

126th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan Day 2: Tuesday 28 November 2017
Technical Session 3A 08:30-10:10 [Hall A]
SNF/HLW Disposal-2 : Safety Case Development-1
Chair : Morimasa Naito (JAEA)
3A-1

Overview of the NUMO Safety Case at Pre-siting Stage
Tetsuo Fujiyama, Satoru Suzuki, Akira Deguchi and Hiroyuki Umeki

3A-2

Geosynthesis of a State-of-the-art Knowledge Base into SDMS in the NUMO Safety Case
Kunio Ota, Hiromitsu Saegusa, Takanori Kunimaru and Saori Yamada

3A-3

Layout Design of Underground Facilities Tailored to SDM in the NUMO Safety Case
Takahiro Goto, Yoichi Yamamoto, Satoru Suzuki and Shigeru Kubota

3A-4

Preliminary Study of Pre-closure Safety Assessment in the NUMO Safety Case
Kazuhisa Yamashina, Satoru Suzuki and Shigeru Kubota

Technical Session 3B 08:30-10:10 [Hall B]
LLW/ILW Disposal-1
Chair : Teruyuki Yamada (FEPC)
3B-1

Conceptual Study on Disposal Facility for Waste from Decommissioning of NPPs
Norie Kawahata, Ryoichi Hojo, Takayuki Hamanaka and Takashi Kozawa

3B-2

An Integrated Framework for Simulating Radionuclide Decay Transport of Low-level
Radioactive Waste with Tunnel Disposal in Nearshore Environment
Chin-Chang Lu, Ming-Hsu Li and Yu-Ru Chen

3B-3

Sensitivity Analysis of Simulating Radionuclide Decay Transport of Low-level
Radioactive Waste in Nearshore Environment
Yu-Ru Chen, Chin-Chang Lu and Ming-Hsu Li

Technical Session 3C 08:30-10:10 [Hall C]
SNF/HLW Disposal -3 : Barrier Technology (Overpack etc. )Chair : Yoichi Yamamoto (NUMO)
3C-1

Seismic Response of Canister in Buffer Material Under Water Invasion Condition by
Centrifuge Modeling
W.Y. Hung, J.J. Xu, Y.C. Wu and M.H. Hsieh

3C-2

Study on the Applicability of Cast Steel Overpack –Evaluation of Casting Defects andCorrosion Resistance Using Full-scale Prototype–
Yusuke Ogawa, Satoru Suzuki, Kazuhisa Yamashina and Shigeru Kubota

136th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

3C-3

Structural Integrity Assessment of Disposal Package for Radioactive Waste Failure
Assessments for Overpack Using Finite Element Analysis
Masahiro Kawakubo and Masato Kobayashi

3C-4

Stress Analysis for the Canister Under Earthquake Induced Fracture Shear
Displacement considering Long-term Creep Effect of Copper Shell
Yuan-Chieh Wu, Hsi-Ting Lo and Yi-Ren Wang

Technical Session 4A 10:30-11:45 [Hall A]
SNF/HLW Disposal -4 : Safety Case Development-2
Chair : Morimasa Naito (JAEA)
4A-1

Preliminary Study of Post-closure Safety Assessment in the NUMO Safety Case
Keisuke Ishida, Takafumi Hamamoto, Sanae Shibutani, Kiyoshi Fujisaki and Motoyuki Yamada

4A-2

Assessment of Sorption and Diffusion in the Rock Matrix in the NUMO Safety Case
Takafumi Hamamoto, Sanae Shibutani, Keisuke Ishida, Kiyoshi Fujisaki, Motoyuki Yamada
and Yukio Tachi

4A-3

How Will New Knowledge Be Reflected to the Management of Geological Disposal? Influence of an FEP Not Considered So Far to Some Sub-scenarios Koichi Shin

Technical Session 4B 10:30-11:45 [Hall B]
LLW/ILW Disposal-2
Chair: Minoru Emori (RWMC)
4B-1

Barrier of Near-surface LILW Disposal Facility in Korea: Hydraulic Model Development
Mi-Jin Kwon, Ki-Jung Kwon and Jin Beak Park

4B-2

Numerical Simulation of the Thermal Performance of a Dry Storage Cask for Spent
Nuclear Fuel
Chi-Ming Lai, Ron-Horng Chen, Chu-Tsen Liao, Ting-Yi Wang and Heui-Yung Chang

Technical Session 4C 10:30-11:45 [Hall C]
SNF/HLW Disposal -5 : Site Investigation Technology-1
Chair : Hiromitsu Saegusa (NUMO)
4C-1

Influence of inherent and Stress-induced Anisotropy of Hydraulic Conductivity on the
Groundwater Flow Around a Rock Tunnel
Huang-Kuei Chu, Po-Sung Lai and Jia-Jyun Dong

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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

4C-2

Data Qualification Methodology in the Literature Survey Stage
Saori Yamada, Takanori Kunimaru, Kunio Ota, Stratis Vomvoris and Niels Giroud

Technical Session 5A 13:00-14:40 [Hall A]
SNF/HLW Disposal -6 : R&D by URL
Chair : Hiroyuki Tsuchi (SHIMIZU Co.)
5A-1

Mizunami Underground Research Laboratory Project - Achievement During Phase I/II
and Important Issues for Phase III Katsuhiro Hama

5A-2

Virtual Reality Geological Modeling for the Horonobe Underground Research Project
Takayuki Motoshima, Makito Nago and Yuji Ijiri

5A-3

Latest Rock Grouting Technologies Under Sea Water in Nordic Countries and Japan
Masakuni Tsuji, Mitsunobu Okihara, Hitoshi Nakashima, Toshinori Sato and Kazuhei Aoyagi

5A-4Strategic and Technical Aspects in RD&D Program Development for HLW Disposal
System in Korea
JeongHyoun Yoon, JeongHwan Lee and SeungHyun Kim

Technical Session 5B 13:00-14:40 [Hall B]
SNF/HLW Disposal -7: Barrier Performance (Buffer-1)
Chair : Takamitsu Ishidera (JAEA)
5B-1

Study on Gas Migration Behavior through Bentonite Buffer Material
Tomoyuki Shimura, Shinichi Takahashi, Masanobu Nishimura, Kazumasa Koga
and Hitoshi Owada

5B-2

Migration Experiment of Void Air in Buffer Material During Seepage
Naohiko Takamoto, Ichizo Kobayashi, Masahiro Kawakubo, Masataka Imai and Tomoko Ishii

5B-3

Study on Piping and Erosion of Buffer Material During the Re-saturation Period
Takayuki Abe, Ichizo Kobayashi, Masahiro Kawakubo, Masataka Imai and Tomoko Ishii

5B-4

Numerical Analysis of Inflow Control for Quality Management of Buffer Material Using
Discrete Fracture Network Model
Kazuhiko Masumoto, Makoto Nakajima, Hayato Nonaka, Tomoko Ishii, Mayumi Jo,
Masataka Imai and Hiroyuki Atsumi

Technical Session 5C 13:00-14:40 [Hall C]
SNF/HLW Disposal -8 : Nuclear & Disposal Related Technology
Chair : Tetsuo Fujiyama (NUMO)

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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

5C-1

Development of Wireless Monitoring Systems for Geological Disposal
Kazuhiro Tsubono, Masato Kobayashi, Hiromitsu Yamakawa and Akinori Hasui, Kazuhiko
Masumoto, Tomoaki Matsushita, Norihisa Sugahara,Tatsuya Tanaka and Chiaki Nagai

5C-2

Feasibility Study of Ventilation Design for Underground Facilities
Naoki Katsumata, Ryoji Yahagi, Hiromi Kurosaki and Shigeru Kubota

5C-3

Establishment and Application of Control Room Habitability Methodology for Maanshan
Nuclear Power Plant
Kai-Chun Yang, Jong-Rong Wang, Hsiung-Chih Chen, Shao-Wen Chen, Chunkuan Shihand Wen-Sheng Hsu

Technical Session 6A 14:55-16:35 [Hall A]
Radioactive Waste Management Policies, Regulations and Programs-3:
International Cooperation
Chair : Hiroyoshi Ueda (RWMC)
6A-1

International Cooperation Activities of KORAD Including Mid-term and Long-term
Strategy of International Cooperation for Radioactive Waste Management
Ou Jeong Yoo and Gyeong Hwan Park

6A-2

Metadata in Geological Disposal of Radioactive Waste: The RepMet Libraries
Russell C. Camphouse. Massimo Ciambrella and Kevin McMahon

6A-3

Metadata in Geological Disposal of Radioactive Waste: The RepMet Initiative
Massimo Ciambrella, Kevin McMahon and József I. Fekete

6A-4

Spent Nuclear Fuel Final Disposal Knowhow, Methodology and Technology Transfer
Magnus Holmqvist

Technical Session 6B 14:55-16:35 [Hall B]
SNF/HLW Disposal -9 : Barrier Performance (Buffer-2)
Chair : Masamichi Obata (TOSHIBA Co.)
6B-1

Sulfide Corrosion by Sulphate-Reducing Bacteria in MX-80 Bentonites
Szu-wei Lee and Ching-Tu Chang

6B-2

Adsorption of Uranium(VI) on the MX-80 Bentonite
Ying-Chieh Lin

6B-3

Am(III)/Nd(III) Interactions with Borate: Experimental Investigations ofNd(OH)3(micro cr) Solubility in NaCl Solutions in Equilibrium with BORAX
Yongliang Xiong, Leslie Kirkes, Cassie Marrs and Jandi Knox

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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Technical Session 6C 14:55-16:35 [Hall C]
SNF/HLW Disposal -10 : Site Investigation Technology-2
Chair : Soshi Nishimoto (CRIEPI)
6C-1

Microbial DNA; A Brand-new Tracer of Groundwater Flow
Ayumi Sugiyama, Kenji Kato, Kazuyo Nagaosa, Tetsuo Ibara, Kazuyoshi Takenobu andAtsunao Marui

6C-2

Seismic Analysis for the Deposition Tunnel in Fractured Rock by 3DEC
Meng-Hsiu Hsieh and Yuan-Chieh Wu

6C-3

Groundwater Flow Analysis for Evaluating Factors on Water Inflow to the Facility
During the Operation Period
Masataka Imai, Minoru Emori, Tomoko Ishii, Masahiro Kawakubo, Kazuhiro Tsubono,
Masayoshi Yamaura, Tsuneyuki Maemura and Eiko Sugimoto

17

Presentation materials for the
Plenary Session( 6 Materials )19Nationwide Map of Scientific Features for Geological Disposal of
High-Level Radioactive Waste in Japan

Agency for Natural Resources and Energy
Ministry of Economy, Trade and Industry
Katsumoto Yoshimura

History of Legislation on Geological Disposal of High-Level Radioactive Waste
1976: JAEC “Report on Radioactive Waste Management”
 PNC and JAERI (now JAEA) started R&D for HLW geological disposal*JAEC: Japan Atomic Energy Commission, PNC: Power Reactor and Nuclear Fuel Development Corporation,
JAERI: Japan Atomic Energy Research Institute, JAEA: Japan Atomic Energy Agency

1998: JAEC “Report on Basic Concept of HLW Geological Disposal”
 Basic concept about securing way of financing, foundation of implementer, site selection process,public communication and planning URL projects etc. was summarized.

1999: JNC (now JAEA) “H12 report”
 JNC summarized geological disposal concept, geological environment features, engineering
technology and safety assessment etc. to evaluate technical feasibility of HLW geological disposal
in Japan.
*JNC: Japan Nuclear Cycle Development Institute,
H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan

2000: “Final Disposal Act” Promulgated NUMO was established as an implementer of the HLW geological disposal in Japan

21

1

Overview of the “Final Disposal Act”
 The followings were included in the “Final Disposal Act” (enacted in 2000) in order tosystematically and reliably conduct the final disposal (disposal in a stratum more than 300 m
deep underground) of high-level radioactive waste generated after the reprocessing ofspent fuel used in nuclear power generation.
• Minister of Economy, Trade and Industry stated a basic policy for final disposal of specified radioactive waste(Cabinet decision).
• Nuclear Waste Management Organization (NUMO) established as an implementer for geological disposal.
• Three-stage selection investigation process was set for selection of repository sites, etc.

◆Selection process outlined in the “Final Disposal Act”
Literature survey

Preliminary
investigation

Assessment withliterature surveys

Assessment withborehole surveys, etc.

Past history ofearthquakes, locationof active faults andvolcanos, etc.

Characteristics and stateof deep rocks andgroundwater

Detail investigation

Selection of geologicaldisposal facility

Detailed assessment withsurveys of undergroundfacilities, etc.

About 20 years※The government will hear the opinions of local municipalities in each stage of the investigation process(not proceed to the next stage if local municipalities oppose).

2

Implementation System for Geological Disposal
Fundamental responsibility
for geological disposal

Governmentalorganization

Supervision

METI

NUMO
Selection of repository
sites/implementation ofgeological disposal, etc.

Regulator

NRA*NRA: Nuclear Regulation Authority

Implementer

Nuclear power

Payment reactor installationof funds
companies
Electric powercompanies, etc.
Cooperation in terms of humanresources and technology

Regulation
Sharing R&D results

National Research andDevelopment Agency

JAEA
Horonobe/ Mizunami URLs andTokai Laboratory

3

22

Timeline of Review of Initiatives for Geological Disposal
 In December 2002, NUMO started open solicitation for all municipalities toapply for an investigation.
 In January 2007, Toyo Town in Kochi Prefecture submitted a formal application.
Following this submission, a controversy erupted around the pros and cons of
their participation in the survey, which divided the town politically.
The application was withdrawn after the mayoral election in April of the same year.
 To date, no formal investigation has been carried out.

Review of initiatives

Est. of Ministerial Meeting on Final Disposal (Dec 2013)
Discussion on direction of review

Strategic Energy Plan (Apr. 2014)
Revision of the basic policy based on the Final Disposal Act(May 22, 2015)
4

Points on the Revision of the Basic Policy Based on the Final Disposal Act (May 22, 2015)(1) Responsibility of current generations and potential for future generations to choose The responsibility of the current generation, which has generated the waste, is to successfully promotemeasures for geological disposal, in order not to pass the burdens on to future generations.
 Specifically, current generations must ensure the potential for reversibility and retrievability (R&R) to offer
future generations the potential to change to more ideal disposal methods. R&D for the alternative disposal
options will be conducted to ensure there is a broad range of options.

(2) Encourage national public understanding and regional understanding It is important to encourage the public to share respect and appreciation for areas that contribute to final
disposal projects and to recognize the need to return profits to those areas.
 Seamless provision of information from the national government to local governments throughout Japanthrough numerous careful dialogues.

(3) Activities spearheaded by the national government Proposals made to related local governments to obtain their understanding and cooperation in nationwide
scientific screening that is considered to be more scientifically suitable and the status of activities to promote
understanding of geological disposal including geological feature in Japan.

(4) Support for the region to contribute to the project Support of activities and the establishment of a Forum for Dialogue in which a diverse set of residents can takepart towards proactively building consensus in the area.
 Consider enacting comprehensive support measures to contribute to sustainable development of the regions.

(5) Improvements of the organizational structures
 Strengthen the organizational system of the project entity NUMO (Nuclear Waste Management Organization ofJapan).
 Clarification of the involvement of the Japan Atomic Energy Commission and implementation of ongoingassessments in order to ensure reliability. The Nuclear Regulation Authority, Japan will successively present
safety considerations related to safety regulations according to the progress of NUMO’s surveys.
 Promote the expansion of storage capacity for spent fuel.

23

5

Discussion of requirement and criteria for nationwide screening
Nationwide Map –Making Process 2014: Geological Disposal Technology WG started
 2015: Interim summary was published 2016: OECD/NEA International Peer Review(https://www.oecd-nea.org/rwm/pubs/2016/7331-japan-peer-review-gdrw.pdf)

 2016: JAEC review
 2017: Final Report was published : Requirement and criteria were
set up for making nationwide map July 2017: Publication of Nationwide Map of Scientific features(http://www.enecho.meti.go.jp/category/electricity_and_gas/nuclear/rw/kagakutekitokuseimap/)
6“Nationwide Map of Scientific Features”

724

Status of the “Nationwide Map of Scientific Features for Geological Disposal”

Classification of areas

Criteria to identify unfavorable features••••

Assumed to be
unfavorable

If any one isapplicable

Vicinity of volcanoes
Vicinity of active faults
Significant uplift/erosion
High geothermal gradientetc.

from the viewpoint of long-term
stability of the deep geologicalenvironment
from the viewpoint of the risk
of future inadvertent
human intrusion

If applicable• Existence of mineral resources

Assumed to be favorable

If none is applicable

Criteria to identify preferable features
Relatively short distance from coastline(including sub-seabed and islands)

Assumed to be preferable also
from the viewpointof safe waste transportation

If applicable

8

Requirement and Criteria
Requirement

Geological
Environment
Features andLong-Term
Stability

Construction
andOperation of
Facilities

Criteria

Volcanic/igneous
activity• Magma intrusion affecting physicalisolation

× Vicinity of volcanoes: Within an area of 15km fromthe center of individual Quaternary volcanoes (or thecaldera rim if this is greater)

Fault movement• Fault movement affecting containment

× Vicinity of active faults: Within the fracture zone
around an active fault, the width of which is about
1/100 of the fault length

Uplift/erosion• Uplift/erosion affecting physicalisolation

× Significant uplift/erosion: Net erosion greater than
300m/100,000 years; in coastal areas, accounting for
sea-level change, uplift rate greater than
90m/100,000 years

Geothermal activity• Geothermal activity affecting
containment

× High geothermal gradient: Geothermal gradientgreater than about 15oC/100m

Volcanic thermal fluids
and deep-seated fluids• Intrusion of exotic groundwater
affecting containment

× Presence of hydrothermal water or other deep-seatedgroundwater: Groundwater with pH less than 4.8

Mineral resources• Future inadvertent human intrusion

× Existence of mineral resources: Known oil, gas andcoal fields, metal ores

Volcanic eruption• Volcanic eruption affecting safe
construction

× Susceptibility to distant impacts from volcanic
eruptions: Traces of Holocene pyroclastic flows or
associated pyroclastic rocks

Unconsolidated• Geotechnical instability affecting safe
construction
geological formation

Transportation Transportation

× Location in unconsolidated geological formations:Geological formations younger than Middle
Pleistocene as cover to a depth of greater than 300m

• Safe waste transportation in terms of Relatively short distance from coastline (including subradiation exposure and nuclear security
seabed and islands): Within about 20km form coastline✓ : preferable

25

× : unfavorable

9

Step-by-step approach toward site selection and geological disposal The publication of the map is the first step on a long way toward completion ofgeological disposal.
 With the aim that multiple municipalities will undertake site investigations, we willcontinue to hold public dialogues to ensure a deeper public understanding of the issue.
Deepen national andregional
comprehension

Site investigations
Conducted by NUMO based onregional comprehension

Listening to public
voices• Prioritized activities
focusing on “green (coastalarea)”

Literature survey
With the aim thatmultiplemunicipalities will
accept site
investigations
Preliminary investigation(borehole survey, etc.）• Supporting regional
deliberation

Selection of
final disposal site

Publication of
Nationwide Map of ScientificFeaturesNationwide explanatory
meetings using the map• Promotion of R&D
Detailed investigation• International cooperation& contribution(construction & studies inunderground facilities)

10

Concluding remarks<Purpose of the Map>
 The Nationwide Map of Scientific features for Geological Disposal was
published in order to deepen the public understanding, not for the site
selection directly.
<First step by map>
 Publication of the map is the first step in a long way and a newchallenge toward geological disposal completion. Government and
NUMO will implement and hope multiple regions’ acceptance of site
investigation.
<International cooperation>
 Sharing learned knowledge with countries would be helpful for
implement of geological disposal. Japan can share the reaction of themap publication.
11

26

1

How to Integrate Various Research Fields
for Reasonable and Reliable Radwaste
Management

Yaohiro Inagaki
Kyushu University
Chair of the Division of Nuclear Fuel Cycle & Environment,
AESJ

Waste management in the present nuclear fuel cycle
Mining &milling U

Fuelfabrication
Nuclear fuelProducts(U, Pu)2

Key process & operationin Up-stream
Key process & operationin Down-stream
Stream of materials

Reprocessing
Spent fuel
HLW

Vitrification

Reactor(LWR)

Correlation & interaction

Features of cycle

ILW

LLW

- Highly integrated system

Volume red.

Volume red.

immobilization

immobilization- Correlation between
processes & operations- Waste management is
one of key processes

Storage

Storage

Storage

GeologicaldisposalIntermid. landdisposal

Shallow landdisposal27

Waste management for progress in nuclear fuel cycle

3

Progress in nuclear fuel cycle
High burn-up LWR → MOX (U+Pu) -LWR → FBR → ?
Changes in waste properties
Nuclide composition, Radioactivity, Heat generation

For progress in total performance of cycle- Consistency between many processes & operations- Evaluation of the effects on waste management- Evaluation of total cycle performance from up-stream
to down-stream with developing evaluation methods
An example of the effects on waste management:

4

Heat generation rate of SNF, kW/THM

Heat generation rate of spent nuclear fuel (SNF)
Cooling period of SNF after
14.0discharge: 4y

14.0
12.0

UO2-LWR

12.0

10.0

4.0
2.0
0

Np: 0.01%
Am: 6.79%

10.0

8.0
6.0

MOX-LWR8.0
Np: 0.03%
Am: 1.52%
Cm: 4.51%
FP+U+Pu:
93.97%

Np: 0.03%
Am: 1.41%
Cm: 7.99%

6.0

FP+U+Pu:
90.60%

2.0

70

Burn-up, GWd/THM

Cm: 32.89%

Cm: 28.47%

4.0

0
45

Np: 0.01%
Am: 7.55%

FP+U+Pu:
63.97%

45

FP+U+Pu:
60.31%

70

Burn-up, GWd/THM

Heat generation rate of SNF after 4 year cooling calculated forhigh burn-up UO2–LWR and MOX-LWR (Inagaki, et al., 2009).
- For MOX-LWR, the heat generation rate is larger than that of UO2-LWR
by a factor of 2 by an increase in contribution of MA(Am, Cm).
- Contribution of MA increases with burn-up.
28

An example of the effects on waste management:

5

Number of glass units per GWd

Number of HLW glass units generated per GWd
0.07

Restrictions for present vitrification
Heat generation rate < 2.3kW/glass unit
Content of MoO3 < 1.50 wt%
Content of Noble metals < 1.25 wt%

0.06

MOX-LWR

0.05
0.04
0.03

Dominant restriction factor
UO2-LWR: Heat generation rate
MOX-LWR : Heat generation rate

UO2-LWR
0.02
0.01
0

Cooling period of SNF after discharge: 4y

20

30

40

50

60

70

80

Burnup, GWd/THM

Number of glass units generated per GWd as a function of burn-up for
UO2-LWR and MOX-LWR assuming present vitrification. (Inagaki, 2009)- Dominant restriction factor is “heat generation rate” for both cases.
- Number of glass units for MOX-LWR is larger than that of UO2-LWR.
- Number of glass units per GWd is constant with burn-up for both cases.

An example of the effects on waste management:

6

Number of glass units per GWd

Reduction of HLW glass units for MOX-LWR
0.07

MOX-LWR

MOX-LWR
0.06

Before MA separation

0.05

30%
Restriction by noble metals

0.04

Separation of MA from SNF

After MA separation

0.03
0.02

Decrease in heat generation rate

Restriction by MoO3

0.01
0

Increase in number of HLW glass
units per GWd by increase in MA

Cooling period of SNF: 4y
20

30

40

50

60

70

80

Decrease in number of HLW
glass units per GWd

Burnup, GWd/THM

Effects of MA separation on number of glass units generated per GWd(Inagaki, 2009)- MA separation from SNF is effective to reduce number of glass units.
- In this case, content of noble metals is dominant restriction factor.
- Separation of noble metals can provide much reduction of glass units.
29Basic concept of HLW geological disposal system

7

Bentonite

HLW glass

Host rock

Overpack(Carbon steel)

Buffer material(Bentonite, sand)Engineered barrier

Geological formation(Deeper than 300m)
Natural barrier

Basic concept of geological disposal system in Japan (JNC, 2000)- Multi-barrier system consists of engineered and natural barriers.
- The system is expected to isolate waste for more than 100,000 years.

Barrier functions of geological disposal system

HLW glassBuffer material

Overpack

Low permeability
against water intrusion

Physical containmentof HLW glass for morethan 1,000 y

Buffer material

Chemical durability
against waste release

Low solute diffusivity
to retard waste
migration

Functions of each barrier
Integration of functions of all barriers
Geological formation

Total performance ofgeological disposal system

Retardation, dilution,
Dispersion of waste
30

8

9

Various research fields for performance assessment

HLW glass

Overpack

Buffer material

Materials science,
Materials. eng., etc.

Geochemistry,
Mineralogy, etc.

Buffer material

Materials science,
Geochemistry, etc.

Geochemistry,
Mineralogy, etc.

Performance of each barrier
Integration of various research fields
Geological formation

Total performance ofgeological disposal system

Geology, Mineralogy, Civil eng. etc.

Correlations & interactions between barriers

Correlation &interaction

Correlation &interaction

Buffer material

Overpack corrosion affects
buffer permeability,
glass corrosion

Correlation &interaction

HLW glass

Overpack

10

Glass corrosion affects
overpack corrosion,
buffer properties

Buffer material
Solute sorption affects
glass corrosion

Performance of each barrier
Correlation &interaction

Correlations & interactions
between barriers

Geological conditions
affect buffer properties

Total performance ofgeological disposal system

Geological formation
31

Method for reliable performance assessment

11

Structure
Performance
Perfom. of
Buffer material

Engineering
Perform. ofOverpack

Perform. ofGeo. formation

Scientific basis

Performance ofGeological disposal

Interaction
Barrier 1
Barrier 2

Perform. of
HLW glass

Barrier N

Engineering

Scientific basis

Total
performance

Engineering

Scientific basis

Scientific basis

Scientific basis

Integration of research fields for waste management
Engineering

Science- Radiochemistry- Geochemistry- Materials science- Geology- Mineralogy- Natural analogue- Radiology
etc.

12

Performanceassessment

Regulation- Chemical eng.
- Materials eng.
- Civil eng.
- Geotechnology- Environment. eng.
- Radiation protection- Simulation/cal. tech.
etc.

Public acceptance- Safety - Reliability - Responsibility - Cost
- Ethics - Equity- Transparency etc.

“How to integrate various research fields”
The most important subject to be considered for implementation ofreasonable & reliable waste management32

13

Summary- Nuclear fuel cycle is highly integrated system consisting ofmany processes with correlations & interactions between them.
- Waste management is one of key processes for progress intotal cycle performance.
- Waste management is also highly integrated system consistingof several barriers & functions with correlations & interactions.
- Performance assessment should be based on integration ofvarious research fields from scientific basis to engineering inconsideration of correlations & interactions.
- “How to integrate various research fields including social
issues” is one of the most important subjects to be developedfor implementation of reasonable & reliable waste management.

14

Thank you for your attention&
I hope you will enjoy
active & successful discussions
at EAFORM 201733

Current Situation of Radioactive Waste
Management in Taiwan
Ching-Tsuen Huang, Ph. D.
Chairman, Academic Committee onRadioactive Waste Management,
Chung Hua Nuclear Society (CHNS)
Professor, Inst. of Nuclear Eng. & Sci.,
National Tsing Hua University

1

Outline Introduction
 Low Level Radwastes Management Spent Nuclear Fuels Management New Policy & Challenges
 Summary234

Introduction
 Peaceful Use of Nuclear Energy in Taiwan

3

Introduction
 Peaceful Use of Nuclear Energy in Taiwan
 Supplies safe, clean, inexpensive electricity
60

Year

52.41

50

47.87

48.54

41.03

32.4

30

2014

2015

2016

38.32

35.24

32.41
37.79
34.59

28.79

30.36

24.77
25.32
26.35
23.64

29.07

22.91

21.57

20

20.93

21.5

13.49
19.49 19.61
19.30
18.37
18.61
20.65
19.3
19.02 18.78
16.04

20.25

10

0
1982
1983
1984
1985
1986
1987
1988
1989
1990
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016

Nuclear share %

30.61

2013

Capacity factor (%) 86.06 88.94 90.55 77.99 67.60

43.77

40

2012

4

35

Introduction
 Peaceful Use of Nuclear Energy in Taiwan
 Supplies safe, clean, inexpensive electricity
average cost

nuclear3
2
NT$/KWh
1
0
1989

1992

1995

1998

2001

2004

2007

2011

2013 2016

5

Introduction Administration System
 Legal Framework of RWs Management
By
Legislature

By AEC

Treatment

By FCMARadioactive Wastes andMaterials Administration Act

Act on Sites for Establishmentof Low Level Radwaste Final
Disposal Facility

15 Regulations

Forbidden-area
Regulations on LLWs
Final Disposal Site

StorageTransportation

D&D

Disposal

Regulatory Guides
636

Introduction Administration System Organization of RWs Management
Executive Yuan
Atomic Energy
Council
FCMA

INER

Envi. Protection
Administration

Regulatory
Body

Technical
Support

Ministry Of
Economic Affairs

EIA

Utilities otherthan NPPs

Supervision

Taipower
Operation

NPPs
7

Introduction
 Congenital Deficiency of RWs
Management Small territory and dense population
lead to serious NIMBY syndrome Leading to the rapid growth of anti-nuclear
force after the 311 nuclear accident
 Causing the difficulty of radioactive waste
final disposal

837

LLWs Management
 Generation of LLWs
 About 90% from NPP, 10% from others All LLWs are safe in storage
In NPPs (drum)
Solidified
Waste
Resin
Compressible
Miscelaneous
Subtotal

In Orchid Island(drum)In INER (drum)

38,282 Solidified Waste 3,904
17,302
13,89536,651
118,320

Solidified
Resin
165
100,277
Waste
Metal
10,430Spent source 11,458
26,282100,277

Total

244,879
9

LLWs Management
 Minimization of LLWs
 Source reduction
 Application of High Efficiency
Solidification Technologies

Total Annual Generation of Solidified
LLW in 3 NPPs
D
r
u
m

500
450

449
388

400

356

Drums ／ Year

350
300
250
200

176

150
100
77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16

50

34

24

99

2000 2001 2002 2003 2004 2005 2006 2007 2008 2009

18

17

27

18

19

30

17

12

22

0
95

96

97

98

Calender Year

10

38

LLWs Management
 Storage of LLWs
 Orchid Island storage site
 100,277drums in 23 trenches

17-35cm water-tightseal plate

35cm waterproof concrete

40cm water-proof
concrete

11

LLWs Management
 Storage of LLWs
 NPPs & INER warehouse storage
 118,320 drums in 3 NPPs
 26,282 drums in INER

1239

SNFs Management
 Storage Capacity and Inventory of
Spent Fuel
 Full loading of fuel pools
Status of Spent Fuel Storage (July 2017)
Date OfCapacity
License Expires (Assembly)

Unit
ChinshanKuosheng
Maanshan

#1
#2
#1
#2
#1
#2

Dec. 5, 2018
July 15, 2019
Dec. 27, 2021
Mar. 14, 2023
July 26, 2024
May 17, 2025

Total

Storage
Full
Inventory discharge( Assembly) per cycle

3,083
3,083
4,838
4,398
2,160
2,160

3,074
3,076
4,548
4,388
1,452
1,468

19,72218,006

~110
~110
~180
~180
~70
~70
13

SNFs Management On-Site Dry Storage
 Chinshan NPP
 AEC granted the Construction
License in Dec, 2008.
 Completed pre-operational tests inJan., 2013
 Hot test approved in Sept., 2013by AEC 2 casks of spent fuelPhotograph Current status Local government refused
to issue 〈Certificate ofCompletion of Soil andWater Conservation〉

14

40

SNFs Management On-Site Dry Storage
 Kuosheng NPP
 MAGNASTOR concretecasks will be used
 Licensing applicationin
March, 2012
 Construction License
granted by AEC in
Aug., 2015
 Current status
 Waiting for the local
government to approvethe 〈Runoff Waste
Water Pollution
Reduction Plan〉
15

SNFs Management Small-scale abroad reprocessing plan in2015 1,200 bundles of used BWR fuel to be
reprocessed abroad To relieve fuel pool volume for 3-year
operation of Chinshan and Kuosheng NPPs

 As a plan B of interim storage project The budget was rejected by the legislature
16

41

SNFs Management
 Promotion of Public Acceptance Public Hearings on SNF Dry Storage
Facility Public Observation on SNF Dry
Storage Facility
 Public Participation in Environmental
Radiation Monitoring17

SNFs Management
 Promotion of Public Acceptance Summary of Public Hearing & Observation
 The pre-operation of SNF dry storage facility (ISFSI) must include
a mockup simulation of the retrieval operation of SNFs
 The operator must make sure a facility available for retrieving
SNFs in the NPP’s decommissioning plan.
 The maintenance and surveillance plan of the ISFSI shall include
the monitoring of SCC of the storage facility.
 The operator must re-evaluate the seismic design of the ISFSI.
 The storage facility should be decommissioned when the 40-years
license expires

Public Participation

Public Hearing

18

42

18

SNFs Management
 Promotion of Public
Acceptance

Time:2017-09-21 16:55

 Establishment of
Real-time
Environmental
Radiation
Monitoring
System 28 stations, 2008 46 stations, 201719

SNFs Management
 Promotion of Public Acceptance Public Participation in Environmental Monitoring in Orchid
Island Observational check by third party (2008 to 2017)
 Invite scholars, experts, NGO and congress man toparticipate in environmental monitoring and sampling Parallel Environmental monitoring and sampling (Nov.,
2012) All results were far below regulatory concern and at thelevel of environmental radiation background

Introduction before sampling

Detection record

Environmental detection

Water sampling

Soil & grass sampling

20

43

SNFs Management
 Promotion of Public Acceptance Environmental Monitoring in Orchid Island
 Monitoring data showed that Orchid Island
has much lower environmental radiation thanother area in Taiwan.

21

New Policy & Challenges
 「Nuclear-free Homeland」Policy
License Expires

Unit
#1*

Dec. 5, 2018#2*

July 15, 2019#1

Dec. 27, 2021

#2*Mar. 14, 2023

#1
#2

July 26, 2024
May 17, 2025

Restart denied after outage in
Dec., 2014, due to fuel assembly
handle loose during fuel transfer
Shutdown in June, 2017
due to full load of fuel pool
Fuel pool re-racking in June, 2017
allowing 2 more refueling cycle
Shutdown in May, 2016
Due to lightning arrester failure
OperatingOperating

ChinshanKuosheng

MaanshanStatus22

44

New Policy & Challenges
 「Nuclear-free Homeland」Policy New 《Electricity Act》- January 26, 2017 Article 95：Nuclear power generation facilities
shall be shut down before 2025

 Tough target of renewable energy share 20% share of electricity in 2025, totally 27.4GW
 Solar 20GW
 Wind 4.2GW

 Advices from international independent scientists,
conservationists, and energy experts
 Michael Shellenberger：Taiwan is uniquely
unsuited for renewables like solar and wind23

New Policy & Challenges
 「Nuclear-free Homeland」Policy Dramatic warning of consequences ofnuclear phase-out Electricity blackout on Aug. 15th, 2017
 Lost 4,384 MWe due to gas supply
interrupted» Half of Taiwanese households lost power,
as did a large number of companies.

 Too low POR (Percent Operating Reserve)2445

New Policy & Challenges
 「Nuclear-free Homeland」Policy Electricity blackout – A dramatic warning
Taipei 101
Taipei(night101
view)(blackout view)

25

New Policy & Challenges
 「Nuclear-free Homeland」Policy
 Be beset with difficulties
 Local government boycotts the on-site
SNF interim storage facilities
 Obstruct the decommissioning of NPPs

 Local government refuses to conduct thereferendum for LLW final repository siting LLWs in Orchid Island storage site can not
be removed out as the governing party
promised26

46

New Policy & Challenges
 「Nuclear-Free Homeland 」Policy
 Promotion Project Purpose Strategic planning Promote public acceptance Reforming of RWM laws and regulations

 Advisory Committee
 Organized bygovernment
representatives
and 14 civil members
27

New Policy & Challenges
 The「Nuclear-free Homeland」Policy
 Contingency Plan of RWM
 Centralized storage facility
 An off-site in-door storage facility
 To store SNFs and LLWs for 40 years» $2.635bn investment as estimated» Needs an area at least 26 hectares

28

47

New Policy & Challenges
 The「Nuclear-free Homeland」
Policy Contingency Plan of RWM
 Challenge Anti-nuclear attitude of local government
has no change Facility site selection is a mission almost
impossible» More difficult than the site selection of
LLWs final repository
29

New Policy & Challenges
 Changes Anticipated Legal Framework Reform
Laws and Competent Authority
Current Status
Atomic Energy Act
Nuclear Materials andRadioactive Waste
Management Act
Act on Sites for
Establishment of Low
Level Radwaste Final
Disposal Facility
-

AEC

AEC
AEC

After Reform
Atomic Energy Act (amend/abolish？)
Nuclear Materials and Radioactive
Wastes Regulatory Act
Radioactive Wastes Management Act
Act on Sites for Establishment of Low
Level Radwaste Final Disposal Facility
Act on Sites for Establishment of Spent
Nuclear Fuel Final Disposal Facility
Act on the Establishment of
Independent Administrative Inst. of
Radwaste Management Center (RWMC)

AEC ?
AEC
MOEA
MOEA
MOEA
MOEA
30

48

New Policy & Challenges
 Changes Anticipated Organization & Duty
 AEC：Authority of Nuclear Safety
 FCMA：Safety Regulation of Nuclear
Materials and RWs

 MOEA：Authority of RWs Administration
 Taipower：RWs Treatment and storage
 Nuclear Power Backend Fund Management
Committee：Backend fund management
 Radioactive waste Management Center：RWs
Final Disposal
31

Summary The routine operation of RWM in Taiwan is as
good and safe as it was, and made the bestrecord of LLW minimization.
 People's fear of nuclear energy burst out after
Fukushima nuclear accident, causing local
governments to oppose all nuclear facilities.
 The legislature has amended the “Electricity
Act” claiming to shut down all NPPs before2025.
 Local government's attitude towards nuclear
facilities has not changed. Decommissioning of
nuclear power units is beset with difficulties.
32

49

The 6th East Asia Forum on Radwaste Management

November 27, 2017

Joowan PARK

Contents
Ⅰ

Overview

Ⅱ

National Plan for HLW Management

III

Future Plan

Ⅳ
Ⅰ

Summary

50

Ⅰ

Overview

Nuclear Fuel Cycle Facilities in Korea

Hanul 1,2,3,4,5&6

In Operation

24 Units(22,529MW)

Under Const.

5 Units(6,000MW)

ShinHanul 1&2 (Under Const)

Seoul

UlchinKORAD HQ

Daejeon
DaeDuk Science Town
NF
: KAERI, KINS, KEPCO-NF,
KORAD R&D Institute,
KHNP-CRI

Gyeongju
Wolsong
Kori

Hanbit 1,2,3,4,5&6

Dok-doLILW Repository
Wolsong 1,2,3&4
ShinWolsong 1&2

Yeonggwang

Kori 1(shutdown),2,3&4
ShinKori 1,2&3
ShinKori 4 ,5&6
(Under Const)

451

National Framework of RWM

President

Implementer & Promoter

Regulator
Prime Minister

NSSC

AEC

(Nuclear Safety andSecurity Commission)

Atomic Energy Commission

Regulatory Authority

MOTIE(Ministry of Trade,Industry and Energy )MSIT(Ministry of Scienceand ICT)

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

Korea Radioactive
Waste Agency(KORAD)

Korea Atomic
Energy ResearchInstitute (KAERI)• NPP Operator• RWM Business
Implementer• R&D in Nuclear
Technologies

KINS(Korea Institute of
Nuclear Safety)

KINAC(Korea Institute of
Nuclear Nonproliferation& Control)

Regulatory Expert
Organizations

5

Radioactive Waste Management Act□ Independent organization for radioactive waste management (article 10)
“Korea Radioactive Waste Agency(KORAD) is designated as the dedicatedorganization for RWM business”□ Scope of radioactive waste management business (article 9)
 Transportation, storage, treatment and disposal of radioactive waste Siting, construction, operation and post-closure management of radioactive
waste management facilities
 Collection, investigation, and analysis of data on radioactive waste Public relations
 Other activities such as R&D, international cooperation and training□ Radioactive waste management fund (article 28-33)
 Establishing the Government Fund to improve the transparency andstability of financial resources for RWM
 Administration of RWM fund is committed to the KORAD
652

Current SNF Management Practice in NPPs
□ PWR Spent Nuclear Fuel
- Stored in on-site SNF pool (Wet storage)- Re-racking, Trans-shipment to neighboring NPP□ PHWR(CANDU) Spent Nuclear Fuel
- Stored in both on-site SNF pool and dry storage facilities

On-site SNF pool
(Wet storage)

Wolsong On-site Dry Storage Facilities
7

SNF Generation in NPP sites & Saturation of storage• SNF generated 15,401 tons as of end of Jun. 201717,574 assemblies(7,256 tons) from PWR, 430,576 bundles(8,145 tons) from PHWR

• Expected : 38,941 tons (presumed 30 reactors to be operated)
63,146 assemblies(26,521 tons)from PWR, 657,148 bundles(12,420 tons) from PHWR

SNF Generation&
Future Estimate

499,632 86.2%
430,576(Jun. 2017, unit : assembly or bundle)

66.9%

9,017

6,034

capacity

8,115 74.2%
6,024

currently stored

7,066 74.5%
5,263
1,046

24.2%
253

Wolsong

Hanbit

Kori

Hanul

Shin wolsong

Saturationofon-site storage

Wolsong
Hanul

Hanbit/Kori

Shinwolsong853

ⅡNational Plan for HLW Management

Efforts made in HLW Management Policy-making (1/3)
1Dec. 2004 : The 253rd Atomic Energy Commission(AEC)

decided that the project of HLW management should be carried out
separately from that of LILW management• Consent-based national policy for SNF management should be madewith consideration of related R&D, international trend, etc.
•2•

Dec. 2009 : Legal ground for the public engagement

established the public engagement process and function of the public
engagement commission in the Article 6.2 of RWMA
3

Nov. 2011~Aug. 2012 : SNF Policy Forum•consisted of 23 forum members (including the chair)•shared opinions and discussed on SNF management options and the
public engagement process among experts and stakeholders•

operated 2 working groups for review of SNF management options andreview of public engagement process
1054Efforts made in HLW Management Policy-making (2/3)
4Oct. 2013~Jun. 2015 : Public Engagement Commission on SNF(PECOS)
collected opinions of the public, local residents of NPP areas andstakeholders through town hall meetings, deliberative poll, etc.

•

(about 370,000 people participated)submitted the recommendation report to the government (Jun. 2015)•

Town hall meeting5•

Discussion at the National Assembly

Deliberative Poll

Jul. 2015~Apr. 2016 : TFT for national plan for HLW management

was organized consisting of 50 people(including experts of academic field, relatedorganizations, and government agencies and lawyers) for have in-depth review of thedraft national plan

11

Efforts made in HLW Management Policy-making (3/3)•

6Jun. 2016 : Public Hearing for National Plan on HLW management

7

Jul. 2016 : Public Announcement of National Plan for HLW management

6th Atomic Energy Commission deliberated and decided the National
Plan in the Article 3 of Nuclear Promotion Act(NPA)8•The Legislative Bill submitted to the National Assembly
9•Nov. 2016 : Legislative Bill [ The Site Selection Process and Host
Community Support for HLW Management Facilities ]Jul. 2017 : Announcement for 5-year Government Operation Plan

Review of National Plan on HLW Management by the 2nd PECOS
12

55

Outline of National Plan on HLW Management• Article 6 of the Radioactive Waste Management Act(RWMA)

Legal basis• Presenting the principles and directions for radioactive waste
management(RWM) as the highest level of plan
• KORAD should establish the implementation plan based on thenational plan annually according to the Article 7 of RWMA

Process• The Atomic Energy Commission(AEC) approves the national plan for
RWM prepared by MOTIE, according to the Nuclear Energy Promotion
Act

Contents• ①Principles of HLW management, ②Current situation & estimate of
HLW generation, ③Facilities development plan including siting process④Investment plan, ⑤Plan for social trust & confidence, ⑥R&D plan13

Principles of HLW Management
1

Governmental responsibility• As HLW needs to be safely managed for long-term period, HLW should be managed undercontrol of government, conforming to domestic and overseas safety standards.

2

Top priority on safety• The public health and the environment should be protected by eco/environmental friendly

management of HLW.

3

Trust & confidence• All the relevant information should be open to the public.
• HLW management project should be under the public consensus.

4

Due burden on present generation• Present generation should be responsible for HLW• ‘Polluter Pays’ principle should be applied

5

Effectiveness of HLW management• Technologies on transportation, storage, disposal and reduction of toxicity & volume

should be developed for effective management of HLW
1456

Roadmap for HLW Management Facilities
Start ofsitingprocessSiteselected+12 years

NPP

Site selection for interim
storage and repository+7 years
Construction of interim storage
Management of SNF generated
from NPPs for certain period
before disposal+24 years+10 years
Construction of generic URL- Site to be selected separatelyfrom the management facilities- R&D on HLW disposal system

Construction of site-specific URL and implementationof RD&D(+14 years)
RD&D on deep geology of candidate site
for construction of repository

Construction of Repository(+10 years)
Permanent isolation of HLW by means of
natural barrier and engineered barrier(00m ~1,000m below sea level)

15

Strategies of HLW management

Siting &Construction• The site for integrated facilities, hosting interim storage,
URL, and repository, will be selected after 12 year-process.
• Interim storage, URL and repository will be constructed in phase.

R&D• On-site storage will be inevitably needed before site selection.
• Key technologies for safe management of SNF will be
developed in timely manner• Technical cooperation with domestic & overseasorganizations will be madeCommunication• Information regarding operation of facilities will be open to
the public and local people.
• Various communication programs with local people andstakeholders will be continued.
16

57

Contents of Legislative Bill
□ Siting process
Exclusion of
unsuitable
sites

Volunteering/
Decision-makingof the local people

Unsuitable sites tohost HLW
management facilities
will be excluded.

Local governments
will decide whether
to apply for site
investigation aftercollecting theresidential opinions

3 years

Preliminarysite
investigation

Site characteristics &suitability will be
investigated.

5 years

In-depth site
investigation

Site selection

Detailed
investigation will be
carried out into thesites turned out‘suitable’ after
preliminary
investigation

Site will be selected
after the in-depth
investigation.

4 years□ The Third party (the Committee on Site Selection) for
transparent siting process to be organized under the MOTIE□ The Committee on Host Community Support to beorganized under the Prime Minister

17

III

Future Plan

58

The 2nd Public Engagement
Why?
- In accordance with change of governmental roadmap on
long-term energy transition- Improvement of social acceptanceHow?
- Organization of the 2nd Public Engagement Commission on
SNF(PECOS)

What?
- Focused on the controversial issues included in the national
basic plan
- Detailed agendas to be decided by the 2nd PECOS
19

The 2nd Public Engagement

Step 1

Preparation
Preparation for the
public engagement tobe madeStep 2

Step 3

Implementation

Amendment of
national plan

The 2nd PECOS to beorganized

In-depth review of therecommendation of the
2nd PECOS

2059

Ⅳ Summary

Summary□ Safety is the most important fundamental objective of the national
HLW management program in Korea.
□ RWM, including HLW management, could be more effectively andtransparently proceeded by establishment of the KORAD in 2009,
separated from the RW generators.
□ Based on the recommendations by Public Engagement Commission
on Spent Nuclear Fuel(PECOS), a national basic plan for HLW
management was set up in July 2016.
□ New government is to make in-depth review of the National Plan on
HLW Management by the 2nd PECOS to raise the public acceptance□ KORAD is continuing the efforts, including RD&D activities, toward
successful implementation of the national HLW management
program.
22

60

61

Photos placed in horizontal position
with even amount of white space
between photos and header

Recertification of WIPP: From Submittal of the
CRA-2014 to an EPA Recertification Decision
EAFORM2017: 6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017

Todd R. Zeitler

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly
owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.
SAND2017-12125C

Outline WIPP Background and Performance Assessment- Regulatory requirements- Performance Assessment- Release Mechanisms and Compliance Metric 2014 WIPP Compliance Recertification Application- Submittal in March, 2014
- Completion of three EPA-mandated sensitivity studies- Recertification in July, 2017

 Summary

262

Waste Isolation Pilot Plant Location3

About The WIPPThe WIPP is a permanent disposal facility for transuranic waste• The nation's only licensed deep
geologic repository for nuclear waste•

Operated by U.S. Department ofEnergy (DOE)•

Long-term performance regulated
by U.S. Environmental
Protection Agency (EPA)•

Waste is emplaced in a salt
formation deep underground•Long-term regulatory complianceis demonstrated via Performance
Assessment (PA) undertaken by
Sandia National Laboratories (SNL)
Carlsbad4

63

Long-Term Regulatory Requirements
 Regulatory requirements guide the
WIPP PA framework.
 The WIPP must be designed to provide
reasonable expectation that cumulative
releases of radionuclides to the accessible
environment for 10,000 years after closure
from all significant processes and events
shall be less than specified release limits

5

Regulatory Requirements
 Reasonable expectation:
regulations acknowledge
substantial uncertainties
 10,000 years: PA must predict
behavior for entire regulatory
time period
 Significant processes andevents: PA must include all ofthese, including the possibility ofhuman intrusion
664

From 40 CFR 191

Qi  1106 curies R   Li C
R = Normalized release in “EPA units”
Qi = 10,000-year cumulative release (incuries) of radionuclide i
Li = Release Limit for radionuclide i
C = the total transuranic inventory (incuries of a emitters w/halflives > 20
years)

WIPP Performance Assessment
 PA calculations cover 24 peer-reviewed conceptual models
 PA codes include 10 principal codes and many utility codes

7

Release Pathways in WIPP PA
Direct Releases fromDrilling Events
Ground Surface

Ground Surface
CulebraArea

Marker Bed

Pressurized Brine
Source865

Land
Withdrawal
Boundary

Waste

Shaft

BoreholeLand
Withdrawal
Boundary

Marker Bed

Total Release CCDF is the Measure of ComplianceTotal releases from the repository are compared to regulatory
release limits to determine compliance via a Complementary
Cumulative Distribution Function (CCDF).
CRA-2014
PA

Less than 1 chance in 10
of exceeding 1 EPA unit

Less than 1 chance in1000 of exceeding
10 EPA units

9

WIPP Recertification
Federal regulations require that the WIPP be recertified every
five years following the first waste shipment in 1999.
•

The prior regulatory baseline was established by the 2009 Performance
Assessment Baseline Calculation (PABC-2009).

•

The 2014 Compliance Recertification Performance Assessment(CRA-2014 PA) demonstrates continued compliance of the WIPP withfederal containment requirements.

• The CRA-2014 was submitted to the EPA in March, 2014.
•

A number of changes/refinements were included in the CRA-2014 PA(e.g. incorporate new data and experimental results).

•

Following its review of DOE’s application, the EPA recertified WIPP facility
in July, 2017.

10

66

CRA-2014 Timeline
PA Calcs.

CRA-2014
Submittal

EPA Comments/
DOE Responses

Planning/
Preparation

EPA-Requested PA
Sensitivity Studies

Documentation/
Review

2012

2013

EPA
Completeness
Determination

2014

2015

DOE/EPA
Tech.
Exchange
Radiological
Release at
WIPP

2016

2017

DOE/EPA
Tech.
Exchange
DOE/EPA
EPA
Tech. Recertification
Exchange Decision(operational impact,
but no CRA/PA impact)

11

Three Sensitivity Studies
 EPA-directed (EPA-mandated
parameters)
 “…the EPA requested that the DOE…
conduct a set of sensitivity studies toaddress some of the significant
technical concerns arising from the
EPA’s CRA-2014 review.” –EPA FRN
 Not compliance calculations
 Formally analyzed by SNL Codified in EPA Federal Register Notice
regarding WIPP recertification decision
1267

Three Sensitivity Studies
EXP CRA14_SEN2: Operations (OPS) andOPS
Experimental (EXP) Areas CRA14_SEN3: OPS/EXP + Panel
Closures (PCS) CRA14_SEN4: Chemistry, brineintersection probability, etc.

PCS
13

Three Sensitivity Studies
Name

Changes to Baseline

Results Summary

CRA14_SEN2

Faster creep closure of
OPS/EXP areas

Insignificant increase inreleases

CRA14_SEN3

Faster creep closure of
OPS/EXP areas and
PCS

Minor increase inreleases

CRA14_SEN4

Actinide solubility
uncertainties

Significant increase inreleases

Prob. intersecting brine
Error corrections

14

68

Three Sensitivity Studies
Total Release Comparisons

CRA-2014
vs.
CRA14_SEN2
CRA14_SEN3

CRA-2014
vs.
CRA14_SEN4
15

Summary of Sensitivity Studies
 “The inputs to these sensitivity studies broadly address
many of the EPA’s technical concerns that could potentially
impact long-term repository performance.” –EPA TSD
 “The Agency has reviewed the results of these studies anddetermined that there exists an adequate level ofconfidence—that is, a reasonable expectation—that therepository will continue to comply with EPA regulations.” –EPA TSD “The Agency does not consider these studies as being
inclusive in addressing all the technical issues identified in
our CRA-2014 review.” –EPA TSD
16

69

EPA Recertification Decision EPA Federal Register Notice issued July 2017
 “With this notice, the Environmental Protection Agency (EPA or
the Agency) recertifies that the U.S. Department of Energy’s(DOE) Waste Isolation Pilot Plant (WIPP) continues to comply
with the ‘‘Environmental Standards for the Management andDisposal of Spent Nuclear Fuel, High-Level and Transuranic (TRU)
Radioactive Waste.’’ “The Agency has determined that the DOE continues to meet all
applicable requirements of the WIPP Compliance Criteria, andwith this action, recertifies the WIPP facility.”

17

EPA Recertification Decision (continued)
 “The EPA has also identified areas in which the DOE’s technicalanalyses and justifications could be improved for the nextrecertification application.” “The EPA recommends that the performance assessment
technical basis be evaluated for improvement in these areas: (1)Calculations of actinide solubility, (2) modeling the chemical
conditions in the repository, and (3) modeling direct brinereleases.”

18

70

Summary• WIPP PA calculations are performed by SNL and used by
DOE to demonstrate long-term repository performance• The CRA-2014 PA was submitted to the EPA as part ofcompliance recertification application• EPA-requested sensitivity studies aided recertification
decision• The EPA recertified WIPP in July 2017
• The WIPP remains in compliance with long-term federal
containment requirements• The EPA has made recommendations for changes to PA
for later CRA submittals• Thanks to WIPP PA Team for support in all PA calculations
19
performed for CRA-2014 activities71

( in preparation from China )72

Presentation materials for theSpecial Session( 4 Materials )73Nuclear Damage Compensation and Decommissioning Facilitation Corporation

Technical Strategy for Decommissioningof the Fukushima Daiichi NPS
27 November 2017
Nuclear Damage Compensation andDecommissioning Facilitation Corporation

NDF
Kazuyuki KATO
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

Contents

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．Future Actions

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation75

1. Roles & Responsibilities of Organizations Relevant toFukushima Daiichi NPS Decommissioning3

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．R&D Initiatives
７．Enhancement of International
Cooperation
８．Future Actions
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation76２－１．Purpose and Positioning of the Strategic PlanFY 2015

FY 2016

FY 2017

The purpose of the Strategic Plan is to provide a firm technicalbasis for contributing to the steady implementation of theGovernment-developed Medium- to Long-Term Roadmap.

Providing strategic proposals for “determination offuel debris retrieval policies for each unit” and“compilation of the basic concept of processing anddisposal for solid waste”

Published on July 13 th, 2016

Published on Apr 30 th, 2015

Strategic Plan 2016

Strategic Plan 2017
 Strategic Plan 2017 provides
strategic proposals for the milestonesof Mid-and Long-Term Roadmap
based on discussions so far.

Strategic Plan 2015
Concepts and Direction of Initiatives

Basic Attitude The decommissioning of Fukushima Daiichi NPS is a project in which considerable uncertainty inheres.
 To realize prompt decommissioning, it is necessary to assume the attitude of taking flexible and prompt
approach, based on the directions determined with previously obtained experience and knowledge, placing
safety at the top priority, even though a certain extent of uncertainty remains.
→ The attitude of optimizing the entire project from the long-term and comprehensive viewpoint, and the attitudeof making preparations for unexpected cases, are also important.
5

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

２－２．Basic Concept of the Strategic Plan
Fundamental Policy
To continuously and promptly reduce the risks associated with theradioactive materials generated by the accidents
Five Guiding Principles
 Principle 1：Safety→ Reduction of risks * posed by radioactive materials and ensuring worksafety （*）Environmental impacts and exposure to the workers
 Principle２：Proven→ Highly reliable and flexible technologies
 Principle３：Efficient→ Effective utilization of resources(e.g. human, physical, financial and space) Principle４：Timely→ Awareness of time axis
 Principle５：Field Oriented→ Thorough application of Three Actuals(actual field, actual things and actual situation)
6

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

77

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．R&D Initiatives
７．Enforcement of International
Cooperation
８．Future Actions
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

３－１．Progress in Decommissioning of Fukushima Daiichi NPS①
 Over the last year, progress has been made in the following areas.
Contaminated Water Management【Removal】 Removal of contamination resources is underway using advanced multi-nuclide removal system.
【Keeping Away】 Freezing operation of the land-side impermeable walls has bee extended; The operation on themountain side is yet to be completed only at one location. After the operation on the sea side was completedin October 2016, the amount of groundwater pumped up from the area of 4.0m above sea level was reduced
to one-third (an average of about 118 m3/day in March 2017). As a result of implementing preventive/multilayered measures such as subdrain, the amount of groundwater flowing into the buildings decreased to an
average of about 120 m3/day in March 2017.
【Preventing Leakage】 The concentration of radioactive materials in the surrounding sea area is constantly low.
【R/B Stagnant Water Management】 The stagnant water level in the Unit 1 Turbine Building was reduced to thelowest floor level in March 2017.
Removal of Spent Fuel from Spent Fuel Pool
•Unit 1: The removal of the cover wall panels for thereactor Building was completed. Information gathering fordeveloping a rubble removal plan is in progress.
•Unit 2: A yard around the Reactor Building was prepared and aworking platform for access to the operating floor was constructed.
•Unit 3: Measures to reduce the dose rates on the operating floor
were completed. Installation of a cover for fuel removal isunderway.
8

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation78

３－2．Progress in Decommissioning of Fukushima Daiichi NPS②
 Over the last year, progress has been made in the following areas.
Fuel Debris Retrieval•Unit 1: An investigation using a robot in the basement outside the pedestal
and in the vicinity of the access in PCV was conducted.
•Unit 2: A survey inside the pedestal in PCV using a robot and otherdevices was conducted.
•Unit 3: An investigation using muon detection system and an investigationin PCV using a robot were performed.
Waste Management•The waste reduction measures are continued to be in place. An operation to reduce waste protective
clothing with an incinerator was started.
•The solid waste storage management plan was updated.
•Sampling and analysis is underway in order to characterize the solid waste.
Other Specific Measures•Since the rubble removal and facing in the area 4m above sea intended for improving site environment
have resulted in the reduced risks of contamination, the classification for the protective equipment has
been changed to “ordinary clothing area”.

9

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

３－３．Risk Estimation and Evaluation
 Risk level defined by the combination of the level of the impact in case of
the release of radioactive materials and the likelihood of impact
 Risk estimation and evaluation is performed based on the SED score*developed by the U.K. Nuclear Decommissioning Authority (NDA)*Safety and Environmental Detriment Score(1) Hazard Potential (Index of the level of impact)
Sample of risk analysis of the Fukushima Daiichi
 Use SED score as it is.
 Consider the amount of all radioactive materials, theform of risk sources (e.g. gas, liquid and solid) and theavailable time for recovery in the event of a loss ofsafety function.

(2) Safety Management (Index of the likelihood of
impact) Partially revised to flexibly cope with the situation of the
Fukushima Daiichi NPS referencing the SED score.
 Graded and quantified the risk sources by the factors ofintegrity of the facility, containment function and apossibility of change in the state of the risk sources.

It will be appropriate to develop risk reduction measures
based on the classification such as Category Ⅰ ~ Ⅲ.
【the interim goal】 To bring their risk levels to the “region of
sufficiently stable management ”

10

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79

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．R&D Initiatives
７．Enhancement of International
Cooperation
８．Future Actions
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

４－１．Basic Concept for Ensuring Safety in the Fuel Debris Retrieval Operation“Concept of safety function” and “principles for realizing safety functions” have been developed
Taking into account factors, such as the international safety fundamentals, the safety-related characteristics of the
Fukushima Daiichi NPS and the risk associated with the fuel debris retrieval operation.

Concept of safety function
Features to ensure the containment function (release prevention and management) and measures to prevent
loss of the features are necessaryDecay heat removal and criticality control, as well as prevention of pulverization and spread of fuel debris to the extent
possible should be consideredPreparedness for normal operations as well as for anticipated external events, such as earthquakes and tsunamis, andinternal events, such as failures and operator errors, is required

It is necessary to observe the dose limits for workers under normal operating conditions to reduce the exposure ofworkers to the extent possible, and to observe the emergency dose limits for workers engaged in emergency response
activities

Principles for realizing safety functions
Considering that the reactors after the accident are decommissioned, it is important to effectively use existing
equipment, to provide the necessary equipment and to consider the combination of equipment and work
managementIn order to reduce the exposure of workers engaged in preparatory work and maintenance work, it is important toensure the containment function by maintaining the PCV at a negative pressure and to consider agile action using both
permanent and mobile equipment

In applying defense-in-depth, it is necessary to carefully consider the hierarchy of defense levels and the need for the
independence of each level

It is important to recognize the presence of uncertainty in the conditions in the PCV and to be able to change plans in a
flexible manner

In safety assessment, it is important to set a realistic management goal and assessment conditions. For example,
actual lifestyles should be considered.

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４－２．Current Status of Each Unit

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４－３．Feasibility of the Fuel Debris Retrieval Method
 In order to evaluate the feasibility of the fuel debris retrieval method, which
are based on combination of water level and access routes, three priority
method have been selected.

a. SubmersionTop Access methodb. Partial Submersion Top Access method

c. Partial Submersion Side Access method

Fuel debris is accessed from the top.
PCV will be completely filled withwater or filled with water to a levelat which the fuel debris is present.

Fuel debris is accessed from thetop. Part of fuel debris will be
dealt with in the air withoutwater covering.

Fuel debris is accessed from theside of the PCV. Part of fuel debris
will be dealt with in the air
without water covering.

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４－４． Evaluation of Feasibility of the Methods ①
 Evaluate the feasibility based on the status of R&D, occupational radiation
exposure and so on.
Submersion-Top AccessPartial Submersion - Top AccessPartial Submersion - Side AccessLiquid Phase
Gas Phase

Ensuring
Containment
Capability

・Hard to ensure water seal capability of・Technical difficulty is slightly lower as・Technical difficulty is slightly lower as
resisting hydrostatic pressure when beinghydrostatic pressure is lower than that ofhydrostatic pressure is lower than that ofsubmerged
submersion case.
submersion case.
・Hard to ensure capability to remotely fix
・Penetration holes on upper PCV to be fixed ・Penetration holes on upper PCV to be fixedpenetration holes for upper PCV with lots
are limited.
are limited.
of holes・It is possible to prevent water leakage even ・It is possible to prevent water leakage even・Emergency water leak prevention measureon emergencies depending on water levelon emergencies depending on water levelis required as a large amount of water issettings.
settings.
supposed to be kept

・Although air conditioning system
・Air conditioning system with capability ・Air conditioning system with capabilityof maintaining negative pressure forwith capability of maintainingof maintaining negative pressure forcontaining alpha-emitting nuclides isnegative pressure is required, small containing alpha-emitting nuclides isnecessary. The scale of the equipment necessary. The scale of the equipment
scale equipment may be good
will be large, but it can be feasible.
will be large, but it can be feasible.
enough.

・Preventing criticality when reactor ・There is a low probability of re・There is a low probability of reCriticality
core is covered with water is an
criticality as reactor core will not be criticality as reactor core will not beManagement issue.
covered with water.
covered with water.
・Although the total weight of coolant in ・Although the total weight of fuel
Structural
PCV and fuel debris retrieval
debris retrieval equipment to be
Soundness /installed at upper R/B increases,
Seismic Resistant equipment to be installed at upper
R/B increases, seismic margin will be seismic margin will be ensured for
Features of PCV
ensured for major components.
major components.
and R/B

・Better seismic margin will be ensured
as fuel debris retrieval equipment will
be installed on the first floor.

・Occupational radiation exposure
・Occupational radiation exposure
would be of several times of the past would be less than the past annual
annual total exposure when sealing
total exposure when sealing lower
upper PCV as there are lots of
PCV.
penetration holes on the upper PCV.

・Occupational radiation exposure would
be less than the past annual total
exposure when sealing lower PCV.

Reducing
Occupational
Radiation
Exposure

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４－５． Evaluation of Feasibility of the Methods ②
Submersion-Top AccessPartial Submersion - Top AccessPartial Submersion - Side Access

Inner
RPV
PCV
BottomEstablishing
Access Route・Scale of work concerning retrieval ・Scale of work concerning retrieval ・Building an access route to fuelof fuel debris located in RPV could of fuel debris located in RPV could debris located in RPV is difficult at
be significant as inner structures
be significant as inner structures of present.
of reactor must be removed.
reactor must be removed.
・Scale of work concerning retrieval ・Scale of work concerning retrieval ・Scale of work concerning retrieval ofof fuel debris located at theof fuel debris located at the bottom fuel debris located at the bottom ofbottom of PCV could be moreof PCV could be more significant
PCV could be less significant thansignificant than that of sidethan that of side-access method as that of top-access method.
access method as it is required to it is required to bore the bottom ofbore the bottom of RPV.
RPV.

Evaluation of feasibility
Submersion-Top AccessPartial Submersion - Top AccessPartial Submersion - Side Access・Development of technologies
for remotely fixing
penetration holes for watersealing is difficult.
・Total occupational exposure
concerning repair work could
be enormous.

・It is necessary to continue
development of technology
for maintaining negative
pressure in order to containalpha-emitting nuclides.
・Both top-access and sideaccess would be required.

・It is necessary to continue
development of technology
for maintaining negative
pressure in order to containalpha-emitting nuclides.
・Both side-access and topaccess would be required.

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４－６．Comprehensive Evaluation
Evaluation of the water level during the fuel debris retrieval operationIn the partially submersion method, though it is necessary to develop a gas-control system and contain radioactive
materials in the gas-phase portion, we should further accelerate the R&D process and evaluate the applicability of themethod to work in the field toward materialization of the partially submersion method in all Units 1 to 3.

With discussing the feasibility of the submersion method in the future scope, we should also store the knowledge obtainedin the R&D process and remaining issues in an appropriate manner to prepare for the possible use of them.

Access routes
At the moment, it is considered difficult to access the fuel debris inside the RPV from the side for retrieval. It is necessary
to access it from the operating floor (from the top).



It is more realistic to access the fuel debris at the bottom of the PCV and retrieve it from the side of the PCV (on the firstfloor level of reactor building). In terms of reducing the exposure of workers and performing maintenance work, it isreasonable to access it from the side of the PCV.

Locations in the reactor from which fuel debris should be retrieved first

It is realistic to retrieve the fuel debris from the bottom of the PCV first by accessing from the side.
The reasons are:1.

The bottom of the PCV in Units 1 to 3 has been investigated and, as a result, a certain amount of knowledge
on the routes for accessing the bottom from the side has been accumulated, which can be used for realistic
engineering.

2. The actual time to reach the fuel debris will be longer when accessing the inside of the RPV from the top thanwhen accessing the bottom of the PCV from the side.
3. In order to streamline the decommissioning process as a whole, preparations for side access to the bottom of
the PCV can be made concurrently with the removal of the fuel from the pool.

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４－７．Proposal for Deciding Fuel Debris Retrieval Policies andApproach after Decision (Strategic Proposal) ①
Proposal for Drafting Fuel Debris Retrieval Policies(1)Develop a comprehensive fuel debris retrieval plan aimed to optimize the entire retrieval process,from preparation work and transfer from the site to treatment, storage and cleanup, including
coordination with other works in the field.
(2)Move forward in a flexible manner according to the information gained little by little via a step bystep approach after deciding the retrieval method to be focused on.
(3)Assume that combination of a variety of methods will be required to complete the fuel debris
retrieval.
(4)Promote preliminary engineering and R&D focusing on the partial submersion methods.
(5)Firstly, focus on retrieving the fuel debris located at the bottom of the PCV and keep reviewing themethods based on the newly gained expertise/experience through the retrieval.
(6)At first, focus on the route from the side of the PCV (the side-access method) for the first access tothe fuel debris located at the bottom of the PCV. The following are the points to be stressed about themethod.
At “Uncertainty”, it is important to analyze small but importantsuccessful experiences, expand the scope and accumulate useful
information gradually（Step by Step Approach）

Judging thefeasibility

The basic concept
which seems mostreliable

1. Implementation

To be United

ModifyEngineering

Result of internal survey

2. Reflecting result

Fuel debris
retrieval

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83

４－８. Proposal for Deciding Fuel Debris Retrieval Policies andApproach after Decision (Strategic Proposal) ②
 Once the fuel debris retrieval policies have been defined, the following priority issues
should be dealt with towards “determination of fuel debris retrieval method for the
first implementing unit” and accelerating the development of the actual construction
planPreliminary engineeringThe applicability of the results of research and development and system concept to the actual field will be evaluated and theprocesses involved in the retrieval will be specified.

As needed, the fuel debris retrieval methods may be revised based on the results of the preliminary engineering.

Acceleration of technology development based on the selection and prioritizationof research and development projects and practical application of technology
Extra investigations into internal PCVInvestigations into internal RPVJudging feasibility of alpha emitting radionuclides management system required for partial submersion methods

Promoting the necessary research and development projects to materialize the side-access method into practicaluse and discussing the significance of mockup facility

Conducting research and development on a system for containing, transferring and storing fuel debris, on the
preparation of a storage facility, and on the waste to be generated by fuel debris retrieval

Path to Commencement of Fuel Debris Retrieval
 It is important to remember the following when promoting the fuel debris project.
"Considerations to the continuity of the project", "Optimization of the entire retrieval process" and"Close communication with the local government and community"

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無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．R&D Initiatives
７．Enhancement of International
Cooperation
８．Future Actions
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation84

５－１． Study Policy on the Strategic Plan for Waste Management
 It is considered that the solid waste generated due to the accident has different characteristics. Hence,characterization of solid waste has been conducted, on the basis of which the study towards processing
and disposal has progressed. The management such as storage of solid waste also has been
implemented in accordance with risk reduction approach.
 The international safety principles on radioactive waste management are compiled and the policies onsolid waste management to be applied are shown, in light of the current state that solid waste
characteristics data has been accumulated.

International Safety Principles on Radioactive Waste Management
 Described below are some principles in the safety principles on radioactive waste
management compiled by ICRP and IAEA for all radioactive waste
 To contain all radioactive waste and isolate it from the accessible biosphere To be processed to have stable solid form, and reduced volume to the extent possible
and immobilized during the management activity of radioactive waste

 At various steps in the predisposal management of radioactive waste, theradioactive waste shall be characterized and classified.
 During the predisposal management, storage is important as a measure to givethe management with flexibility.
 To be stored in such a manner that it can be inspected, monitored, retrieved andpreserved in a condition suitable for its subsequent management.
 The anticipated needs for any future steps in radioactive waste management have to betaken into account as far as possible in making decisions on the processing of the waste
21

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

５－２．Current State of Solid Waste Management
Effort１．Storage

Current State Solid waste is safely stored continuously according to its type and surface dose rate.
Volume reduction of solid waste such as used protective clothing is being carriedout by operation of miscellaneous solid waste incineration facility.
 In the solid waste storage management plan, the amount of solid waste generated
in the next ten years is estimated, and a plan is shown to reduce volume of waste
stored outdoors as much as possible and transfer to indoor storage in order toachieve further reduction of the risk on storing solid waste.

Miscellaneous solid waste incineration facility

Outline of solid waste contamination source and nuclide migration pathway

22

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５－３．Current State of Solid Waste Management
EffortCurrent State

２．WasteCharacterization



３．
Processing
and disposal








Sampling and analysis were carried out mainly on “rubble, etc.” and secondary waste from water
treatment generated after the accident. Analysis results of about 300 samples have been accumulated
over the past 6 years.
Although some contamination characteristics of solid waste are being estimated based on
accumulation of analysis results, further analysis is still required. In addition, in order to promotecharacterization, a method of understanding the characteristics by using analysis data and nuclide
migration model evaluation data is being developed.
For secondary waste from water treatment system, the technologies for stabilization andimmobilization have been developed with priority. In order to narrow down the applicable candidatesof conditioning technology, fundamental tests for solidification have been conducted with simulated
non-radioactive waste using proven technologies. Through these tests, data was obtained on thepossibility of solidification and the soundness of the solidified products.
As for disposal, waste classification methodology has been studied on a trial basis for existing disposal
concepts as an example.
The case-based study of overseas disposal facilities has been conducted in order to contribute to thestudy on disposal concepts which is based on the characteristics of solid waste

Sampling（JAEA）

Examples of sampling location23

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５－４．Proposal for Compilation of the Basic Concept of Processing andDisposal for Solid Waste (Strategic Proposal) ①
Characteristics of Solid Waste

Described below are the characteristics of solid waste that have been presumed.


The amount of solid waste is greater than that of radioactive solid waste generated by normal operation of
nuclear power stations, and a lot of solid waste have relatively high dose rate.

As the major source of contamination is fuel debris, radioactive concentration of nuclides in solid waste does notexceed that of spent fuel.

Composition and radioactive concentration of nuclides in solid waste have a lot more varieties compared withthose of radioactive solid waste generated by normal operation of nuclear power stations.



Secondary waste generated from water treatment has high fluidity, high dose rate that can causes thegeneration of hydrogen, and the materials/substances which have never been dealt with in Japan.

Solid waste generated immediately after the accident include materials/substances whose chemical
characteristics might impact on the safety of the predisposal and disposal and/or whose chemical hazard might
affect the environment.



Information about the total amount of solid waste and their characteristics, which are essential for discussions
on disposal, will be revealed sequentially as decommissioning activities proceeds.

Solid Waste Management Policies
Characterization, storage and preceding processes in predisposal management are mainly focused on until a prospect of disposal is obtained.

◇Thorough containment and isolation ◇Reduction of the amount of solid waste◇Promotion of characterization⇒ Increasing analytical capability systematically by constructing required facilities/equipment, keeping on
analytical talented staff and bringing up analytical staff, and improving characterization by R&D from theperspective of efficiency.
◇Thorough storage ◇Establishment of a selection system of preceding processing methods in consideration of
disposal⇒ Establishment of a selection system of preceding processing methods for stabilization and immobilization ofsolid waste before technical requirements are set.

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５－５．Proposal for Compilation of the Basic Concept of Processing andDisposal for Solid Waste (Strategic Proposal) ②
 Present Efforts/R&D Based on Solid Waste Management Policies
Thorough containment and isolation


The measure such as store in containers or immobilization is performed as needed. Solid waste is isolated bystoring in the storage place which is set up appropriately.

Reduction of the amount of solid wasteContinue such efforts as carry-in control, re-use/recycling and volume reduction.

Promotion of characterization【Increasing analytical capability】 Construct new facilities/equipment and use existing ones systematically from a mid-tolong term perspective, and establish a system for human resource development (HRD) and technology transfer.
【R&D for improving characterization from the perspective of efficiency】 Develop a characterization method withcomplimentarily combining analytical data and evaluation data based on nuclide migration models, and facilitate R&D
for optimizing the number of analytical samples and for simplifying analytical methods to speed up analysis.

Thorough storageReduce the amount of solid waste that have been stored temporarily outside of storage facilities by moving
them to storage facilities after volume reduction as much as possible.
Discuss the methods of evaluating the amount of hydrogen gas generated from the secondary waste fromwater treatment during the period of storage, estimate the timing of implementing additional safety measuresand consider what kind of the measures will be required.
Study storage method for solid waste to be generated by fuel debris retrieval.

Establishment of a selection system of preceding processing methods in consideration of disposalEstablish a selection system of preceding processing methods. In this system, safety assessment ofspecifications of a provisional waste package is performed relatively in each developed disposal concept. The
selection of preceding processing method depends on the result of safety assessment.

Efficient implementation of R&D projects from the perspective of overall solid waste management

Development of a system for continuous operations
Measures to reduce the exposure of workers to radiation25

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

１．Introduction
２．Strategic Plan
３．Strategy for Reducing Risks Posed
by Radioactive Materials
４．Fuel Debris Retrieval
５．Waste Management
６．R&D Initiatives
７．Enhancement of International
Cooperation
８．Future Actions
無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

87

８．Future Actions
 The importance of project management will increase as the decommissioning activities at the
Fukushima Daiichi NPS progress to an advanced phase and technically difficult activities, such asfuel debris retrieval, are fully performed

Enhancing Project Management Capabilities

The fuel debris retrieval operation, which is technically difficult and needs to be performed,
taking into account the relationship with and the continuity of the activities that have been
performed, will be fully performed.
Project risk management that identifies the risks which may impact on the implementation of
the project, analyzes the significance of the risks and take required actions is essential.
TEPCO, as the organization responsible for the implementation of decommission, and NDF, as
the organization responsible for the management and supervision of decommission, need toclearly define their roles and accountability, and strengthen project governance in order tofacilitate sound progress of the project, respectively.

Engagement with Stakeholders
 Based on providing plain and accurate information, it is necessary to have good conversation,
make continuous effort mutually in order to reduce the gap between the providers and receivers,
and repeat these processes.
 Just anxiety about potential risks may cause serious harm to the reputation. To avoid further
reputational damages, it is more significant, than anything else, to manage radioactive materials
in an appropriate manner in order not to let them leak and reduce the existing risks promptly.

Consideration for the Project Sustainability
 It is necessary to ensure the mechanism that enables continuous project management, etc. andto secure a large variety of human resources taking the roles in these works

27

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation

Thank you for your attention!
Technical strategic plan 2017 is available
URL:http://www.dd.ndf.go.jp/jp/strategic-plan/index2017.html

28

無断複製・転載禁止 原子力損害賠償・廃炉等支援機構©Nuclear Damage Compensation and Decommissioning Facilitation Corporation88

1
The 6th Conference of East Asia Forum on Radwaste Management,
27 November 2017 Osaka, Japan

Study onPost-accident Waste Management Scenarios
for Fukushima Daiichi Nuclear Power Station
Satoshi Yanagihara
Subcommittee on Radioactive Waste Management
Study Committee on Decommissioning of the
Fukushima Daiichi NPP
Atomic Energy Society of Japan

2

Consequences of the Fukushima Daiichi NPS Accident

Contamination of theCooling/ground water

Secondarywaste

Trimmed branches
cut-down trees

Contaminatedwater Approx. 120-130 t/d of contaminated watercollected (as of Oct., 2017)
 Generation of secondary waste from water
treatment Accumulation of Tritium containing water A need for disposal of tanks storing HTO in
the future

Contamination of
the Environment
Rubbles

Pipes and tanks,
etc.

Soils

 Wide area contaminated Contamination of soil and groundwater A significant amount of waste generated
from clean up and remediation activities

Water treatment

Severe Contamination inside
the Reactor Facilities
Fuel
DebrisRadioactive Inventory before the Accident-> around 6 × 1019 Bq (in the Unit-1)
The amount released into the environment-> around 7 × 1017 Bq (Total)

Reactorcomponents

Facilities andbuildings

 High dose rate inside the facilities Difficulty in radiological characterization
 Measures to ensure nuclear safeguards andsecurity

Sources: Special committee of Atomic Energy Society of Japan, Treatment and disposal of radioactive waste arising from Fukushima Daiichi NPS accident, 2015
Shunsuke Uchida, et. al, NUCLEAR TECHNOLOGY VOL.188 DEC. 2014, JAEA-Data/Code201-018

893Mid-and-long-Term Roadmap towards Decommissioningof Fukushima Daiichi Nuclear Power Station Units 1-4

http://www.meti.go.jp/english/earthquake/nuclear/decommissioning/index.html

4Organization of the Review Committeeof Atomic Energy Society of Japan

Review committee

Risk Assessment and Management

Robotics

Radioactive Waste Management

Structural Performance of Reactor Building

Follow-up the Proposal for Lessons Learned905Outline of the Activities by Subcommittee ofRadioactive Waste Management
Literature review on past accidents and lessons learned(Examples) As taken into account in the Fukushima Daiichi Decommissioning Roadmap, astrategy must be developed to identify the steps of post-accident clean out,
including key milestones, hold points and end points. -----IAEA

 The development of a long-term strategic approach to the lifecycle design andoperation of nuclear facilities, which takes into account waste management,
decommissioning and site remediation, as well as the reuse of sites. ----OECD/NEA
Study on Radioactive Waste Management Scenarios Survey of the radioactive waste arising from the post-accident clean-up phase Discussion on the interim and final end points for decommissioning and remediation
 Estimation of the radioactive waste arising from all phases until site release

6

Radioactive Waste Storage Facilities
at Fukushima Daiichi NPS

Solid waste
storage facility

Cesium absorption vesselstorage facility

Used protective
clothing, etc.

91

Temporary storage ofradioactive waste

7Radioactive Waste Arising fromthe Post-Accident Cleanup Phase Secondary waste by treatment of contaminated water：
9,972 m3 + 1,216 Bessels+2,580 Containers+9 Columns (Building) Debris ：216,200 m3
 Trimmed Trees：133,700 m3
 Contaminated water（886 Tanks） *：1,041,878 m3
Water Stagnanted in Reactor Facility
3
Facility
Volume（m )Unit-1
6000
Unit-2
12990
Unit-3
12770
Unit-4
13640
Total
45400

Description of the 1000 m3 Tank
• Diameter：12 m• Hight：10 m
Water Stagnanted in Storage Building• Capacity：1000m3
3

Facility
Process Main Bld.
High Temperature
Incineration Bld.
Total

Volume （m )
4,060
3840
7900

Source: TEPCO, October, 20178

Timeline from Accident to the End State for Site Release

Stabilization

Emergency
Response

Safe
Enclosure(Optional)

PostAccident
Cleanup

Remediation

Decommissioning

Care &
Surveillance

Operation
Accident
OccursEnd State of
Post-Accident
Cleanup

End Stateof Site Release
End State of
Decommissioning

IAEA, Experiences and Lessons Learned Worldwide in the Cleanup and Decommissioning of
Nuclear Facilities in the Aftermath of Accidents, Nuclear Energy series, No.NW-T-2.7, 201492

End Stateof Remediation

9

Present Status of the Radioactive Waste
at Fukushima Daiichi NPS

Water
treatment
facilities

 Water treatment produces the secondary waste, such as
adsorption towers, treatment columns, and filters.

 Solid waste treatment will be effective in volume reduction andstabilization. These facilities need to be dismantled anddisposed of, once all of the solid waste is treated.
 Treated water containing tritiated water is stored in tanks.
About 900 tanks are in use as of October 2017, which amounts
to about one million tons of waste tanks after disposition oftritiated water.

Units 1 - 4

 Approximately 350 thousandof solid waste, as of October
2017, was generated from the post-accident cleanup andstabilization activities, and is stored on-site. They are sorted by
the types as well as the dose rate.

Solid waste
storagefacilities

Units
5 and 6

1F Sitem3

Solid waste
treatment
facilities
Treatedwater
storagetanks

10Radioactive Waste Management Scenarios~ End State Options ~End State
Condition
OP-1 Green Field

OP-2 Brown Field

Facilities on Site

Site Remediation

Radioactive Waste
Management

Completely dismantled
and demolished.

Decontaminated andremediated to a levelthat allows an
uncontrolled access.

Radioactive waste istransported out to aninterim storage or adisposal facility.*

Partially dismantled.
Remaining facilities
enclosed for a long-term
storage.

Decontamination andremediation for longterm care andmaintenance.
Restricted access.

Radioactive waste isconditioned, packaged,
and stored on site until
a final disposal site
becomes available.

*: There is an agreement between national government and Fukushima prefectural government that the final disposal facilities for
Fukushima accident waste will be constructed outside Fukushima prefecture.

93

11

Estimation of Waste Arising from
Decommissioning

Units 1 - 4

Water
treatment
facilities
Solid waste
treatment
facilities
Treatedwater
storagetanks

 Dismantling of the Units 5 and 6 will generate 700 to 800
thousand tons* of waste. Most the waste is expected to
be non-radioactive, similar to the facilities that have notexperienced accidents.
 Dismantling of the Units 1-4 is expected produce 800 to
900 thousand tons* of waste. Most or all of these waste
may be radioactively contaminated from the explosions.
 Demolition of service buildings and site remediation will
generate more waste.
OP-1

Volume of
Waste

Solid waste
storagefacilities

Units
5&6

1F SiteOP-2

site cleanup/
preparation

decommissioning

remediation

time

12

Major Waste Management Decisions to makefor Fukushima Daiichi NPS Decommissioning• Emergency Response• Stabilization• Post-Accident Cleanup

Fuel Debris
RemovalDecommissioning

Care &
Surveillance

Remediation

Accident
OccursAction Plan for Tritiated Water
Waste Management Systems- Release to the environment, or- Interim storage on or off site, or- Continued storage- Final disposal→ Work space for fuel debris→ Waste segregation andremoval, and/or fuel debris
characterization system
storage area
Radioactive waste management decisions

94

Decommissioning End State
Option- Complete dismantling/
demolition, or- Partial dismantling/enclosure andcare/maintenance13

Considerations for the Analysis ofRadioactive Waste Management Scenarios
Safety of the public, the work force, and the environment

Proper Waste Management

Minimization of the Overall
Environmental Impact

Effective
Project Management

 Physical, chemical, andradiologicalcharacterization
 Waste minimization
 Handling / conditioning /treatment / storage /disposal Reduction/mitigation of theenvironmental impact not only
via physical measures, but
economical and administrative
actions combined
 Balancing between the benefitof decommissioning/remediation and the impact
of waste management

Balancing the competing
project constraints
Scope/Quality/Schedule/
Budget/Resources/Risks
Integrated risk management
Cost overrun/Schedule
delay/Technical feasibility

14

Concluding Remarks
 The subcommittee is developing potential radioactive waste management scenarios forthe waste arising from Fukushima Daiichi NPS decommissioning and site remediation.
 In the process, we have come to realize that defining the final end point anddetermining the long-term strategy toward the goal is essential in considering waste
management options.
 The end point options for decommissioning and site remediation will be discussed on
the technical point of view, including both the green field and the brown field options.
 The waste management scenarios leading to the end point will be discussed andevaluated thoroughly and carefully in terms of feasibility, radiation safety, technology
availability, and consistency with the national framework.
 The subcommittee will discuss and identify the requirements on waste management
imposed by the constraints of space and time for decommissioning/remediation
activities (i.e. securing the space for handling/storage for fuel debris, interim storage
facility and treatment /conditioning facilities).

95

15

Subcommittee Members of Radioactive
Waste Management
Satoshi Yanagihara (Univ. of Fukui)
Yuichi Niibori (Tohoku Univ.)
Naoko Watanabe (Hokkaido Univ.)
Daisuke Kawasaki (Univ. of Fukui)
Masahisa Watanabe (JAEA*)
Katsumi Kishimoto (JAEA*)
Noriko Asanuma (Tokai Univ.)
Kazuya Idemitsu (Kyushu Univ.)
Shunichi Suzuki (Univ. of Tokyo)
Takao Ohi (JAEA*)

Masaaki Kaneko (Toshiba Corp.)
Osamu Kontani (Kajima Corp.)
Kenji Noshita (Hitachi-GE Nuclear Energy)
Shigenobu Hirusawa (NSRA*)
Kazuto Maeda (MHI*)
Yasuaki Miyamoto (JAEA*)
Toyoaki Yamauchi (JAPC*)
Toshiyuki Nakazawa (MITSUBISHI MATERIAL)

JAEA: Japan Atomic Energy Agency
JAPC: The Japan Atomic Power Company
NSRA: Nuclear Safety Research Association
MHI: Mitsubishi Heavy Industrie96

EAFORM 2017 Osaka, November 2017

JAEA’s R&D for Decommissioningof Fukushima Daiichi Nuclear Power Station

Shinichi NAKAYAMA
Fukushima Research InstituteJapan Atomic Energy Agency (JAEA)1

JAPAN ATOMIC ENERGY AGENCY
 Government-funded nuclear
research and development
institute

Horonobe Established in 2005
(JNC and JAERI merged)

 3,100 personnel
Aomori 5 R&D sectors

Tokai, Oarai(Ibaraki) Fukushima
 Nuclear Safety Research andEmergency Preparedness

Fukushima

TsurugaTokyo

 Decommissioning andRadioactive Waste
Management

Tono

Ningyo-toge
Harima Fast Reactor

 Nuclear Science Research297Which is Fukushima?

A

B

C

3

A

C
Shikoku island

1.96 million / 13,783 km2

B

Australia
498March 11th, 2011
Hydrogen explosions TEPCO HP
Evacuation
 Acute external exposure was nil Acute internal exposure was minimal
 Remained were regional scale of contaminatedarea and >160 thousand of evacuees→ required minimization of long-term lowlevel radiation exposure from 134,137Cs
Remediation and living back in order needed Environmental decontamination(Clean-up)
 Infrastructure rebuilding

5

EVACUATION ORDER
Difficult to return

IitateMarch 31,2017

Not permitted to live

Minamisoma

Ready to be lifted

Namie

March 31, 2017

Kawamata
March 31,2017
Fukushima evacuees
164,865 May 2012
129,154 June 2014
79,446 February 2017
In Fukushima Pref.
To outside Fukushima

39,60839,818

TomiokaMarch 31
2017
6

99

Radioactive Contamination in Fukushima（August 2011）>19 mSv/hr

7

MISCONCEPTION- MY PLACE IS NOT CONTAMINATED People in
NE District
think

People near
Tokyo think

People in
Hokkaido
think Hokkaido

North Eastof JapanTokyo

People in
Osaka &
Kyoto think

People in
Okinawa
think

The World
think

Kyoto
Osaka

Okinawa
http://www.yukawanet.com/archives/3930193.html

100

8

JAEA R&D BASE IN FUKUSHIMA“OFFSITE” AND “ONSITE”
TEPCO Holdings
Fukushima Daiichi NPS

Minamisoma City
Fukushima Environmental
Safety Center (monitoring)
Fukushima City
Fukushima Office

Okuma Town
Okuma Analysis and ResearchCenterUnder construction

Miharu Town
Fukushima Environmental
Safety Center

Tomioka Town
CLADS - Collaborative
Laboratories for Advanced
Decommissioning Science

Iwaki City
Iwaki OfficeNaraha Town
Naraha Remote Technology
Development Center2 offices and 5 R&D bases
300 staff in Fukushima

9

DECOMMISSIONING R&D
NARAHA – CLADS - OKUMA
Naraha Remote
Technology
Development
Center

CLADS - Collaborative
Laboratories for Advanced
Decommissioning Science

Okuma
Analysis andResearch Center(in Tomioka)(to be opened 2021)

Development of- tele-operated andremote sensing
technical tools,- training tools fordecommissioning
workers, and- emergency
responding robots

Core of JAEA’s R&D,
- acting as both aninternational research hub
and a symbol of the sharedinternational interest- (OECD/NEA News Bulletin)- strongly linked with Naraha
Center and Okuma Center

Characterization ofradioactive wastes
and fuel debris, forradioactive waste
management, andnecessary R&D

101

Naraha Remote Technology Development Center (1/3)
 JAEA’s R&D base in cooperation
with academia and industries Launched in autumn 2015, 38 users (JFY 2016) and more than
8,000 visitors since the opening

IRID Full-scale Plug Test

1/8 portion of Suppression Chamber

IRID: International Research Institute for Nuclear Decommissioning

Mock-up stairs

11

Naraha Remote Technology Development Center (2/3)

Motion capture studio(equipped with 16 tracking cameras)

Virtual Reality system, 1F
reactor data incorporated, for operator training
 robot simulator
Robot testing water tank 12102

Naraha Remote Technology Development Center (3/3)- Education &Training for Decommissioning -Decommissioning
Robot Competition
13 colleges competedorganized by
National Institute ofTechnology, Fukushima College
funded by
MEXT(December 3, 2016)

Robotics Summer School
The University of Tokyo
funded by MEXT
(August 8-10, 2016)

DECOMMISSIONING R&D
NARAHA – CLADS - OKUMA
Naraha Remote
Technology
Development
Center

CLADS - Collaborative
Laboratories for Advanced
Decommissioning Science

Okuma
Analysis andResearch Center(in Tomioka)(to be opened 2021)

Development of- tele-operated andremote sensing
technical tools,- training tools fordecommissioning
workers, and- emergency
responding robots

Core of JAEA’s R&D,
- acting as both aninternational research hub
and a symbol of the sharedinternational interest- (OECD/NEA News Bulletin)- strongly linked with Naraha
Center and Okuma Center

Characterization ofradioactive wastes
and fuel debris, forradioactive waste
management, andnecessary R&D

103

13

CLADS TOMIOKA
Research: R&D topics includes
 PCV dose rate evaluation
 Laser-based in-situ remote analysis
 Radioactive Particle Characterization
 Radiation Imaging using Compact
Compton Camera
mounted Drone
Human Resource Development
through R&D

 BWR core degradation
behavior investigation

Fukushima Research
Conference on postaccident waste
management safety(2016)DECOMMISSIONING R&D
NARAHA – CLADS - OKUMA
Naraha Remote
Technology
Development
Center

CLADS - Collaborative
Laboratories for Advanced
Decommissioning Science

Okuma
Analysis andResearch Center(in Tomioka)(to be opened 2021)

Development of- tele-operated andremote sensing
technical tools,- training tools fordecommissioning
workers, and- emergency
responding robots

Core of JAEA’s R&D,
- acting as both aninternational research hub
and a symbol of the sharedinternational interest- (OECD/NEA News Bulletin)- strongly linked with Naraha
Center and Okuma Center

Characterization ofradioactive wastes
and fuel debris, forradioactive waste
management, andnecessary R&D

104

Okuma Analysis and Research Center
Under construction Fully open in JFY 2021

 plays an essential role in terms
E&T/human resource development for;
hot facility construction,
hot facility maintenance,analysis technicians, andradiochemists
Laboratory-2

Lab-1

Laboratory-1
Administrative
Building

Admin.
Bldg.

105

Lab-2

Fukushima Daiichi NPS

 first large scale hot laboratory constructed in Japan to deal specificallywith post-accident radioactive wastes (> 1Sv/h @ surface) and fuel debris, world unique facility specific to severely-damaged nuclear reactors
 300 workers including facility operators, engineers, analysis
technicians and radiochemists

17

Distribution of Radionuclides nearby
the Fukushima Daiichi Nuclear Power Station(FDNPS)
Fukushima Environmental Safety Center
Japan Atomic Energy Agency ( JAEA )Terumi DOHI

EAFORM 201727 November 2017 Osaka, Japan
2

Outline1. Background• Regional contamination and Waste management challenge• Radiocaesium (Cs) distribution in the Fukushima Prefecture• Background and Objective of the F-TRACE project

2. Researches on transport behavior of Cs in• Forest and River system

3. Predicitive modeling• Mechanistic approaches• Radiocesium behavior in the Ukedo river system

4. Summary106

3Outline1. Background• Regional contamination and Waste management challenge• Radiocaesium (Cs) distribution in the Fukushima Prefecture• Background and Objective of the F-TRACE project

2. Researches on transport behavior of Cs in• Forest and River system

3. Predicitive modeling• Mechanistic approaches• Radiocesium behavior in the Ukedo river system

4. Summary
For Revitalization of Fukushima…

4

Understanding of the Cs Contamination status in the living sphere Where is the contaminated place especially ?

Source What is the characteristics of Cs contamination ?
 How and When do the contaminants migrate ?
(contaminant behavior)

Determination of decontamination place and method
 Based on the Decontamination Pilot Project

Understanding of the kind of waste and their amount
Waste treatment and final disposal
107

5

Regional contamination (Development of decontamination)

Pruning and removal of branches
and leaves

FDNPS

Stripping of topsoil

High pressure water washing
Distribution map of air dose rate in the Fukushima
prefecture. (measured from the 1st Sept. to the 7th Nov. 2014 by
NRA and corrected to the 7th Nov. 2014

JAEA, The report of decontamination project5
in evacuation zone. (2012))

6

Waste Management Challenges
Cs removed or
reduced materials

Reuse

Volume Reduction

Soils & wastes

Cs accumulatedmaterials

16~22Mm3

current estimation

Wet particle-size
separation

Final disposal
Incineration Facility R&D Facility

 Clean-up efforts are generating

Volume (Mm3)

Contaminated soil
and wastes

Storage Facility

Emplacement &Segregation Facility
Monitoring facility

10.35Mm3

10.06Mm3

20km

Incinerated ash

Katsurao

Namie

Okuma
KawauchiTomiokaNaraha

1.55Mm3

00.01Mm3Below
8kBq/kgHeat treatment

Public Information Center

105Chemicaltreatment

Over
8kBq/kg
below
100kBq/kg

Hirono0.02Mm3

Over
Incinerated
Waste in
100kBq/kg ash over countermeasure
100kBq/kg
area over
100kBq/kg

Contaminated soil and wastes

108

Futaba1F

Interimstoragesite

huge volumes of contaminatedsoil and waste, which must be
managed in a safe and costeffective manner, wherever
possible implementing waste
volume reduction
 Future reuse of soil for
construction purposes is an
important option, if constraints
in terms of allowable organicand clay content can be
managed6

How to Enhance Fukushima Environmental Resilience7

Address the needs for better scientific andtechnological capabilities to assess, predict,and minimize the impact of radiologicalcontamination
Enhance the understanding of radiation andassociated risks in the public
Kai Vetter, 2015

The concept of Resilience；
The ability to recover from or more successfully adapt to adverse events
7

Fukushima people concern about…

 Where is the Cs contaminated places in the living
sphere ?
 Will the Cs contaminants move to their living areain the future ?
 Will the Cs contaminants redistribute despite adecontamination ?
109

8Initial Cs distribution: Forests occupied the mostLand
use
Forest

Area
(km2)
5,330

Initial Cs
(TBq)
920

Total
Forest%

8,370
64%

1,300
69%

9

Wasteland
Paddy field (14%)
Beach
Ocean
River and Lake

Forest (64%)

Built-up area (5%)
road / railways
Golf field
Crop field (10%)
Other land use
revised Kitamura et al., Anthropocene (2014).
JAEA website
https://fukushima.jaea.go.jp/QA/ftrace/QA3-1f.html

Land use distribution in eastern Fukushima

10

Background of the project
Redistribution and evolution ofconcentration of radioactive Cs

Consequence of redistribution
and evolution

transfer to forest
products

Internal exposuretransfer to agricultural / aquatic
products and drinking water

air dust
Cs ?

wash off
degradation
sorption /desorption

dissolved
Cs

sedimentation

particulate
Cs

dissolution

erosionrun-off / river transport

sedimentationnear living sphere
accumulation in forests
near living sphere

External exposure
110transport tothe ocean

11

Research area
Fukushimacity

8 river systems,
4 dam lake,
3 forest sites

・By comparing behaviors in several river
systems with different characteristics,
Koriyama
we intended to clarify factors thatcity of Cs.
dominated transport

Kawamata
Yokokawa dam

Ota

Odaka
Ogaki dam
Ukedo/
Takase
Maeda

FDNPS

South of NPS
Sakashita dam

Kuma

Takigawa dam

・Systematic research on each river
system from forests to estuary makes it
possible to evaluate the behavior of Cs inwhole catchment.

Tomioka

Kawauchi
Ide
Kido

Distribution map of air dose rate in the Fukushima prefecture.
(measured from the 1st Sept. to the 7th Nov. 2014 by NRA and corrected to the 7th Nov. 2014 )12

Research area reported in this presentation
Fukushimacity

Kawamata

Ogaki dam
Ukedo/
TakaseKoriyama
city

FDNPS

2 forest sites & 2 river systems

KawauchiDistribution map of air dose rate in the Fukushima prefecture.
(measured from the 1st Sept. to the 7th Nov. 2014 by NRA and corrected to the 7th Nov. 2014 )
111

13

Outline1. Background• Regional contamination and Waste management challenge• Radiocaesium (Cs) distribution in the Fukushima Prefecture• Background and Objective of the F-TRACE project

2. Researches on transport behavior of Cs in• Forest and River system

3. Predicitive modeling• Mechanistic approaches• Radiocesium behavior in the Ukedo river system

4. Summary
14

Research area reported in this presentation
Fukushimacity

Kawamata

Diciduous Forest

Koriyama
city

Ogaki dam
Ukedo/
TakaseFDNPS
Evergreen Forest(Japanese cedars)
KawauchiDistribution map of air dose rate in the Fukushima prefecture.
(measured from the 1st Sept. to the 7th Nov. 2014 by NRA and corrected to the 7th Nov. 2014 )112

Cs transport processes in forests

15

Rain

fall

Infiltration

Surface runoff
Soil loss

The investigation result of
caesium distribution in the forest(Press release (2014.4.1) byministry of agriculture, forestry
and fisheries of Japan)

Basement

Litter fall

Stem flow

Throughfall

Cs moves downward along with time

Organic layer
Mineral soil

Water, litter and soil movement16

Cs-137 wash-off rate accompanied by the particulate matter.
2013
2014
2015*
2016
(duration) (duration) (duration) (duration)
Kawauchi (KA)

Evergreen(Japanesecedar)

0.10%

0.06%

0.30%

0.04%

(6.10-11.18) (4.11-10.8) (6.24 - 10.28) (6.16 - 11.28)

Steep slope
Kawamata(KE)
T. Niizato et al., J. Environ. Radioact.,
161, 11-21 (2016).

Kawamata(KW)

Deciduous
Steep slope

Wash-off rates were quite low inall observation areas.

0.02%

0.10%

0.05%

0.11%

Deciduous (3.29-11.19) (4.7-10.20)
Gentle slope

0.23%

0.02%

(6.28-11.19) (4.7-10.20) (6.30 - 11.5) (6.15 - 11.29)*New observation plot was set up.

reivised T. Niizato et al., J. Environ. Radioact., 161, 11-21 (2016).

113

Water, litter and soil movement

2013.7.18
Surface runoff
2
Soil loss 289 kBq/mlitter2014.7.21
410 kBq/m2

litter
2012.8.28
368 kBq/m2

Depth (cm)Depth (cm)Cs deposition (kBq/kg)

Cs deposition (kBq/kg) Cs deposition (kBq/kg)

litter
2012.12.13
368 kBq/m2

Cs deposition (kBq/kg) Cs deposition (kBq/kg)

litter2015.7.26
359 kBq/m2

Depth (cm)litterDepth (cm)Cs deposition (kBq/kg)

2012.1.17
215 kBq/m2

Depth (cm)2011.7.2
260 kBq/m2

litterLitter fall

Depth (cm)litterStem flowfall

Cs deposition (kBq/kg)

Depth (cm)Rain

Depth (cm)Throughfall

Cs deposition (kBq/kg)

17

litter2016.8.7420 kBq/m2

Basement

Infiltration

Evolution of depth profile of Cs in the Japanese cedar
forest floor in Kawamata.

Organic layer
Mineral soil

Downward infiltration of Cs wasquite slow so that Cs concentrations
in litter and topsoil were stable.
Y. Onda, results of commissioned study by JAEA

18

Consequently, Cs tends to be preserved on the forest floor andcycled in the forest ecosystem.
114

19

Research area reported in this presentation
Fukushimacity

Kawamata

Ogaki dam
Ukedo/
TakaseUkedo river
Takase river

Koriyama
city

FDNPS

KawauchiDistribution map of air dose rate in the Fukushima prefecture.
(measured from the 1st Sept. to the 7th Nov. 2014 by NRA and corrected to the 7th Nov. 2014 )20Transport in river system: Ukedo/Takase river

Ukedo river Takase river
Dam

Ukedo river

Contamination levelat Upper stream〇（at middle)

×HighlyLess

Dam (Ogaki dam lake)

Takase river
FDNPS

115

21

Cs in river system
Riverbed

River

Riverbed
Suspended solids(SS)bound Cs
Dissolved Cs

Cs of bottom sediment

In the Typhoon

Ordinary water level

Comparison of behavior in rivers with/without dam
SS （left） and radiocesium （right） concentration in lower Ukedo (with dam)and Takase (without dam) river waters in the typhoon in 2013.
300>60 um250（Bq/L）

Ukedo

TakaseDissolved 137Cs

0.31
±0.03

0.05
±0.02

SS-bound 137Cs

2.2
±0.2

2.1
±0.1

134+137Cs

3.3

2.9

3-60 um

SS (mg/L)

200

0.45-3 um

150

100

50

0

Ukedo

Takase SS: Ukedo << Takase
 SS supplied upstream did notreach to lower Ukedo river due tothe presence of the dam.
 Dissolved Cs: Ukedo > Takase
 Higher Cs fallen out along
mountain streams was brought.

JAEA website
https://fukushima.jaea.go.jp/QA/ftrace/QA4-1-3-2.html

116

22

23

Stock and flow of Cs at the high water event in Sept. 2015
Takase riv.

6

Ukedo riv.

Ogaki dam lake

90% of SS flowing into the lake
was expected to be deposited.

180
7
千t/km

38

1.7

0.1

230

60
SS
137Cs Conc.

kBq/kgDischarged
SS (103 t)

千t/km

Deposited 137Cs
on floodplain

12

Depositedsoils (103 t)
137Cs

140

1.5

840

Discharged 137Cs to
the ocean

activity(GBq)

90 In spite of highly contaminated upper, smaller amount of Cs was discharged to theocean from Ukedo river than Takase river due to deposition in Ogaki dam lake.
JAEA website

https://fukushima.jaea.go.jp/initiatives/cat01/pdf1702/hokokukai11.pdf

24

Cs discharged to the ocean in FY 2015
 Discharge rates of Cs to the ocean were
extremely low, especially in rivers withdams.

Estimated amount of discharge
Annual discharge
River
Amount(TBq)

Discharge at the highwater event on Sept.

Ratio toAmountinitial(TBq)deposition

Ratio toannualdischarge

Odaka

0.074

0.5 %

0.053

72 %

Ukedo

0.22

0.1 %

0.090

48 %

Takase

1.2

0.4 %

0.84

79 %

Maeda

0.057

0.1 %

0.035

73 %

Kuma

0.31

0.5 %

0.23

79 %

with dam

Tomioka

0.022

0.1 %

0.013

58 %

JAEA website

https://fukushima.jaea.go.jp/initiatives/cat01/pdf1702/hokokukai11.pdf

with dam117

25

Outline1. Background• Regional contamination and Waste management challenge• Radiocaesium (Cs) distribution in the Fukushima Prefecture• Background and Objective of the F-TRACE project

2. Researches on transport behavior of Cs in• Forest and River system

3. Predicitive modeling• Mechanistic approaches• Radiocesium behavior in the Ukedo river system

4. Summary
26

Mechanistic approach
Suspended solid

advection dispersion inflow/outflow sedimentation/resuspension

Dissolved Cs

advection dispersion decay inflow/outflow sorption/desorption to SS and river bed

SS-bound Cs

advection dispersion decay inflow/outflow sorption/desorption sedimentation/resuspension

Y. Onishi and W.A.Perkins.,
TODAM. In “the U.S.
Department of Energy,
under Contract DE-AC0676RLO 1830”, (1994).

118

27

Prediction of Cs sedimentation on the flood plain
Nays2D: 2D transport model for contaminant in rivers

11 Dec. 2012

Distribution map of air dose
rate by air-borne survey
Y. Sanada et al., Exploration
Geophysics, 45(1), 3-7 (2013).

Accumulation density of Cs-137（KBq/m2）

Distribution map of accumulation of Cs in Ukedo river after the typhoon on Sept. 2011.
A. Malins et al., AGU Fall Meeting (2015).

Areas of higher accumulation density of Cs seems good agreement withthose of higher air dose rate. => applicable to predict sedimentation behavior.
28

Prediction of Cs discharge from the river
3 days soil loss [kg]

3 days water flow [m3]6.0E+75.0E+7
4.0E+73.0E+72.0E+7
1.0E+7

3 days discharge of 137Cs [GBq]

GETFLOWS (physical model for soil loss and water transportation):
Water flow, soil loss and Cs-137 discharge for 3 days during each high water event.
1.0E+78.0E+6
6.0E+6
4.0E+6
2.0E+6
0.0E+00.0E+06.0E+25.0E+24.0E+2
3.0E+22.0E+2
1.0E+20.0E+01.0E+75.0E+78.0E+6

4.0E+7

Soil loss [kg]Water flow [m3]6.0E+7

137Cs discharge [GBq]

Relationship to
precipitation

3.0E+72.0E+7
1.0E+70.0E+06.0E+6
4.0E+6
2.0E+6
0.0E+0

0

100

200

300

Precipitation per event [mm]

0

100

200

300

Precipitation per event [mm]6.0E+25.0E+24.0E+2
3.0E+22.0E+2
1.0E+20.0E+0
0

100

200

300

Precipitation per event [mm]It can be applied to predict the behavior of Cs depending on precipitation.
A. Kitamura et al., AGU Fall Meeting (2015).

119

29

Estimation of behavior of Cs in the river system
 Stock and flow of Cs in the Ukedo river system can be estimated based on
the models, and discharge flux agreed with the observation results.

0.96 TBq/yUkedo riv.

Ogaki damdownstream
0.07 TBq/y
請戸川：
約0.2 TBq/年

高瀬川：
約0.5 TBq/年

Ukedo riv.:
0.2 TBq/yTakase riv.:
0.5 TBq/yTotal flux
to theocean
1.0 TBq/y

3 km

Distribution mapping1) of air dose rate and overview of flux from each catchment of
Ukedo River basin estimated from observation results2),3),4) and model calculation5).
1) http://ramap.jmc.or.jp/map/map.html (as of 2014/11/7, in Japanese)2) Ishii et al., Proceedings of ICONE-23 (2015).
3) Kurikami et al., J. Environ. Radio., 137, 10 (2014).

4) http://fukushima.jaea.go.jp/initiatives/cat03/pdf06/2-4.pdf (inJapanese)5) Kitamura et al., Anthropocene, 5, 22 (2014).

30

Outline1. Background• Regional contamination and Waste management challenge• Radiocaesium (Cs) distribution in the Fukushima Prefecture• Background and Objective of the F-TRACE project

2. Researches on transport behavior of Cs in• Forest and River system

3. Predicitive modeling• Mechanistic approaches• Radiocesium behavior in the Ukedo river system

4. Summary
120

31

Summary• Forest• Less discharge from forest with soil (<0.1%/y) leads tocontinuous accumulation of Cs in the forest floor, mostly within
10 cm depth.
⇒ Further investigation on; generation processes of dissolved Cs,
 future cycle in the forest, especially to trees, mushrooms, wildlife, etc.

• River system• Dissolved Cs: <1 Bq/L in all river systems.
• Discharge to the ocean: 0.1 – 0.5 %/y.
• Transport of Cs can be predicted.
⇒ Further investigation on; generation and buffering processes of dissolved Cs,
 future transfer to agricultural and aquatic products.

For Revitalization of Fukushima…

32

Understanding of the Cs Contamination status in the living sphere

Efficiency
strategy

Prediction &Estimation

 Where is the contaminated place especially ?
 What is the characteristics of Cs contamination ?
 How and When do the contaminants migrate ?
(contaminant behavior)

Determination of decontamination place and method
 Based on the Decontamination Pilot Project

Understanding of the kind of waste and their amount
Waste treatment and final disposal
121

33

Acknowledgement• Regional municipalities gave

us opportunities of researchin forests, rivers, lakes andestuaries in detail.
• Minami-soma city
• Namie town
• Futaba town
• Okuma town
• Tomioka town
• Naraha town
• Kawamata town
• Kawauchi village

Thank you for your attention !

122

Abstracts for theTechnical Sessions* Full-Length Papers for the technical sessions will be uploaded at the following web-site.
http://eaform2017.aesj.or.jp/123Session 1A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

High-frequency melting technology of radioactive metal waste
Nobutake Horiuchi, Takeshi Nishikawa, Naoto Sasaki and Hirokazu Tanaka
Energy Project & Technology Center
Mitsubishi Materials Corporation
1002-14 Mukoyama, Naka-city, Ibaraki-pref., Japan
Key words; Melting, radioactive metal scrap, decontamination, volume reduction, recycling
Abstract
Mitsubishi Materials Corporation has been developing a melting decontamination technology
of metal waste by high-frequency melting, for volume reduction, decontamination andrecycling of metal scrap contaminated with radioactive materials from nuclear power plants
and nuclear facilities. By the melting treatment, high volume reduction of metal waste anddecontamination of radionuclides such as cesium, strontium and α nuclides are expected. In
addition, the melting process is also suitable for the metal recycling, because the process canhomogenize the waste and cast it into any shape.
The boiling point of Cs is lower than the melting point of iron (1,600 deg.-C), so iron-based
metal waste can be decontaminated by volatilizing Cs. On the other hand, Sr and α nuclides,
such as U, utilize the property that they tend to be more oxide than iron. In theelectromagnetic field of the high-frequency induction furnace, since electromagnetic force
does not work on oxide, it is ejected to the outward as the metal moves to the center of thefurnace. Furthermore, the metal has good homogeneity because molten metal is stirred byelectromagnetic forth.
Through development, we confirmed the followings; - the volume of metal waste can be
reduced from 1/5 to 1/3 by melting, - the radioactivity concentration of Cs of metal that
actually contaminated due to a nuclear accident can be set below the detection limit, demonstration scale test with simulated substances can decontaminate Cs, Sr and U. In
addition, we have been developing manufacturing technology to make the metal waste
containers by centrifugal casting method, which was applied to advanced volume reduction
facilities of Japan Atomic Energy Agency (JAEA).

125

Session 1A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

C-14 release behavior and thermal decomposition characteristics of crud particles
collected from the coolant filter of commercial PWR
Shintaro Tsujia, Toshikazu Wakia, Akira Sakashitab and Kenichiro Kinoc
a The Kansai Electric Power Co. Inc.
b Mitsubishi Heavy Industries, LTD.
c Nuclear Development Corporation
a 3-6-16 Nakanoshima Kita-ku Osaka-shi, Osaka 530-8270 Japanb 16-5 Konan 2-chome, Minato-ku, Tokyo 106-8215 Japanc 622-12, Funaishikawa, Tokaimura Ibaraki 319-1111 Japan

Key words; Waste Disposal, Radionuclide migration, Safety assessment, C-14, TG/MS
Abstract
In the safety assessment of radioactive waste buried repository, it is extremely important toevaluate the migration behavior of long half-life radionuclides in the disposal environment.
C-14 is a key radionuclide in waste disposal because of the long half-life (5,730years) and theconsiderable effect on public exposure due to the less capability of capturing C-14 byartificial barrier material. In addition, C-14 is a radioactive isotope of carbon which forms
various chemical forms in the process of its formation, and the migration behavior in theburied environment greatly differs depending on its chemical form.
Some previous studies have reported the generation of insoluble C-14 in pressurized water
reactor (PWR), and suggested that ion-exchange resin, which is used for purifying primarycoolant, could be its origin. However, this hypothesis has not been well evidenced yet,because the information about the generation mechanism and the chemical form of insoluble
C-14 is still limited.
In this study, in order to investigate the chemical form of C-14 contained in radioactive
particles generated by PWR, radioactive particles collected from the coolant filter of PWR
were analyzed by TG/MS, and the gas released by TG and residual C-14 in the compounds
was also analyzed at the same time.
As a result, organic molecules presumed to originate from ion exchange resin were detectedfrom the released gas components, and it was confirmed that 90% or more of C-14 wasdetected in residual components of heating in the inert gas.
From the results of this study, the chemical form of insoluble C-14 generated by PWR
could not be clarified, but it was revealed that the majority of C-14 may exist in amorphousand thermally stable compounds.

126

Session 1A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanEVALUATION OF CARBON 14 RELEASE FROM IRRADIATED ZIRCALOY FUEL
CLADDING THROUGH A LONG-TERM STATIC LEACHING TEST
Hiroyoshi Ueda, Tomofumi Sakuragi, Naoki Fujii and Hitoshi Owada
Radioactive Waste Management Funding and Research Center (RWMC)
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan
Key words; geological disposal, spent nuclear fuel, TRU waste, C-14, Zircaloy, corrosion
Abstract
Zircaloy cladding of the spent nuclear fuel, which is highly activated and contaminated, is
expected to be disposed of in a deep underground geological repository, both in theonce-through and recycling fuel cycles. Carbon 14 (C-14), mainly originated from thenitrogen impurity in Zircaloy cladding and formed through the 14N(n, p)14C reaction in thereactor, is a typical activation product. In a preliminary performance assessment of the deep
geological disposal in Japan, C-14 gave a significant dose impact, which is due to therelatively large inventory, relatively long half-life, i.e. 5730 years, higher release rate and thechemical speciation and consequent migration parameters. In a preliminary Japanese safetycase, relatively high IRF (instant release fraction), 20%, was assumed regarding the oxide
film formed on the metal surface as a source of the instant release.
With respect to this source term issue, we have prepared an irradiated BWR fuel cladding(Zircaloy-2, average rod burnup of 41.6 GWd/t) which has an external oxide film of 25.3 μm
thickness. Although the specific activity in oxide was 2.8 times the base metal activity due tothe additional C-14 generation by the 17O(n,α)14C reaction, the C-14 abundance in the oxide
was less than 10% of total inventory. A static leaching test using the cladding tube was carriedout in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) atroom temperature. After 6.5 years, C-14 was found in each leachate fraction of gas phase anddissolved organics and inorganics, the total of which was less than 0.01% of the entire C-14
inventory of the immersed cladding tube. Both the C-14 abundance and the low leaching ratesuggests that C-14 in oxide does not have a significant impact on the IRF in the safety case.

127Session 1B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

The Assessments and Comparisons of RASCAL and EPZDose on Atmospheric
Dispersion and Dose Consequences in Radioactive Material Release Accidents
Shaohsuan Chen1,2, Jongrong Wang 1,2, Chunkuan Shih1,2, YuTing Ku2, Yu Chiang2,
ShaoWen Chen2 and WenSheng Hsu3
1 : Nuclear and New Energy Education and Research Foundation
2 : Institute of Nuclear Engineering and Science, National Tsing Hua University
3 : Nuclear Science & Technology Development Center, National Tsing-Hua University.

No. 101, Section 2, Kuang-Fu Road, Hsinchu, Taiwan 30013, R.O.C.
Key words; Dose, EPZDose, RASCAL, SBO
Abstract
RASCAL has been developed by NRC for making dose projections during radiologicalincidents and emergencies. It evaluates atmospheric releases from nuclear power plants, spent
fuel storage pools and casks, fuel cycle facilities, and radioactive material handling facilities.
EPZDose is a dose assessment code designed and developed by NTHU. It calculates the dose
consequences according to the input time-varying source term and the meteorology conditions
and is capable of demonstrating the dose dispersion progress. In this study, we compared theatmospheric dispersion and dose consequences evaluated by RASCAL and EPZDose. First,
we used RASCAL to simulate a long term SBO accident for Maanshan NPP and exported thecalculated released source term. And then we converted the source term file generated by
RASCAL to the input source term format of EPZDose. Finally, We used RASCAL and
EPZDose respectively to evaluate the dose consequences of that specific long term SBO
accident with the source term under the same postulated meteorology conditions andcompared the assessment results of these two codes. The accumulated thyroid dose
distribution map at 8 hours and 1 day after the start of release obtained by RASCAL and
EPZDose appeared to be very similar. We went further to verify by comparing the dose values
derived by the two codes at some specific locations. For example, it showed that, at EPZ
boundary of 8 km in downwind direction after 8 hours of release, the difference ofaccumulated thyroid doses calculated by these two codes is less than 10%. More comparisons
will be demonstrated and concluded in the full paper.

128

Session 1B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Confirmation of appropriate operation condition with Blasting device
Yuuki TSUHARA and Hirotaka IZUKA KANDEN POWER-TECH, Corp.
3-6-2, Bingomachi, Chuou-ku, Osaka, 541-0051, Japan
Shinro HIRANO, Yuji MATSUNAGA, Yasuhiro SUGAHARATHE KANSAI ELECTRIC POWER CO., INC.NUCLEAR POWER DIVISION
13-8, Goichi, Mihama-cho, Mikata-gun, Fukui, 919-1141, Japan
Yoji KANAMORIATOX CO., LTD.
4-11-3, Shiba, Minatoku, Tokyo, 108-0014, Japan
Key words; decommissioning, decontamination, radioactive waste, Blasting
Abstract
The decontamination method of metallic waste was considered to reduce the radioactive
waste in decommissioning the nuclear power plant. The blasting method is widely known as
one of the decontamination techniques. It is necessary to introduce a blasting device thatmatches the type of metals and the metallic waste volume for each plant.
Stainless steel occupies most for material of system equipment of PWR. The radioactive
materials are stuck in the surface in the equipment as metal oxide (e.g. chromium oxide, ironoxide). It is necessary to remove the metal oxides efficiently from the stainless steel with theblasting device.
The type of abrasives, the blasting velocity and the blasting angle as blasting operation
condition were tested and confirmed abrade condition stainless steel. The type of abrasives,
the blasting velocity and the blasting angle were selected as the appropriate operation
conditions from the test results. And the next, the decontamination test using contaminatedsamples was performed under the appropriate operation conditions. The metallic radioactive
waste was confirmed to be able to decontaminate to the clearance level from thedecontamination test results.

129

Session 1B-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Development of In-situ Radioactivity Inspection System
for radioactive waste and decontamination system using microalgae
Sujung Min1, Dohyung Kim1, Woyoung Kim1, Sheunghyun Ha1, San Chae1, Jongman Lee2and Unjang Lee1
ORIONENC Co., Ltd.
#305, 37, Seongsui-ro 22-gil, Seongdong-gu, Seoul, 133-832, Korea1,
Korea Research Institute of Standards and Science
267, Gajeong-ro, Yuseong-gu, Daejeon, Republic of Korea2
Key words; Decommissioning, Microalgae, Inspection, Decontamination, Radioactivity
Abstract
In 2011, Fukushima nuclear power plant accident in Japan caused contamination of lakes
and rivers due to radioactive materials released. As time passes, the radioactivity ofcontaminated water has been lowered, but there is still some level of radioactive
contamination. And, concerns about nuclear safety have been raised globally.
At present, decontamination of radioactive contaminated water uses methods of removing
radioactive materials in polluted water by using resin or filter. These conventional
decontamination methods can lower the radioactive concentration of contaminated water to acertain level. However, it is difficult to reduce the amount of contaminated water to a level
lower than the very low level due to technical and economic reasons. Also, it is atime-consuming and costly process to dispose of a large amount of radioactive waste
generated in the dismantling of a nuclear power plant, and many researches are needed tominimize and recycle radioactive waste in nuclear power plants around the world.
Dismantling of Nuclear power plant is the process of decontaminating structures, equipment
and polluted water, disposing them and restoring the site to a usable natural state that can beused after the nuclear power plant is permanently shut down.
In this study, we propose methods to (1) decontaminate radioactive contaminated waterusing microalgae, and (2) method to automatically detect the amount of activity of radioactive
waste generated from nuclear dismantling process using In-situ Radiation Inspection System
previously developed.

130Session 1B-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

RAPID REMOVAL OF CESIUM-137 FROM URBAN AREA AFTER THE
FUKUSHIMA DAI-ICHI NUCLEAR POWER PLANT ACCIDENT
Kazuya Yoshimura
Sector of Fukushima Research and Development, Japan Atomic Energy Agency
10-2 Fukasaku, Miharu-machi, Fukushima 963-7707, Japan
Key words; Cesium-137, urban, distribution
Abstract
Urbanized areas have a large impact on external exposure to residents. Diverse
decontamination methods had been evaluated by the model decontamination project carriedout by Japan Atomic energy Agency in 2011 and 2012 to reduce the external exposure. On theother hand, quantitative information of distribution of radioactive substances, especially 137Cs,
in urban area is limited, although the information is crucial for decontamination planning. To
provide quantitative information of the distribution of 137Cs in urban area, this study evaluatedrelative 137Cs inventories, which are defined as the relative values of 137Cs inventory on eachurban component such as roof and road to that on a nearby permeable plane field, for 11
buildings in the evacuation zone in 2015 and 2016. Additionally, the results were analyzed
with the data obtained in the model decontamination project to evaluate temporal trend in the
distribution during the initial five years after the accident. The relative inventory obtained forpaved ground were 0.2 and for other building components such as roof, wall and window
were less than 0.04 in 2016. These values indicate the limited contamination of urban area.
The relative inventories of paved ground, roof and wall in 2011 and 2012 showed similarvalues with those obtained in this study, suggesting that initial run-off and the following
wash-off effects during the first year after the accident largely defined the distribution of 137Cs
in urban area. Some of the component showed rapid decrease in the relative inventory morethan decay during five years after the accident. The fast decrease and the obviously low
relative inventories on the components in urban area demonstrate the rapid removal of 137Csfrom urban area.

131

Session 1B-5
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

A Study on the Performance of Flocculating agent for Radioactively Contaminated Soilby Soil Washing Process
Jong Soon Song and Sun Il Kim*
Department of Nuclear Engineering, Chosun University
309 Pilmun-daero, Dong-gu, Gwangju 61452, Republic of Korea
Key words; Soil Washing, Flocculating Agent, Soil remediation
Abstract
Radioactive substances that are discharged in the process of nuclear power plant
decommissioning or facility maintenance cause soil pollution. Such soil pollution requires
decontamination in order to ensure people’s residence and reuse of land as an industrial site.
Among the various soil remediation technologies, this study was selected soil washingtechnology which has actual example of being used on soils polluted by radioactive
substances Korea and abroad of the country. Soil washing process uses washing agent toweaken the surface tension of soil and Cs in order to separate soil and Cs. For this reason, themajor factor that determines decontamination efficiency is the selection of washing water.
While the decontamination process which uses acidic wash water weakens surface tension
greater than the one that uses water, therefore separating Cs and soil with high efficiency, but
it produces much acid effluent. Using water instead of acidic washing water, which produces
great amount of effluent, and reusing washing water with flocculating agent may ensure amore economical and efficient process. In order to enhance the efficiency of soil washingprocess, the washing water can be reused. This study used flocculating agent to perform theexperiment for testing the removal performance of micro-soil and Cs that are included in thewashing water. Cs removal experiment was performed by using ICP-OES. In addition,binding species of Cs to soil according to respective pH was predicted using Visual MINTEQ
code and whether washing watering used in the soil washing process can be reusable wasassessed by turbidity.

132

Session 1C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Spent Nuclear Fuel Final Disposal Management in Taiwan
Allan Lee
Department of Nuclear Backend Management,Taiwan Power Company
7F, Annex Bldg., 242, Sec. 3, Roosevelt Rd. Taipei, 10016, Taiwan
Key words; Spent Fuel, Final Disposal, Radioactive Waste Management
Abstract
Based on “Nuclear Materials and Radioactive Waste Management Act”, Taiwan Power
Company (TPC) has responsibility for taking good care and manage the spent nuclear fuelthat generated from its nuclear power plants. So, in 2004 TPC submitted a “Spent Nuclear
Fuel Final Disposal Plan in Taiwan (hereinafter referred to as the “Plan”)” to the regulatory
authority. According to the Plan (revised in 2014), the period from 2005 to 2017 is considered
the “Potential Host Rock Characterization and Evaluation Stage”, which has 2 missions
including 1: Two milestone reports “Preliminary Feasibility Assessment Report for the Spent
Nuclear Fuel Final Disposal Technology in Taiwan (abbreviated as the SNFD2009 report)”and “Technical Feasibility Assessment Report on Spent Nuclear Fuel Final Disposal(abbreviated as the SNFD2017 report)” are to demonstrate the technical capabilities in Taiwan
and 2: A number of survey areas are recommended to continue more detailed investigation in
next stage.
In this study, some achievements from SNFD2017 report and proposed siting procedures forsurvey areas of Taiwan are shown in this topic. Furthermore, in order to respond the new
government’s policy “no nuclear power plants operate in Taiwan after 2025”, a “task force of
nuclear-free homeland” had been established in Taiwan to discuss national nuclear strategies.
Thus, the relevant current status and strategies are also presented (e.g., the status of a
dedicated agency of radioactive waste management in Taiwan, the future siting procedure ofcandidate sites, and public engagement).

133

Session 1C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanSCIENTIFIC BASIS FOR NATIONWIDE SCREENING
OF GEOLOGICAL DISPOSAL SITES IN JAPAN
Takehiro Matsumoto1, Hiromitsu Saegusa1, Hideaki Hyodo1, Akira Deguchi1,
and Hiroyuki Umeki1
1Nuclear Waste Management Organization of Japan (NUMO)

4-1-23, Shiba, Minato-ku, Tokyo, 108-0014 Japan
Key words; geological disposal, site selection process, nationwide screening
Abstract
The legal system was established for geological disposal of radioactive waste in 2000,
which specifies three steps of siting process as literature survey, preliminary investigation anddetail investigation to be conducted by the Nuclear Waste Management Organization of Japan(NUMO). NUMO has been promoting the siting process by a volunteering approach, but novolunteer so far has appeared and selection of geological disposal sites in Japan has notadvanced significantly over 15 years. Taking such situation into consideration, thegovernment reinforced measures for achieving solutions and promoting geological disposal
through the “Strategic Energy Plan” published in April 2014. In association with thereinforcement, a new process for selecting repository sites, which is the nationwide scientificscreening by the government before initiating legally defined siting process, has been addedby the amendment in the Basic Policy in May 2015. The nationwide scientific screening isaiming at encouraging municipalities to consider accepting the literature survey.
Discussion for the nationwide screening for stepping into the literature survey was initiatedin December 2014 at the Geological Disposal Technical Working Group (WG), which is anadvisory committee for the Ministry of Economy, Trade and Industry. The WG identified
requirements and criteria for the screening, and summarized them in the report which was
published in April 2017.
The requirements and criteria are discussed from the viewpoints of 1) geologicalenvironment characteristics and their long-term stability favorable to geological disposal; 2)safety of repository facilities; 3) safety of waste transportation; 4) technical feasibility.
NUMO has been developing the nationwide knowledge base that contributed to the discussion
by providing the state of the art scientific basis.
This paper presents the detail of the scientific basis of these viewpoints for therequirements and criteria.

134

Session 1C-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanSENSITIVITY OF WIPP PERFORMANCE ASSESSMENT RESULTS TOREGULATOR-PRESCRIBED CHANGES
Todd R. Zeitler and Brad A. Day
Sandia National Laboratories
Address: 4100 National Parks Hwy. Bldg. A, Carlsbad, NM, 88220
Key words: performance assessment, transuranic waste, WIPP, regulatory
Abstract
The U.S. Environmental Protection Agency (EPA) requested of the U.S. Department ofEnergy (DOE) a sensitivity study (CRA14_SEN4) of the current Waste Isolation Pilot Plant(WIPP) Performance Assessment (PA) model that incorporated several changes to thebaseline WIPP PA model (CRA14), including: parameter changes, use of an updated code,
and a revised computational grid. The modifications to the repository model resulted inincreased releases in all primary release mechanisms. The impacts of each EPA-requestedchange to CRA14 were analyzed with respect to each release mechanism in the
CRA14_SEN4 study. Overall, total high-probability (P[Release>R] = 0.1) predicted mean
releases from the repository were increased by about 15%, which corresponds to a 0.6%
reduction in the margin to the limit of 1. Total low-probability (P[Release>R] = 0.001)
predicted mean releases were increased by about 107%, which corresponds to a 2.9%
reduction in the margin to the limit of 10. The upper 95% confidence level on the mean
increased for high-probability and low-probability releases by 18 and 119%, respectively. It
is concluded that the EPA-requested changes to the CRA14 result in increases to the predicted
total releases from the repository. However, releases calculated in the CRA14_SEN4
analysis remain below regulatory limits, demonstrating continued compliance of the WIPP.
Sandia National Laboratories is a multi-mission laboratory managed and operated by National
Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of
Honeywell International, Inc., for the U.S. Department of Energy’s National Nuclear Security
Administration under contract DE-NA0003525. SAND2017-6687A

135

Session 2A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Geochemical modeling of water-rock interactions in granitic rocks, Eastern Taiwan
Yu-Te Chang
Material and Chemical Research Laboratories, Industrial Technology Research Institute8F, -4, No.9, Sec. 2, Roosevelt Rd. Zhongzheng Dist., Taipei City 10093, Taiwan
Key words: granitic rocks, groundwater, reaction-path modeling
Abstract
Crystalline rocks such as granite have been considered as suitable host rocks of radioactive
waste repository. As groundwater flows through fractured-rock aquifers, the water-rockinteractions could induce dissolution and precipitation of minerals and change thegeochemical, structural, and mineral characteristics of the host rocks. Therefore, study ofpossible water-rock reaction processes and mechanisms is crucial for suitability evaluation ofpotential disposal host rocks. The granitic rocks in eastern Taiwan are mainly surrounded bymarble, which could be predicted that the groundwater quality (especially pH value) will
mainly controlled by carbonate system. In this study, the groundwater and rock compositions
in eastern Taiwan were used for reaction-path modeling by 3 scenario cases (shallow open
fracture, deep open fracture and deep closed fracture), which can be used to understand theconditions of rock composition and the surrounding water quality, and their chemical changes
over time, e.g., the variation of pH and Eh and the formation of secondary minerals. The
simulation results show that the calcite and sericite are the major secondary minerals under
the final equilibrium state. Besides calcite and sericite, the rest of the main secondary
minerals will change gradually from andradite, saponite-Ca to dolomite, nontronite-Na, anddawsonite when the fugacity of carbon dioxide increases. For the fractured-rock aquifers ineastern Taiwan granitic rocks, the simulated pH and Eh ranges (pH = 7.12 to 8.93 and Eh =-0.37 volts to -0.15 volts) should be considered in the sensitivity analysis of the nuclide
solubility and/or adsorption characteristics. Compared to the similar simulation results in
Taiwan’s granitic offshore island (pH = 6.99 to 9.75 and Eh = -0.45 volts to -0.17 volts), it
shows that the groundwater characteristics (especially pH value) of granitic rocks in eastern
Taiwan is obviously affected by carbonate system and with less pH fluctuations.

136

Session 2A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Centrifuge model test to gain reliability of the future predictionin terms of long term THM processes in deep geological repository
Soshi Nishimoto
Central Research Institute of Electric Power Industry (CRIEPI)
1646 Abiko, Abiko-shi, Chiba, 270-1194 Japan, soshi-n@criepi.denken.or.jp
Key words; centrifuge, scaling law, near-field, THM processes, geomechanical interaction
Abstract
The long-term behavior in the near-field of deep geological repository is governed by coupled
thermal-hydraulic-mechanical (THM) processes. THM processes will continue for hundredsof years in the initial stage of disposal. These processes influence the overpack displacement,
the swelling behavior of the buffer, the deformation of the disposal hole, and so on. The
overpack displacement is influenced by the swelling behavior of buffer. The swelling pressureof the buffer generated is influenced by the depth of the repository and the stiffness of thesurrounding bedrock. To clarify the long term behavior in the near-field, the researches by thefull-scale tests and the numerical analyses have been carried out. However, the former are
difficult due to location, time, and economic restraints, and the latter are necessary to verify
the applicability of the numerical model. Centrifuge model test can be used to replicate an
event, similar to what can be done with a prototype, and in fact is a reduced-scale version of a
prototype. Based on scaling laws, any two investigations of the same conditions using acentrifuge model test and a prototype are similar and related. However, the centrifuge model
test has the advantage that it can greatly shorten the long time needed to see behavior
resulting from the typically slow flow of groundwater that satisfies Darcy’s law. If the time
acceleration test using the reduced-scale model of the near-field is available on the basis of
the centrifugal scaling law, then the long-term reliability of the geological repository can be
improved by empirical laboratory data. Our aim is; to conduct the time acceleration test using
the centrifugal equipment; to measure the equivalent data of long-term behavior of theoverpack, buffer and bedrock; and to evaluate the long-term THM behavior of the near-field
in a geological repository by laboratory measurements.

137

Session 2A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Thermo-hydro-chemical processes influence on buffer material degradationin high level radioactive waste disposal
Wen-Sheng Lin1, Suu-Yan Liang2 and Chen-Wuing Liu3
1Hydrotech Research Institute, National Taiwan University
2Material and Chemical Research Laboratories, Industrial Technology Research Institute3Department of Bioenvironmental Systems Engineering, National Taiwan University

No. 1, Sec. 4, Roosevelt Road, Taipei, 10617 Taiwan
Key words: high-level radioactive waste, buffer material, bentonite degradation, performanceassessment, geological disposal
Abstract
The concept of geological disposal for high-level radioactive waste (HLW) is based on amulti-barrier system which comprises the natural geological barrier provided by therepository host rock and its surroundings and an engineered barrier system (EBS), including
the waste form, waste canisters, buffer materials and backfill. The multi-barrier system should
perform its desired functions and isolate the waste from the biosphere to achieve theperformance of the HLW disposal. Due to the interaction processes of radionuclide decay heat,
hydraulic and water-mineral reaction in the radioactive waste repository, to investigate thebentonite buffer material degradation resulting from the external temperature and pressure
should be required. However, thermo-hydro-chemical (T-H-C) processes influence on buffermaterial degradation in EBS of HLW disposal repository due to the chemical reaction ofsmectite clay dehydration was seldom addressed in Taiwan. Therefore, the reactive chemicaltransport model is applied to simulate the buffer material degradation affected by the T-H-C
processes of HLW disposal repository. The results can form T-H-C based information andpre-techniques for the performance assessment and safety analysis on the final disposal of
HLW in Taiwan.

138

Session 2A-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Coupled THMC Processes in Radionuclide Waste Management
Gour-Tsyh (George) Yeh
CAMRDA, College of Earth Sciences, National Central University
No 300 Jhongda Road, Jhongli District, Taoyuan City 32001, Taiwan
Key words; Multiphase Flow, Thermal Transport, Reactive Transport, Geo-Mechanics
Abstract
Physics-based mathematical-computational models provide an invaluable tool for repository
design, performance/safety assessments, site clean-up, and environmental remediation in
nuclear spent fuel and high level waste disposal in shallow or deep geological media. The
migration of nuclear wastes is controlled by coupled THMC (Thermal-Hydrology-MechanicsChemical) processes. These include multiphase flow or variably saturated flow, thermaltransport, reactive transport, and geo-mechanical deformation. Multiphase or variably
saturated flow in subsurface media is the key driving mechanism for thermal transport,
chemical transport, and geo-mechanics. Conversely thermal transport, chemical transport, andgeo-mechanic deformation will affect flow directly or indirectly. Temperature changes are
induced by fluid injection, associated phase changes, and chemical reactions. Conversely,
temperature change will alter fluid flow, reactive transport, and geo-mechanical deformation.
Faults and fractures will affect fluid pressure, thermal transport, and chemical migration.
Conversely fluid pressure, thermal transport, and chemical migration will induce rockdeformation and fault displacement.. Reactive transport is controlled by fluid flow, affectedby thermal transport, and influenced by geo-mechanical deformation. Conversely, reactive
transport will have significant feedbacks on fluid flow, thermal transport, and geo-mechanics.
This talk presents the development of a series of computational models that fully or partiallycouple these processes. Demonstrative applications are presented to illustrate the interplay of
THMC processes and their implications for radionuclide waste managements.

139Session 2B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Efforts toward Safe, Steady and Efficient Decommissioning
Toyoaki Yamauchi, Satoshi Karigome and Hitoshi Ohata
The Japan Atomic Power Company
1-1, Kanda-mitoshiro-cho, Chiyoda-ku, Tokyo, 101-0053, Japan
Key words; Decommissioning, Tsuruga-1
Abstract
Before Fukushima accident, only four NPPs, Tokai (GCR), Fugen (ATR) andHamaoka-1/2 (BWR), were under decommissioning in Japan. But other NPPs, Mihama-1&2(PWR), Tsuruga-1 (BWR), Genkai-1 (PWR), Shimane-1 (BWR), Ikata-1 (PWR) and Monju(FBR), were decided to terminate operation after Fukushima accident, and some of those
NPPs started Decommissioning work recently. It is supposed that the number of NPPs indecommissioning phase will continue to increase in the future.
In order to advance decommissioning, there are four principal conditions which have to be
met, that is, 1) Spent Fuel path, 2) Radioactive Waste Disposal path, 3) Decommissioning
fund and 4) Culture, mindset and organization suitable for decommissioning project.
In order to responsibly finish all the activities of the closed nuclear power plants, it isnecessary for all the stakeholders involved in nuclear power generation (government,
regulator, utilities and local stakeholders) to play their roles respectively to establish theconditions above. It is clear that decommissioning cannot be completed without spent fuelpath and waste disposal path.
Because it is also important for Japanese utilities to learn decommissioning experience inother countries, The Japan atomic Power Company (JAPC) signed an agreement with
EnergySolutions (ES) on strategic cooperation for Japanese decommissioning projects for the
purpose of introducing successful international experiences in April 2016. As a first step inthis cooperation, decommissioning know-how developed by ES in the US is being applied to
Tsuruga-1 decommissioning project with verifying its applicability to Japanese
decommissioning projects. Outline of JAPC and ES collaboration will be also introduced.

140

Session 2B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Decommissioning Planning for Unit 1 and 2 at Mihama Nuclear Power Plant
Shinya Kato, Koichi Kamahori, Hiromu Isaka and Shinro Hirano
Kansai Electric Power Co., Inc.
8 Yokota, 13 Goichi, Mihama-cho, Mikata-gun, Fukui-ken, 919-1141, Japan
Key words; decommissioning plan, pressurized water reactor, Mihama Unit 1 and 2,
system decontamination
Abstract
The Mihama Unit 1 and Unit 2 has stably supplied electric power for kansai erea. There has
been a new regulatory standards and a 40-year operation regulation in Japan since theaccident occurred at the Fukushima-Daiichi nuclear power plant of Tokyo Electric Power Co.,
Inc. (TEPCO). Because it was difficult to operate these units based on the standards, we
decided that both of these units would be permanently stopped on March 17 in 2015. We
applied for the decommissioning plan of Mihama Unit 1 and Unit 2 on February 12, 2016,
and an application for approval of changes in Tec. Spec. on August 31 in the same year. It
commenced the actual decommissioning phase upon obtaining the approval from NRA after
going through the review process, and finally, we accepted the approval for thedecommissioning plan of Mihama Unit 1 and Unit 2 on April 19, 2017.
Our decommissioning plan is contributed of a process for about 30 years which has 4 stages.
The first stage is the period to decontaminate nuclear reactor vessels(RVs) and steam
generators(SGs), etc. and to investigate remaining radioactivity in the facilities. The second
stage is the period when the equipment around the nuclear reactor will be dissolved andremoved. The 3rd stage is the period when a nuclear reactor will be dismantled and removed.
The 4th stage is the final period when a building will be dismantled and removed.
As the approval certification has been got, it is now shifted to the first stage indecommissioning. The first work in our decommissioning is the system decontamination
which makes the dose for workers as low as reasonably achievable, and it is about to start themain work for system decontamination.

141

Session 2B-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Radiation Survey and Waste Inventory Estimation for Decommissioning of Taiwan NPP
Chun-Cheng Lin1*, Wei-Sin Chen1, Jyi-Lan Wuu1 and Shu-Jun Chang1
1

Division of HP, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village,
Longtan Township, Taoyuan County, Taiwan

Keywords: Radiation survey, radioactive wastes, MARSSIM, Decommissioning.
Abstract
Both radiation survey and waste inventory estimate are parts of the NPP (nuclear power
plant) decommissioning program issues. Radiation survey including site investigation process
can provide the radiation information of a site or a facility, which can be useful for thedecommissioning planning, waste inventory estimation and its characterization. The site
survey results can also support the final decision on whether the site or the facility willcomply with the regulations. This paper describes the radiation survey and site investigation
methods based on MARSSIM (The Multi-Agency Radiation Survey and Site Investigation
Manual); Waste inventory estimate for decommissioning is also essential in the wholedecommissioning study, since D&D (Decommissioning and Dismantling) methods,
decommissioning cost, packing containers selection and repository design fordecommissioning waste etc. This paper has referred to the methodology of Westinghouse that
introduced the implement for decommissioning waste estimation. The first step for the waste
estimation is to define the classifications of waste by the level of radioactivity. And the second
step is to confirm the origins of radioactive waste such as surface contamination, activation ofmaterials or water processing streams. This paper presents the preliminary outcomes ofradiation survey and waste inventory estimate for decommissioning of Taiwan NPP.

142

Session 2C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Status of radioactive waste and spent nuclear fuel management in Finland
Jari Tuunanen
Fortum Power and Heat Oy, Loviisa NPP
Loviisa NPP, P.O.Box 23, 07901 Loviisa, Finland
Key words; spent nuclear fuel, radioactive waste, decommissioning
Abstract
The main principles of the Finnish radioactive waste management program were fixedalready in 1983 in the decision in principle (DiP) of the ministry of trade and industry ThisDiP fixed the overall responsibilities, schedule and reporting principles of the radioactive
waste producers covering the operational waste, spent nuclear fuel and decommissioning. One
key principle of the decision was that the waste producers are themselves responsible formanagement of all the radioactive waste they produce, including the costs.
The status of the program today is such that all the facilities are ready on site of the
NPP's in Olkiluoto and Loviisa for management of operational radioactive waste. Both
nuclear power plant operators have their own waste treatment and storage facilities as well as
own LILW repositories in operation. The spent nuclear fuel is also stored on the NPP sites.
The spent fuel management program is organized through Posiva, a private company ownedby the NPP operators Fortum and TVO, who has got a construction license for spent fuelencapsulation and final disposal facility. Decommissioning plans have also been prepared as
requested by the DiP of the ministry. The money for decommissioning is reserved in theradioactive waste fund and the aim is to dispose all decommissioning waste to the existing
LILW repositories on site of the NPPs.
This paper summarizes the status of the radioactive waste management program in
Finland, with the main focus on the radioactive waste and spent fuel produced at Loviisa NPP.
It also presents some highlights of research and development program on radioactive waste
management and gives an overview of the plans for future decommissioning.

143Session 2C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Communication activities through dialogue in Japanese geological disposal project ofhigh-level radioactive waste
Ayako Araki, Saki Ikeda, Kumiko Ezaki and Kenichi Kaku
Communication and Public Acceptance Department
Nuclear Waste Management Organization of Japan (NUMO)
1-23, Shiba 4-Chome, Minato-ku, Tokyo, Japan,
Key words; Communication, Dialogue, Geological disposal of high-level radioactive waste
Abstract
The Nuclear Waste Management Organization of Japan (NUMO) was established in October2000 based on the “Specified Radioactive Waste Final Disposal Act” as the authorized
implementing entity approved by the Minister of Economy, Trade and Industry (METI).
Our mission is the “Realization of the Geological Disposal of Radioactive Waste”. Since its
establishment, NUMO has been engaging in communication and research and development(R&D) to fulfill the mission.
As a new initiative for enhancing the site selection process of geological disposal, Japanese
government is now preparing for an announcement of the “Nationwide Map of ScientificFeatures for Geological Disposal”, which categorizes all areas in Japan into four categories.
In such a situation, NUMO is working on raising awareness and enhancing understanding of
the geological disposal project among the public. NUMO is also making efforts to listen to theopinions of the public on geological disposal to understand what the public concerns are.
NUMO has been carrying out PR activities such as advertising on TV commercials andnewspapers to raise public awareness of geological disposal. We are currently focusing on
face to face communication activities such as holding seminars with small group discussions
and outdoor events with a communication vehicle, “Geo Mirai”. Utilizing other measures
including cross-media advertisement, opinion surveys and press releases allows us to reachout a wide range of stakeholders in Japan.

144

Session 2C-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanEFFORTS IN THE FIELD OF EDUCATION FOR JAPANESE GEOLOGICALDISPORSAL OF HIGH-LEVEL RADIOACTIVE WASTE
Kumiko Ezaki, Kenichi Kaku and Shunsuke Suzuki
Communication and Public Acceptance Department,
Nuclear Waste Management Organization of Japan(NUMO)
1-23,Shiba 4-Chome, Minato-ku, Tokyo, Japan
Key words; Education, Nuclear Waste Management Organization of Japan
Abstract
Nuclear Waste Management Organization of Japan (NUMO) was established in October2000 based on the “Specified Radioactive Waste Final disposal Act” as the authorized
implementing entity approved by the Minister of Economy, Trade and Industry (METI). Since
its establishment, NUMO has been engaging in public communication to increase anunderstanding of the importance of the geological disposal project.
The geological disposal is the 100-years project from the start of investigation for exploringa repository construction site till the closure of the repository. Therefore, multiple generations
will be involved in this project. For raising awareness of the geological disposal project as oneof social issues among next generations, especially elementary school and junior/senior highschool students, NUMO has been supporting school teachers and college students who arewilling to work on developing teaching plans and materials. NUMO encourages them toactually teach geological disposal to their students in their school classes. In 2016 NUMO
produced basic teaching materials for elementary school students/junior high school
students/and instructional materials for teachers and distributed them to education committee
and schools across Japan.
NUMO also has been supporting college debate classes on the topic of geological disposal ofhigh-level radioactive waste since 2012. NUMO provides its supports by sending experts onradiation and geological disposal for special lectures, by organizing tours to local nuclear
power plant and a research facility of geological disposal and by answering to various
questions from the students. In addition, NUMO has been carrying out communication
activities using a communication vehicle “Geo Mirai”, where exhibits and a 3D-animation
theater are installed, throughout Japan. Thorough these activities various questions andopinions have been obtained from a wide range of generations which will be utilized to
improve NUMO’s communication.

145

Session 3A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanOVERVIEW OF THE NUMO SATETY CASE AT PRE-SITING STAGE
T. Fujiyama, S. Suzuki, A. Deguchi and H. Umeki
Nuclear Waste Management Organization of Japan
Mita NN Bldg. 1-23, Shiba 4-Chome, Minato-ku, Tokyo 108-0014 Japan
Key words; Geological disposal, Safety case, Vitrified waste, TRU wastes
Abstract
NUMO has developed a safety case for geological disposal at pre-siting stage in Japan. Thissafety case provides the multiple lines of arguments and evidence to demonstrate thefeasibility and safety of geological disposal, which will encourage stakeholder confidence inthe safe implementation of geological disposal and will provide a basic structure for a safetycase which will be applicable to any potential site.
This paper will provide a brief overview of this safety case, emphasizing practical approaches
and methodology which will be applicable for the conditions/constraints during an actual
siting process, with a focus on advances from the “H12 Report”, which formed the basis forestablishing NUMO in 2000 as the implementing organization and is considered the first
generic safety case on geological disposal of HLW in Japan.
The geological environments of various potential host rocks are developed as site descriptivemodels, which are, as realistically as possible, based on the current best understanding ofJapanese geology. A rigorous and flexible methodology for tailoring repository design tovolunteer siting environments and safety assessment models and data have been developed
and applied for demonstration of the technical basis of the safety assessment. The NUMO
Safety Case has been extended in key areas, including assessing extreme geological events
during long-term repository evolution, widening discussion of both operational andpost-closure safety, scenario development based on risk-informed approaches, etc.
The safety case will feature extensive use of advanced knowledge management tools and will
be presented in a web-based communication platform, which facilitates not only integrationand documentation of the huge knowledge base that supports a modern safety case, but alsopresentation of output in a user friendly manner to all stakeholders, including the generalpublic.

146

Session 3A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanGEOSYNTHESIS OF A STATE-OF-THE-ART KNOWLEDGE BASE INTO SDMS
IN THE NUMO SAFETY CASE
Kunio Ota, Hiromitsu Saegusa, Takanori Kunimaru and Saori Yamada
Nuclear Waste Management Organization of Japan (NUMO)
4-1-23 Shiba, Minato-ku, 108-0014 Tokyo, Japan
Key words; Safety case, Site selection, Geosynthesis, Site Descriptive Model
Abstract
During the repository site selection process involving stages of literature survey, preliminary
investigation and detailed investigation, the site geological environment is characterised andthe expanding site-specific knowledge base is integrated in a geosynthesis process into site
descriptive models (SDMs), at a progressively increasing level of detail. The SDMs represent
the key geological characteristics and processes required for repository design and safetyassessment and, as importantly, their temporal and spatial evolution. Of great importance isthus to use the SDMs iteratively to illustrate how repository designs are tailored to the site
conditions identified and, with increasing understanding, tailored to optimise their practicality
and performance.
The NUMO Safety Case demonstrates a state-of-the-art knowledge base of all relevant
geological environments in Japan and the integration of such knowledge into the SDMs,
together with the tools and experience to characterise them to the degree required. Five majorrock types that commonly occur in Japan with sufficient spatial extent at relevant depths, havebeen characterised in terms of their structural, thermal, hydrological, geomechanical andgeochemical features of significance to geological disposal. The five rock types have thenbeen categorised into three groups (plutonic, Neogene sedimentary and Pre-Neogene
sedimentary rocks) by focussing on common distinct characteristics and properties. Nested
SDMs have then been developed stepwise to conceptualise illustrative geological settings foreach of the potential host rock groups at scales of several tens of kilometres, several
kilometres, several hundred metres and decimetres. Critical geological and hydrogeologicalinformation represented quantitatively in the SDMs is derived from an actual andquality-assured dataset of deep geological environments for particular siting environments inJapan. The chemistry of model groundwaters has also been developed through the carefulinterpretation of quality-assured hydrochemical data and geochemical modelling, which
covers a wide range of groundwater chemistry for each potential host rock in Japan.

147

Session 3A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanLAYOUT DESIGN OF UNDERGROUND FACILITIESTAILORED TO SDM IN THE NUMO SAFETY CASE
Takahiro Goto, Yoichi Yamamoto, Satoru Suzuki and Shigeru Kubota
Nuclear Waste Management Organization of Japan (NUMO)
Mita NN Bldg. 1-23, Shiba 4-chome, Minato-ku, Tokyo 108-0014 Japan
Key words; repository design, underground facility layout, HLW, TRU waste
Abstract
NUMO has developed a generic safety case for geological disposal in Japan. This safety caseprovides the multiple lines of arguments and evidence to demonstrate the feasibility andsafety of geological disposal, which will encourage stakeholder confidence in the safe
implementation of geological disposal and will provide the basic structure for a safety casewhich will be applicable to any potential site. This paper provides the outline of trial
underground repository layout designs tailored to the site descriptive models (SDMs) in thissafety case. Suitable emplacement areas were first selected in the repository-scale SDMs
based on the Layout Determining Features (LDFs) defined by the geological features such as
faults and permeable rocks. Next, geometry of repository panels was determined. Regarding
the high level radioactive waste, both through type and dead-end type were applied for thevertical and horizontal emplacement concepts. Then, layout of the repository panel was
designed in the suitable emplacement area considering the Emplacement Determining
Features (EDFs) defined by the features such as water inflow from minor faults and fracturesin the disposal tunnels, ensuring the space for the reserved areas. As a result of comparison
among the above design options, the dead-end type geometry with the horizontal
emplacement concept could be identified as a relatively efficient and economical approach
which has more flexibility to the geological structures.

148

Session 3A-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanPRELIMINARY STUDY OF PRE-CLOSURE SAFETY ASSESSMENT
IN THE NUMO SAFETY CASE
K. Yamashina, S. Suzuki and S. Kubota
Nuclear Waste Management Organization of Japan
Mita NN Bldg. 1-23, Shiba 4-Chome, Minato-ku, Tokyo 108-0014 Japan
Key words; repository, safety assessments, waste treatment
Abstract
NUMO has developed a generic safety case for geological disposal in Japan. This safety casepresents technical evidence to support the feasibility and safety of geological disposal, which
will encourage stakeholder confidence in the safe implementation of geological disposal andwill provide the basic structure for a safety case which will be applicable to any potential site.
The target range of the safety case is not only after closure of the repository but also thepre-closure period. The ongoing update of the safety case for co-disposal of HLW and TRU
waste in Japan will include a more extensive assessment of pre-closure safety than has been
carried out in the past. The pre-closure safety case aims to assure both radiological andconventional safety.
After identifying the evaluation scenarios that take into account the progress of abnormal
situation, an impact analysis is conducted to evaluate the damage to waste packages and/or
workers. In the evaluation of the abnormal state, the relationship between the abnormal statethat may occur at the geological disposal facility and the potential countermeasures areschematized by taking the same approach as the event tree analysis method.
In summary, the methodology for assessment of pre-closure safety has been developed andapplied to provide feedback for optimizing repository design.
For conventional safety for workers, the working environment will be maintained to ensure
worker comfort and safety during normal operation.
Radiological safety for public and workers requires radiation shielding and radionuclide
containment within the disposal facilities in case of operational perturbations. Radiation
control and facility design are based on guidelines applied for other nuclear facilities. Within
the radiation-controlled zones, most operations will be remote-handled or will involve
appropriate shielding, avoiding any significant dose to workers.
Mita NN Bldg. 1-23, Shiba 4-Chome, Minato-ku, Tokyo 108-0014 Japan
Key words; repository, safety assessments, waste treatment
Nuclear Waste Management Organization of Japan

149

Session 3B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Conceptual Study on Disposal Facility for Waste from Decommissioning of NPPs
Norie KAWAHATA, Ryoichi HOJO, Takayuki HAMANAKA and Takashi KOZAWARadioactive waste disposal business division, Japan Nuclear Fuel Limited
504-22, Nozuki, Obuchi, Rokkasho-mura, Kamikita-gun, Aomori-ken, Japan
Key words; disposal facility, low-level waste, nuclear power plant, reprocessing plant,
decommissioning waste, high surface dose rate, lower radiation exposure, skyshine,
external exposure, horizontal emplacement
Abstract
1. Introduction
In Japan, low-level waste from operation of nuclear power plants (NPPs) has been containedin 200-litre drums with solidification materials and converted into waste packages. The waste
packages have been disposed of in near-surface disposal facilities with engineered barriers, ofthe Rokkasho LLW Disposal Center.
In the near future, large-size radioactive waste with much higher radiation level will arise
from decommissioning of NPPs and operation of reprocessing plant. Therefore, we studied aconcept of the disposal facility which can reduce radiation exposure to the public, comparedto the above-mentioned near-surface disposal facilities.
2. Concept of Disposal Facility
For the purpose of more efficient processing and containing of waste from decommissioningof NPPs or operation of reprocessing plant, waste producers have been considering the use ofa cubic container (1.6 m on a side) for packaging of the waste, as well as the 200-lite drum.
Additionally, the waste packages are expected to have much higher surface dose rate. Thus,
there would be a concern about the increase in radiation exposure to the public caused bydirect gamma-ray or skyshine gamma-ray. From the viewpoint of reduce the radiation
exposure to the public, it is effective to adopt not a pit-type facility but a box-culvert-type
facility with higher shield effect.
In consideration of the adaptation of the box-culvert-type facility, waste packages will be
emplaced horizontally by use of a forklift truck, not by a crane as used in the Rokkasho LLW
Disposal Center.
3. Advantages of the Concept
There are the following advantages of the box-culvert-type facilities over the pit-type ones;(1) Lower radiation exposure to the public.
(2) All weather operation(3) May use a candidate area effectively, in case of multi-layered structure150

Session 3B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

An Integrated Framework for Simulating Radionuclide Decay Transport of
Low-level Radioactive Waste with Tunnel Disposal in Nearshore Environment
Chin-Chang Lu, Ming-Hsu Li and Yu-Ru Chen
Graduate Institute of Hydrological and Oceanic Sciences,
National Central University, Taiwan#300 Jhongda Rd., Jhongli, Taoyuan City, Taiwan
Key words: Tunnel disposal, Nearshore, Radionuclide decay transport, Low-level waste
Abstract
Tunnel disposal at nearshore was tentatively proposed as one of final disposal sites for
low-level wastes in Taiwan. The potential disposal facilities was located at 50m~200m below
ground surface and designed with several disposal tunnels. Concepts of safety function are
retardation and contaminant provided by disposal tunnels with multiple engineered barriers
and host rock to prevent leaching of radionuclides into biosphere. An integrated simulation
framework was developed with three levels of finite element grids to facilitate simulations ofradionuclide decay transport with 3-D HYDROGEOCHEM5.6 numerical models. Steadyflow and transient transport were performed in this study. The coarsest level grids serve as
far-field simulations. The middle level grids were developed to describe distributions of alldeigned disposal tunnel to serve as near-field (all tunnels) simulations. The finest grids weredeveloped to investigate near-field phenomena at selected tunnel section. Total radioactive
activities of source term and a total of 11 radionuclide (Pu-238, Am-241, Tc-99, Co-60, I-129,
Cs-137, Ni-59, C-14, Sr-90, Y-90, Ni-63) decay chains as given in the report of Taiwan Power
Company were simulated in this study. Flow boundary conditions at finer level grids wereobtained from flow simulations of coarser level grids. Release rates of radionuclides atcoarser grids were obtained from simulations of radionuclide decay transports at finer grids.
For the first 100 years, release rates of C-14, Tc-99, and I-129 are higher than others. After
100 years, release rates of Ni-59, C-14, and Tc-99 are higher than others. Owing to highsorption of Ni-59 onto bentonite barrier, long-lived Ni-59 with high initial concentrationdeliver low release rates for the first 100 years. Low release rate of I-129 after 100 years wascaused by low initial concentration of I-129. Results demonstrate complicated interactions
among initial concentration, half-life, decay chains, near-field processes, and far-fieldprocesses on radionuclide decay transport.

151

Session 3B-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Sensitivity Analysis of Simulating Radionuclide Decay Transportof Low-level Radioactive Waste in Nearshore Environment
Yu-Ru Chen, Chin-Chang Lu and Ming-Hsu Li
Graduate Institute of Hydrological and Oceanic Sciences,
National Central University, Taiwan#300 Jhongda Rd., Jhongli, Taoyuan City, Taiwan
Key words: Sensitivity analysis, Radionuclide decay transport, Low-level waste, Geosphereparameters
Abstract
Tunnel disposal at nearshore environment is one of potential final disposal sites proposed forthe low-level waste in Taiwan. The proposed tunnel disposal facilities were located roughly
800 m away from the shoreline and below sea levels. Geosphere parameters, sea-level, andinfiltration rates are critical to radionuclide decay transports in far-field simulations. A
radionuclide decay chain of 4N+2, including Pu-238  U-234  Th-230  Ra-226, wasused as the study case. The 3-D HYDROFEOCHE5.6 numerical model was used to performsensitivity analysis of far-field radionuclide decay transport for 100,000 years withsimulations of steady flow transient transport. Hydraulic conductivities, diffusion coefficients,
and dispersivities were selected for sensitivity analysis of geosphere parameters. With 10
times higher of hydraulic conductivities than reference values, the Pu-238 concentrations arehigher and peak concentration appear earlier at shorelines. However, the Ra-226
concentrations simulated with 10 times higher/lower hydraulic conductivities are all lower
than those with reference values at shorelines. Higher hydraulic conductivities will favormigration of mother radionuclides of Pu-238, U-234, and Th-230 before seeing Ra-226 atshorelines, while 10 times lower of hydraulic conductivities will hinder migrations of allradionuclides considered in this study. Sensitivities of diffusion coefficients on radionuclide
decay transport are not significant in this study. Cases with 10 times higher/lower
dispersivities than reference values deliver higher/lower concentrations for all radionuclides atshorelines. Considering radionuclide decay transport in association with changes of geosphereparameters are much more complicated than single species simulations. Changes of sea-level
and infiltration rates are designed to investigate changes boundary conditions on radionuclide
decay transport through changes of fields. Reliable site investigations are crucial toreliabilities of far-field radionuclide transport simulations.

152

Session 3C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Seismic response of canister in buffer material under water invasion conditionby centrifuge modeling
W.Y. Hung1, J.J. Xu2, Y.C. Wu3 and M.H. Hsieh4
1Assistant Professor, Dept. of Civil Engineering, National central University, Taiwan (R.O.C)
2Master Student, Dept. of Civil Engineering, National central University, Taiwan (R.O.C)
3Supervisor, Institute of Nuclear Energy Research, Atomic Energy Council, Executive Yuan,

Taiwan (R.O.C)
4Engineer, Institute of Nuclear Energy Research, Atomic Energy Council, Executive Yuan,

Taiwan (R.O.C)
No. 300, Zhongda Rd., Zhongli District, Taoyuan City 32001, Taiwan (R.O.C.)
Key words; Canister, Seismic test, Centrifuge modeling test
Abstract
For high-level waste (HLW) disposal, the HLW is isolated from natural environment by
multi-barrier systems including engineering and natural barriers in the deep ground. The
disposal holes would be drilled in host rock and set up with canister and buffer. Then thecanister is not easily affected by natural operation, humanity activities and tectonic movementwith high stability. However, Taiwan is located at Circum - Pacific seismic zone which
earthquake occurs frequently. As a result, the effect of earthquake-induced vibration on thecanister, buffer and disposal hole is needed to be investigated.
This research assumes that the deposition is set up in granite layer at certain depth. The
disposal hole is regard as an entire intact space without deformation and cracks. Referring toKBS-3V concept and geometry, the test model is built with 1/10 scale of prototype. Throughthe centrifuge modeling, 10 g artificial acceleration field is created to simulate in situ stress ofcanister from buffer material. Seismic events are input by shaking table with 15-cycle
sinusoidal waves. During shaking, acceleration, pore water pressure and total pressure
histories are measured in order to realize the seismic behavior of canister and buffers in thedisposal hole.
Considering different states of buffer material and conditions, in total, 6 dynamic models
are conducted. In 17% water content condition, large acceleration and contact pressure ismeasured at the surface of canister. When water invades to 17% water content buffer, theacceleration and the contact pressure on the canister are both reduced. After the test, cracks
were found on the surrounding buffer materials for the model with 17% water content
condition and water invasion case.

153

Session 3C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanSTUDY ON THE APPLICABILITY OF CAST STEEL OVERPACK -EVALUATIONOF CASTING DEFECTS AND CORROSION RESISTANCEUSING FULL-SCALE PROTOTYPEYusuke Ogawa, Satoru Suzuki, Kazuhisa Yamashina and Shigeru Kubota
Nuclear Waste Management Organization of Japan (NUMO)
2F, Mita NN Bldg, 4-1-23, Shiba, Minato-ku, Tokyo, Japan
Key words; Geological disposal, Overpack, Cast steel, Non-destructive test, Corrosion,
Polarization tests
Abstractn the geological disposal concept in Japan, forged steel and cast steel can be considered as
candidate materials of carbon steel waste container, namely overpack. So far, research anddevelopment of overpack has been carried out mainly on forged steel, while that of cast steel
is still lack. We therefore studied influence of casting defects and heterogeneity of chemicalcomposition and microstructure on the long-term integrity of cast steel overpack.
In order to evaluate the influence of casting defects on structural integrity, the realistic
distribution of defects were simulated by producing the full-scale cast steel overpack.
Although small casting defects such as shrinkage cavity were founded, there was not so large
defect as compared with general cast steel products.
In order to evaluate the influence of heterogeneity of chemical composition andmicrostructure on corrosion resistance, the specimens were collected from the various
locations in the full-scale cast steel overpack. As a result, no significant difference wasobserved in chemical composition and microstructure. In addition, the anode and cathodepolarization test was carried out and corrosion rates obtained by the Tafel extrapolation
method were compared. The result shows no significant difference among the specimens.
From these results, it can be considered that the variation in corrosion resistance in a cast steel
is negligibly small.

154

Session 3C-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanSTRUCTURAL INTEGRITY ASSESSMENT OF DISPOSAL PACKAGE FOR
RADIOACTIVE WASTE FAILURE ASSESSMENTS FOR OVERPACK
USING FINITE ELEMENT ANALYSIS
Masahiro KAWAKUBO*1 and Masato KOBAYASHI*1*1 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan
Key words; Geological disposal, Disposal package, Structural integrity, Failure assessment
Abstract
In Japan, high-level radioactive wastes have been planned as geological disposal after
vitrified and encapsulated into an overpack. The carbon steel overpack is required for at least
1,000 years isolation of the vitrified waste from coming into contact with groundwater. The
overpack is cylindrical package and made of carbon steel as primary candidate material. The
basic specification of the overpack was set as the wall thickness of 190 mm and it isconsidered for corrosion allowance layer (40 mm), radiation shielding layer (150 mm) andmechanical withstanding layer (included in radiation shielding layer).
In this study, to ensure the integrity of the period required for the overpack, failure
assessments of the overpack using finite element analysis were performed for all failure mode
assumed during operation and after emplacement. From the results of failure assessment, it
was showed that the relationship between collapse load and the overpack thickness, and therelationship between welding residual stress and critical crack size of welded part. And it wasclarified that the strength of the weld joint can be sufficiently ensured by partial welding.
This research is under a grant from the Japanese Ministry of Economy, Trade and Industry(METI).

155

Session 3C-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Stress Analysis for the Canister under Earthquake Induced Fracture Shear
Displacement considering Long-term Creep Effect of Copper Shell
Yuan-Chieh Wu*, Hsi-Ting Lo and Yi-Ren Wang
Institute of Nuclear Energy Research, Atomic Energy Council
1000 Wenhua Rd. Jiaan Village, Longtan District, Taoyuan City 32546, Taiwan
Key words; ANSYS, Creep, Canister
Abstract
The design for spent nuclear fuel canister should consider three containment failuremodes, rock confining pressure, rock deformation, and corrosion. For deposal concept in hardrock such as KBS-3 in Sweden, Finland, and current Taiwan test sites, the rock fractures
surrounding deposition hole may turn into the permanent shear displacement during
earthquake; therefore, the stress analysis should be performed to evaluate seismic capacity forshear displacement. Also, for long-term disposal environment, copper canister needs toconsider the creep of copper shell, so the capability to perform analyses combing these two
phenomena should be established in final disposal research team. To perform the analyses, athree dimensional canister and buffer assembly model was built, and model test was carriedout by comparing the results with SKB data to achieve the model confidence. Several shear
displacement scenarios including different loading position and direction were carried out tounderstand the safety margin of canister design through von Mises results compared withallowable stress of copper material for canister shell and cast iron material of canister insert,
and also the long-term creep effect was considered in the analysis procedure. Based on the
results, the influence of copper creep effect will lead different stress distribution on copperand cast iron, and reduce the stress of cast iron insert; also, it demonstrates canister canmaintain integrity either considering creep from strain perspectives. The analysis capability
not only shows we can handle earthquake issue for final disposal in Taiwan where theearthquake issue always be the first natural hazard to face, but also we can reduce the canister
failure rate due to fracture shear displacement if we further consider adjusting deposition holelocation or rejecting holes when the intersecting fractures occur around deposition holes.

156

Session 4A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanPRELIMINARY STUDY OF POST-CLOSURE SAFETY ASSESSMENT
IN THE NUMO SAFETY CASE
K. Ishida, T. Hamamoto, S. Shibutani, K. Fujisaki and M. Yamada
Nuclear Waste Management Organization of Japan
Mita NN Bldg. 1-23, Shiba 4-Chome, Minato-ku, Tokyo 108-0014 Japan
Key words; post-closure safety assessment, risk informed approach, radionuclide migration
Abstract
NUMO has developed a generic safety case for the geological disposal in Japan. This safetycase provides the multiple lines of arguments and evidences to demonstrate the feasibility andthe safety of the geological disposal, which will encourage stakeholder confidence in the safe
implementation of geological disposal and will provide a basic structure for a safety casewhich will be applicable to any potential site.
The key distinguishing features of the safety assessment are “application of the risk-informed
approach”, “scenario development with comprehensiveness, traceability and transparency”,
and “realistic radionuclide migration model reflecting the repository design”.
For the purpose of optimization of protection, the disaggregated approach, as one ofrisk-informed approaches, which makes it possible to evaluate a potential impact of a future
event and its probability separately, is applied as a frame of the safety assessment. Developed
scenarios are classified into 4 categories on the basis of their probabilities; “Likely scenario”,“Less likely scenario”, “Very unlikely scenario” and “Human intrusion scenario”.
The methodology of the scenario development results from a desire to combine a bottom-up
what is called FEP-based approach and a top-down method based on the safety functions,appropriate to risk-informed assessment. The scenarios considered to be “likely” aredeveloped to be as realistic as possible, representing comprehensive current understanding of
relevant FEPs in terms of extent and rate of impact on radionuclide containment and eventual
release and transport. The methodology consisting of overall procedure and associatedtoolkits is aiming to increase traceability and transparency. For a rational repository design, it
is important that the assessment model can show the difference of the performance among
candidate designs, such as the panel layout and the disposal galleries. Then NUMO’s model isbased on the 3D particle tracking method which can effectively describes the radionuclide
migration behavior reflecting the repository design.

157

Session 4A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanASSESSMENT OF SORPTION AND DIFFUSION IN THE ROCK MATRIXIN THE NUMO SAFETY CASE
T. Hamamotoa, S. Shibutania, K. Ishidaa, K. Fujisakia, M. Yamadaa and Y. TachibaNuclear Waste Management Organization of Japan (NUMO)bJapan Atomic Energy Agency (JAEA)a

Mita NN Bldg. 1-23, Shiba 4-Chome, Minato-ku, Tokyo 108-0014 Japanb
4-33, Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 Japan

Key words;Geological disposal; Performance assessment; Radionuclide migration; Sorption;
Diffusion
Abstract
NUMO has developed a generic safety case to demonstrate the feasibility and safety ofgeological disposal of HLW and TRU waste in Japan and to provide a basic structure for thesafety case which will be applicable to any potential site. In this pre-siting safety case, theperformance assessment was carried out for the repositories tailored to site descriptive models(SDMs) developed for three representative rock groups (plutonic, Neogene sedimentary and
Pre-Neogene sedimentary rocks). Radionuclide migration parameters in rocks, i.e. distribution
coefficients and effective diffusion coefficients, were derived to allow performanceassessment for a range of scenarios. The values of these parameters were given statisticallyfrom laboratory data for certain rock types. The data were extracted from the latest sorption
and diffusion database, with interpretation based on the speciation by thermodynamic
modeling using relevant groundwater chemistry.
These parameters have some uncertainties, resulting from simplifications of sorption anddiffusion processes, simplified groundwater chemistry model, insufficient understanding of
the evolution of geosphere and the engineered barriers. In the performance assessment, theparameters assessed as the most realistic were used for a dose calculation of a “likely”
scenario. The assessment of the parameters was based on scenario analysis and numericalanalysis. The variations of the parameters to account for the uncertainties were analyzed in“less-likely” scenarios.
The paper will describe the data selection process in more detail with emphasis on
uncertainties and identify open questions and required R&Ds to resolve them.

158

Session 4A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

How will new knowledge be reflected to the management of geological disposal?
- Influence of an FEP not considered so far to Some sub-scenarios Koichi ShinCentral Research Institute of Electric Power Industry
Abiko 1646, Abiko-shi, Chiba-pref, 270-1194 Japan
Key words; Public Involvement, Stakeholder, Stepwise Decision, Safety Case, FEP,
Pressure Solution
Abstract
Since it became clear that the geological disposal of radioactive high level waste is notrealized without the understanding from the society when all siting approaches in westerncountries in around 1980s tumbled, international communities and OECD/NEA have madeefforts to attain the understanding from the society and established important principles
concerning social and ethical aspects. Some of the key words are responsibility of present
generation, safety case, stepwise decision, stakeholder involvement, preservation of RKM(records, knowledge and memory), and so on. Safety case is stipulated to include theassessment of all features, events and processes (FEPs) that could significantly influence theperformance of the disposal system. And as such, the safety case is a document to share theconsensus on the safety of a disposal system among the present generation, to serve as theground for stepwise decision, and to transmit the technical information about the disposal
system and our sincere endeavor on it to the future generation. In the stepwise decision,decisions made in the previous steps may be changed based on new scientific knowledge andothers.
Meanwhile, an FEP, pressure solution, which has not been dealt with in previous safetycase documents, has been pointed out that it can affect the geological disposal system throughtwo sub-scenarios. One is a process of permeability evolution around a disposal tunnel due
to mass transfer through pressure solution, diffusion and precipitation. Another one is aprocess of the sinking of a heavy container in the buffer due to pressure solution creep. This
report reviews the pressure solution phenomena and above mentioned two sub-scenarios,
points out that the phenomena and sub-scenarios need to be duly addressed in a safety case,and gives a humble thought to fulfilling the responsibility of present generation.

159

Session 4B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanBARRIER OF NEAR-SURFACE LILW DISPOSAL FACILITY IN KOREA :HYDRAULIC MODEL DEVELOPMENT
Mi-Jin Kwon1 , Ki-Jung Kwon2 and Jin Beak Park1
Korea Radioactive Waste Agency
1Technology institute, Korea Radioactive Waste Agency,

168 Gajeongro, Yuseong-gu, Daejeon 34129, KOREA
2Technology institute, Korea Radioactive Waste Agency,

168 Gajeongro, Yuseong-gu, Daejeon 34129, KOREA
Key words; Low and Intermediate level Radioactive Waste (LILW), vault type near-Surface
disposal, multiple engineered barrier cover system, FEFLOW
Abstract
The primary objective of engineered barrier of low and intermediate level radioactive waste
disposal facility is to limit the amount of water ingress that passes through the engineered
barrier and to limit the amount of water contacting to waste packages. From the mid of 2015,
the silo type disposal facility for low and intermediate level radioactive waste disposal started
for its operation as first stage. As second stage, the vault type near-surface disposal facility is
being prepared in Korea. The second stage facility is composed by the multiple engineered
cover barrier system so that the influences by rainfall are reduced, and thus leakage ofradioactive substances can be minimized. The cover barrier is composed of different layers
including surface layer, protective layer, drainage layer and barrier layer. In this paper,
Hydraulic model to predict the rainfall infiltration through the engineered cover barrier isprepared and assessed by two-dimensional groundwater flow model FEFLOW.

160

Session 4B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Numerical Simulation of the Thermal Performance ofa Dry Storage Cask for Spent Nuclear Fuel

Chi-Ming Lai, University Distinguished Professor
Department of Civil Engineering, National Cheng Kung University
1, University Rd., Tainan City, 701, Taiwan R. O. C.; cmlai@mail.ncku.edu.tw
Rong-Horng Chen, University Distinguished Professor
Department of Mechanical and Energy Engineering, National Chiayi University
300, Syuefu Rd., Chiayi City, 600, Taiwan R. O. C.
Chu-Tsen Liao, Assistant Professor
Department of Creative Design and Architecture, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.
Ting-Yi Wang, Undergraduate Student
Department of Civil and Environmental Engineering, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.
Heui-Yung Chang, Associate Professor
Department of Civil and Environmental Engineering, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.
Keywords: dry storage cask; spent nuclear fuel; salt particle collection; thermal
performance; heat transfer
Abstract
In this study, the heat flow characteristics of a dry storage cask were investigated and theeffect of installing a salt particle collection device on the thermal performance of the dry
storage cask was evaluated, via thermal flow experiments and a computational fluid dynamics(CFD) simulation. The results indicate that there are many inner circulations in the cask’s
flow channel (the channel width is 10 cm). They affect the channel airflow efficiency, which
in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures of
the top concrete lid and the upper locations of the concrete cask are higher than permitted by
the design specification. The installation of the salt particle collection device has no negative
impact on the thermal dissipation performance of the dry storage cask.

161

Session 4C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Influence of inherent and stress-induced anisotropy of hydraulic conductivity on the
Groundwater Flow around a rock tunnel
1Huang-Kuei Chu, 1Po-Sung Lai and 1Jia-Jyun Dong
1Graduate Institute of Applied Geology, National Central University, Taiwan (R.O.C.)

No. 300, Zhongda Rd., Zhongli District, Taoyuan City 32001, Taiwan (R.O.C.)
Key words: inherent anisotropy, stress-induced anisotropic, tunnel excavating, Oda model
Abstract
The hydraulic conductivity around the disposal tunnel is one of the key parameters for thesafety assessment of radioactive waste disposal. This study aims to explore the inherent
anisotropy (orientation of the discontinuities) and stress induced anisotropy of the hydraulicconductivity around a rock tunnel. JRC-JCS model is used to estimate the aperture of
discontinuities under stress. Based on the calculated stress field via Kirsch solution and theequivalent continuum model, the hydraulic conductivities around a circular tunnel can becalculated. The groundwater inflow of the tunnel is further evaluated via finite difference
method. The result shows that the hydraulic conductivity on the tunnel wall is about 1 ~ 2orders of magnitude larger than the one away from the tunnel (or the one of rock mass underboundary stress). The major principal hydraulic conductivity on the tunnel wall can be 6 ~ 9times larger than the minor principal value. The principle directions of the hydraulicconductivity near the tunnel wall are also significantly deviated from the tangential and radial
directions when the inherent anisotropy is considered. Groundwater flow analysis shows thatthe total head and the flow velocity are dominated by the inherent and stress induced
anisotropy of hydraulic conductivity. Surprisingly, the inflow of the tunnel is insignificantly
influenced by the spatial variation of hydraulic conductivity around the tunnel wall.

162

Session 4C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanDATA QUALIFICATION METHODOLOGYIN THE LITERATURE SURVEY STAGE
Saori Yamada1, Takanori Kunimaru1, Kunio Ota1, Stratis Vomvoris2 and Niels Giroud2
1Nuclear Waste Management Organization of Japan (NUMO)
2National Cooperative for the Disposal of Radioactive Waste (Nagra)
14-1-23 Shiba, Minato-ku, 108-0014 Tokyo, Japan
2Hardstrasse 73, CH-5430 Wettingen, Switzerland

Key words; Quality Management System, site investigation
Abstract
The site selection for a deep geological repository in Japan will be undertaken by meansof a three-stage process. The initial literature survey (LS) aims at selecting areas for the next
preliminary investigation (PI) only on the basis of data from the existing literature. The resultsof geological characterisation in the PI stage lead to the final detailed investigation.
Key to the LS stage and the evaluation process is the development of a qualified
assessment basis for each area. The challenge at this stage is that the vast amount of data in
the assessment basis originates from a wide spectrum of activities by third parties – field
investigations, laboratory tests and underground constructions – none of which has been
performed under NUMO’s Quality Management System.
NUMO has been executing a project to develop a strategy on how to evaluate, qualify andinclude data and information collected by the third parties in the assessment basis. The basic
policy for data qualification was defined and a qualification methodology was derived andapplied to the data.
The methodology consists of three steps: 1) the assessment of the level of information in agiven data record; 2) the assessment of the reliability of the reported values; and 3) the
assignment to a qualification level. As an example, for hydraulic data from borehole
investigations, the first step considers availability of geological information, information on
the location and depth of the test and the reliability of the data source. The second step
considers the test method and/or the interpretation method applied. The results of these steps
are combined to provide the final qualification level of data: (A) qualified, (B) qualified withsome degree of uncertainty or limitation in level of information, (C) high uncertainty in thedata value or metadata and (D) disqualified.

163

Session 5A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Mizunami Underground Research Laboratory Project- Achievment during Phase I/II and Important Issues for Phase III Katsuhiro Hama
Tono Geoscience Center
Japan Atomic Energy Agency
1-64, Yamanouchi, Akiyo-cho, Mizunami-city, Gifu prefecture 509-6132 JAPANKey words; MIU Project, Site Characterization Technology
Abstract
The Mizunami Underground Research Laboratory (MIU) project is being pursued by theJapan Atomic Energy Agency (JAEA) to enhance the reliability of geological disposal
technologies in the crystalline host rock (granite) at Mizunami City in Gifu Prefecture, central
Japan. The project proceeds in three overlapping phases, Surface-based investigation Phase(Phase I), Construction Phase (Phase II) and Operation Phase (III). The MIU Project has beenongoing the Phase III. During Phase I, a step-wise investigation was conducted by iterating
investigation, interpretation, and assessment, thereby understanding of geologic environment
was progressively and effectively improved with progress of investigation. An optimal
procedure from investigation to assessment was compiled as a set of geosynthesis data flow
diagram for each investigation step. During Phase II, we have evaluated adequacy oftechniques for investigation, analysis and assessment of the deep geological environment
established in the Phase I. For Phase III, three important issues were identified based on thelatest results.
(1) Development of countermeasure technologies for reducing groundwater inflow,(2) Development of modeling technologies for mass transport,(3) Development of drift backfilling technologies.
For the issue (1), post-grouting works have been applied to a gallery at 500 m depth. Three
grouting concepts were applied to the post-grouting works; a new grout material, a new
injection system, and a new post-grouting zone. For the issue (2), in-situ and laboratory masstransport experiments have been carried out. The macroscopic and microscopic observations
were carried out to understand the distribution of tracer (uranine) after the laboratory diffusion
experiment. For the issue (3), with a focus on hydraulic pressure and hydrochemical recovery
processes around underground galleries in fractured crystalline rock, the groundwater
recovery experiment has been conducted to evaluate the natural groundwater andhydrochemical recovery of the rock mass.

164

Session 5A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Virtual reality geological modeling for the Horonobe Underground Research Project
Takayuki MOTOSHIMA, Civil Engineering Dept., Nuclear Facilities Division,
TAISEI Corporation
Makito NAGO, Advanced Technology Development Dept., Taisei Technology Center,
TAISEI Corporation
Yuji Ijiri, Civil Engineering Dept., Nuclear Facilities Division, TAISEI Corporation
1-25-1 Nishi Shinjuku, Shinjuku-ku, Tokyo
Key words; Virtual Reality, Horonobe URL, geological information
Abstract
Since the high level waste disposal program is a project with high public interest, the site
selection process should be carried out through open discussion with necessary and sufficient
information. Providing only information requiring expert knowledge in interpretation, such as
conventional two-dimensional drawings (geological map and etc.) and one-dimensional
information (boring log and etc.) may not be sufficient to develop public understanding at thediscussion forum.
Virtual reality (VR) and Construction Information Management technology are considered tobe effective for collection, analysis and visualization of the geological information of theunderground, which cannot be visually confirmed easily. In this study, focusing on the effectof VR, a VR model was created by integrating geological data, photographs, andthree-dimensional profile of tunnel face acquired during the construction of the Horonobe
Underground Research Laboratory owned and operated by Japan Atomic Energy Agency. The
VR model is new and significant in that it uses measurement data of the field.
In the VR model, various information could be viewed at any point and compared. The useof the VR model allowed stakeholders to have simultaneous visual access to the geologicalinformation and to achieve a smooth communication. In addition, the model provided visuallyclear discussion materials, such as allowing access to geological information covered withconcrete even after construction. The model can also be used for planning the subsequent
construction work and research studies. The use of the VR model will not only provide
information for the discussion forum during the investigation stage but also contribute toenhance the safety and the interpretability of the disposal project by incorporating andupdating the geological information acquired during the construction and the operation of therepository into the model.

165

Session 5A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Latest rock grouting technologies under sea water in Nordic countries and Japan
Masakuni Tsuji, Civil Engineering Technology division, Shimizu Corporation;
Mitsunobu Okihara, Civil Engineering Technology division, Shimizu Corporation;
Hitoshi Nakashima, Research Institute of Technology, Shimizu Corporation;
Toshinori Sato, Horonobe Underground Research Center, JAEA;
Kazuhei Aoyagi, Horonobe Underground Research Center, JAEA.
No. 16-1, Kyobashi 2-chome, Chuo-ku, Tokyo 104-8370, JAPANKey words; geological disposal, Nordic countries, sea water condition, rock grouting,activated silica colloid, colloidal silica grout
Abstract
In Japan, the coastal area is discussed to be a more suitable area for the geologicaldisposal in the aspect of transportation in the council held by Agency for Natural Resources
and Energy. However, the engineering technology or safety assessment has not been fully
studied as an aspect for applying to the coastal area. Regarding the engineering technology,
the rock grouting for the geological disposal has been recently studied as one of the mostimportant engineering technologies. Although the cement grout has not been reported to beaffected by the sea water, the latest material for rock grouting called “colloidal silica” isknown to be affected by the saline water. The colloidal silica grout is composed of activatedsilica colloid and the mechanism of gelling and the grouting methodology under the sea water
is not yet established.
Therefore, a research project on grouting under the sea water was established, as a partof the public offered project by Agency for Natural Resources and Energy, in order to enhance
the rock grouting technology under the sea water. During this project, we studied the latest
grouting technologies for geological disposal in Nordic countries and Japan by article survey
and a grouting workshop held in Helsinki, Finland in January 2017.
As a result, it was found that the problems and the approach for countermeasures forgrouting under sea water are different between Japan, Sweden, and Finland, which are based
respectively on the mixture, on the design method, and on the controlling method. It seems
the best solution for developing this grouting technology is to combine each technology andto optimize grouting design, mixing, and controlling method. As for this Japanese project, it
was found important to develop the generic mixing method under the saline water by referring
to a previous project in Japan.

166

Session 5A-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Strategic and Technical Aspects in RD&D Program
Development for HLW Disposal System in Korea
JeongHyoun Yoon*, JeongHwan Lee, and SeungHyun Kim
Radioactive Waste Disposal R&D Dept,
Korea Radioactive Waste Agency (KORAD)
168 Gajeongro, Yuseong-gu, Daejeon 34129, KOREA
Keywords: HLW disposal, RD&D, Disposal system, URL(Underground ResearchLaboratory)
Abstract
This paper is focused on Strategic and Technical Aspects in RD&D Program for Developmentof HLW Disposal System in Korea. As the High-Level Radioactive Waste Management Basic
Plan was deliberated and voted by the final Nuclear Energy Promotion Committee in 2016,
the URL for licensing is to be constructed and operated within the permanent disposal site toensure timely management facilities. In addition to a permanent disposal facility, the research
URL(generic) is to be built separately from the management facility so that the programs forthe site selection, design, construction and operation of the disposal facility would bematerialized in a timely manner. Therefore, it is required to develop a RD&D program for the
URL operation for the demonstration of the disposal system considering the site selection
schedule and the status and characteristics of domestic high-level radioactive waste. The
preconditions for deriving RD&D programs for research URL are the preferred rocks to bedisposed of, the standard disposal system, the type of spent fuel, and the regulatory
technology guidelines. Consequently, RD&D program in associated with disposal program
should be prepared taking into accounted of strategic and technical aspects to minimize
trial-and-error. This paper shows what are strategic and technical factors in preparation of
RD&D program to implement Korean HLW disposal system and a future plan for theelaboration of the program.

167

Session 5B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Study on gas migration behavior through bentonite buffer material
Tomoyuki SHIMURA*, Shinichi TAKAHASHI*, Masanobu NISHIMURA*,
Kazumasa KOGA** and Hitoshi OWADA*** Obayashi Corporation,** Radioactive Waste Management Funding and Research Center* 2-15-2, Konan, Minato-ku, Tokyo, Japan** Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan.
Key words; Radioactive waste disposal, bentonite buffer, gas migration, performanceassessment
Abstract
Gasses generated due to the corrosion of metals contained in the waste packages andengineered buffer materials will accumulate in the engineered buffer system (EBS) of aradioactive waste repository, and may affect long-term stabilities of the EBS. To evaluate gas
migration behavior through the bentonite buffer materials in conjunction with a reasonable
design model is one of the challenges to be solved for the performance assessment of aradioactive waste repository.
A series of studies funded by the Ministry of Economy, Trade and Industry (METI) have been
conducted for more than 10 years under the leadership of Radioactive Waste Management
Funding and Research Center (RWMC) in Japan to better understand gas migration behavior
through bentonite. In this report, the outline of the framework of the gas migration study by
RWMC as well as some of the results from laboratory gas injection tests are presented.
The laboratory test results focusing on the interface of bentonite materials showed that waterflow through the interface part was blocked (self-healed) due to the swelling of the bentonite.
In addition, the gas injection test after full saturation indicated no significant differences inbreakthrough pressure compared to the intact (e.g. no interface) specimen. These findings
suggest that the interface is unlikely to become a dominant gas migration pathway in asaturated buffer.
It is expected that all of findings acquired through this decade of study are summarized andintegrated within this fiscal year (2017) taking into consideration the parallel studies on gas
scenario development and integration of coupled modelling, and contribute to the reliableadvanced repository design and performance assessment.

168

Session 5B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanMIGRATION EXPERIMENT OF VOID AIR IN BUFFER MATERIALDURING SEEPAGE
Naohiko TAKAMOTO*1, Ichizo KOBAYASHI *2, Masahiro KAWAKUBO*1,
Masataka IMAI*1 and Tomoko ISHII*1*1 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan*2 Kajima Corporation
2-19-1, Tobitakyu, Chofu city, Tokyo, 182-0036, Japan
Key words; Geological disposal, Engineered barrier, Buffer material, Seepage, Void air
Abstract
The buffer material composing engineered barrier system (EBS) for the geologicaldisposal of high-level radioactive waste in Japan plays many important roles, e.g. mechanicalprotection, hydraulic barrier, chemical buffer, controlling migration radionuclides.
Radioactive Waste Management Funding and Research Center has been focusing on
the re-saturation process of the buffer material with groundwater after the emplacement andresearching mainly on the mechanical behavior of the buffer material. The state transition of
the buffer material is the most intense during this period, and the state after the period will
impact to the continual long-term behavior. Thus, it is important to predict the state after
re-saturation process for assessing the long-term integrity of the buffer material. In this study,
we investigated the migration of void air, which is one of the phenomena affecting the statetransition of the buffer material during the re-saturation period. In order to understand thepressure increasing of void air in the swelling buffer material, one-dimensional long-term
simulations were conducted. As a result, it was obtained that the pressure increasing of void
air was mitigated by dissolution of the void air to water in parallel with water seepage into thebuffer material. Therefore, unless the influence of heat and gas from the overpack is taken
into account, it was estimated that there is little concern that the buffer material breaks bypressure increasing of void air.
This research is under a grant from the Japanese Ministry of Economy, Trade andIndustry (METI).

169

Session 5B-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanSTUDY ON PIPING AND EROSION OF BUFFER MATERIALDURING THE RE-SATURATION PERIOD
Takayuki ABE*1, Ichizo KOBAYASHI *2, Masahiro KAWAKUBO*1, Masataka IMAI*1
and Tomoko ISHII*1*1 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan*2 Kajima Corporation
2-19-1, Tobitakyu, Chofu city, Tokyo, 182-0036, Japan
Key words; Geological disposal, Engineered barrier, Buffer material, Piping, Erosion
Abstract
The state transition of buffer material (clay) which composing engineered barriersystem (EBS) for the geological disposal of high-level radioactive waste is the most intense
during the re-saturation period after the emplacement into the disposal cell. For evaluating
and predicting of long-term evolution of buffer material, it is important to grasp the behaviorof the buffer material during the re-saturation period.
The buffer material have to maintain a range of its density in order to perform its
long-term safety function. However, it is concerned that buffer material flows out by piping
and erosion during the re-saturation period.
In this study, several laboratory tests for piping and erosion of buffer material
concerned during the re-saturation period were conducted with various conditions, e.g.
material form, interface condition, liquid type, water flow and water pressure. As a result, it
was obtained knowledge on the flow out of buffer material, and engineering countermeasuresagainst it.
This research is under a grant from the Japanese Ministry of Economy, Trade andIndustry (METI).

170

Session 5B-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanNUMERICAL ANALYSIS OF INFLOW CONTROL FOR QUALITY MANAGEMENTOF BUFFER MATERIAL USING DISCRETE FRACTURE NETWORK MODEL
Kazuhiko MASUMOTO*1, Makoto NAKAJIMA*1, Hayato NONAKA*1,
Tomoko ISHII *2 ,Mayumi JO*2 ,Masataka IMAI*2 and Hiroyuki ATSUMI*2*1 Kajima Corporation
2-19-1, Tobitakyu, Chofu city, Tokyo, 182-0036, Japan*2 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan
Key words; Numerical analysis, Groundwater inflow analysis, Discrete fracture networkmodel, Geological disposal, Engineering barrier, Buffer material, Grouting
technique
Abstract
In the vertical disposal concept of high level radioactive waste, it is important to evaluate
influence of the groundwater inflow into the disposal pits against performance of buffermaterial as an engineering barrier because the inflow might cause the erosion of buffermaterial. The target of this study was to estimate the inflow control technique (e.g. grouting)around the disposal pits using numerical analysis. The model for the inflow simulation should
be a discrete fracture network model when the groundwater mainly flows along the fracturesin the rock mass, therefore, the discrete fracture network model was selected to estimate theinflow control technique in this study.
The numerical flow analysis was conducted by LT-Flow, that is, the groundwater
simulation program using pipe network structure, into which the fracture network wasconverted. Firstly the single disposal tunnel was modelled for the simulation, and secondly
the five disposal tunnels were modelled assuming the disposal panel. The results of numericalflow analysis shows that the effectiveness of grouting for the disposal pits to control the
inflow and the impact of grouting for the other disposal pits or surrounding disposal tunnel
could be evaluated quantitatively.
This research is a part of “Development of Advanced Technology for Engineering
Components of HLW Disposal” under a grant from the Japanese Ministry of Economy, Tradeand Industry (METI).

171

Session 5C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

DEVELOPMENT OF WIRELESS MONITORING SYSTEMSFOR GEOLOGICAL DISPOSAL
Kazuhiro TSUBONO*1, Masato KOBAYASHI*1, Hiromitsu YAMAKAWA*1, Akinori HASUI*1
Kazuhiko MASUMOTO*2, Tomoaki MATSUSHITA*2, Norihisa SUGAHARA*3,
Tatsuya TANAKA*4 and Chiaki NAGAI*4*1 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan*2 Kajima Corporation
2-19-1, Tobitakyu, Chofu city, Tokyo, 182-0036, Japan*3 Sakata Denki Co., Ltd.
2-17-20, Yagisawa, NishiTokyo city, Tokyo, 202-0022, Japan*4 Obayashi Corporation
Shinagawa Intercity Tower B 2-15-2, Konan, Minato-ku, Tokyo, 108-8502, Japan
Key words; Radioactive waste, Geological disposal, Engineering barrier, Wireless monitoring systems,
Wireless transmission, Low frequency electromagnetic waves, Wireless power transfer,
Magnetic resonance coupling
Abstract

In developing a monitoring program for geological disposal of radioactive waste, it isimportant that measuring instruments and cables do not affect the performance of theengineering barriers. Several studies have been conducted with underground wirelesstransmission systems; wireless transmission technology for data transmission and wirelesspower transfer technology for long-term power supply for monitoring systems.
We have studied a wireless transmission technology with low frequency electromagnetic
waves and its applicability to geological disposal monitoring systems. The applicability has
been increased by reducing the size of both transmitting and receiving devices. We have also
increased the redundancy of the wireless transmission systems by developing multistage relay
devices and transmission path changing technologies. Moreover, in cooperation with theHoronobe Underground Research Center and the Mizunami Underground ResearchLaboratory of Japan Atomic Energy Agency, verification tests for radio communication havebeen conducted, with the use of the transmitter and receiver, and we have confirmed its
technical feasibility.
Meanwhile, underground monitoring devices have some challenges such as the limited
capacity of the battery. In view of this, we have studied a wireless power transfer technology
using a magnetic resonance coupling method. Through design and trial productions oftransmission and reception coils, and experiments for power transfer efficiency and optimum
frequency, the power transfer over a buffer material has been confirmed to be viable. In theelemental tests carried out so far, we evaluated the effects on the power transfer efficiency of
cementitious structures and reinforcement steel used in plugs in the repository, demonstrating
its applicability to geological disposal facilities. In the presentation, we report the latest
development results of the underground wireless transmission technology and the wirelesspower transfer technology.
This research and development is under a grant from the Japanese Ministry of Economy,
Trade and Industry (METI).
172

Session 5C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanFEASIBILITY STUDY OF VENTILATION DESIGNFOR UNDERGROUND FACILITIES
Naoki Katsumata1), Ryoji Yahagi2), Hiromi Kurosaki2) and Shigeru Kubota1)
Nuclear Waste Management Organization of Japan (NUMO)1)
1-23, Shiba 4-chome, Minato-ku, Tokyo 108-0014 JapanSHIMIZU CORPORATION2)
2-16-1, Kyobashi, Chuo-ku, Tokyo 104-8370 Japan
Key words; operational safety, ventilation system, underground facility layout
Abstract
NUMO has developed a generic Safety Case on deep geological disposal repository in Japan.
In this report, based on the safety assessment results, the prospect will be shown that safety inthe pre-closure and post-closure phases will be ensured.
This paper provides the outline of the ventilation system design during construction andoperation. The underground facility has very long disposal tunnels and complicatedly.
Therefore, to confirm the feasibility of design for ventilation system essential to the workers
during construction and operation is difficult.
Consequently, to confirm the feasibility of design for ventilation system not only in thenormal state but also in the abnormal state is important for the operational safety.
The outbreak of the fire should be considered as one of the accident in relation to theventilation system in the underground facilities.
It is suitable countermeasures that non-flammable and/or flame-retarding materials are
utilized for all construction as measures to prevent occurrence of fire in the undergroundfacilities. On the other hand, the unusual fire event of vehicle failure or collision should beconsidered.
In spite of these preventing measures, if the incident is greatly beyond the assumption, firewould occur due to the failure or collision of transport vehicle. In case of fire in theunderground facilities, workers safety evacuation and disaster expansion prevention is what
it’s all about.
Therefore, it is indispensable for securing of operational safety to get a prospect of thefeasibility of the ventilation system not only in normal state but also in abnormal state.
As a result, by considering the ventilation system during the operation phase for the safety,
the requirements to the layout of the underground facilities was clarified in this paper.
Currently no potential site has been decided yet, these requirements can be used in futurereference.

173

Session 5C-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Establishment and application of control room habitability methodology
for Maanshan nuclear power plant
Kai-Chun Yang 1,Jong-Rong Wang 1,2, Hsiung-Chih Chen 1,2,3, Shao-Wen Chen 1,2,
Chunkuan Shih 1,2 and Wen-Sheng Hsu 3
1: Institute of Nuclear Engineering and Science, National Tsing Hua University
2 : Nuclear and New Energy Education and Research Foundation
3 : Nuclear Science and Technology Development Center. (R.O.C.)
No. 101, Section 2, Kuang-Fu Road, Hsinchu, Taiwan 30013, R.O.C.

Key words; RADTRAD, HABIT, Habitability
Abstract
In this study, we focus on the establishment and application of the control room habitability
methodology for Maanshan PWR nuclear power plant (NPP) by using RADTRAD and
HABIT codes. Therefore, there are two steps in this study. The first step is to use RADTRAD
and HABIT codes to establish the analysis methodology. The Final Safety Analysis Report(FSAR), reports, and other data for Maanshan NPP were used in this step. In addition, theRADTRAD and HABIT models were also established in this step. Second, this analysis
methodology was applied to evaluate the control room habitability for the LOCA and CO2
storage burst cases. The analysis results of RADTRAD for the LOCA were similar to FSAR
data. The HABIT results for the CO2 storage burst case were below the R.G. 1.78 failure
criteria. The above results indicate that Maanshan NPP habitability can be maintained.

174

Session 6A-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

international cooperation activities of KORAD including mid-term and long-term
strategy of international cooperation for radioactive waste management
Ou Jeong YOO and Gyeong Hwan PARKInternational Cooperation Team, Korea Radioactive Waste Agency (KORAD)
Chunghyocheong-gil 19, Gyeongju-si, Gyeongsangbuk-do, Republic of Korea,
Key words; international cooperation strategy, international cooperation activities, consistent
with radioactive waste management project, meeting, joint work, IAEA projects
Abstract
As radioactive waste management has been a challenging issue in every country using nuclear
energy, international cooperation is an important tool for competency building of eachorganization responsible for radioactive waste management and related activities. Efforts havebeen increasingly made in order to find out the solutions to safe and effective management ofradioactive waste by means from information and knowledge sharing to joint work among thenations.
Under this circumstance, KORAD established the mid-term and long-term strategy ofinternational cooperation, consistent with the roadmap and plan of radioactive waste
management, assuming that international cooperation activities should have the function of
support to the radioactive waste management project. The international strategy was set as
result of the analysis of political, economic, social and technical environment, level ofcompetency of KORAD and overseas implementing organizations, and analysis of SWOT
(Strength, Weakness, Opportunity and Threat). Based on the strategy, KORAD has carried out
various international cooperation activities, such as having cooperation meetings, launching
joint works, and initiating IAEA projects by sending the expert and providing extra budgetarycontribution to IAEA. Also, the international cooperation strategy is regularly revised toenhance its feasibility by reflecting the changes of internal and external environment
regarding radioactive waste management.
The process and result of the international cooperation strategy-making will be introduced andits implementation will be presented in detail at the paper.

175

Session 6A-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Metadata in Geological Disposal of Radioactive Waste:
The RepMet Libraries
1CAMPHOUSE, Russell C., 2 CIAMBRELLA, Massimo; 1 MCMAHON, Kevin
1 Sandia National Laboratories

Performance Assessment and Decision Analysis Department
Carlsbad, New Mexico USA
2 Nuclear Energy Agency (NEA)

Radioactive Waste Management Division
Paris, France
Key words; repository, waste package, site characterization, waste management
Abstract
OECD Nuclear Energy Agency (NEA) launched the Radioactive Waste Repository Metadata
Management initiative (RepMet) in 2014 under the auspices of the Integration Group for the Safety Case
(IGSC) technical body. RepMet’s goal is to recommend sets of metadata that can be used by national
radioactive waste repository programmes to manage their data, information, and records thereof, in a waythat is both harmonised internationally and suitable for long-term management and utilisation, e.g., insafety cases. Furthermore, the initiative that involves over ten different countries’ programmes worked on
the formulation of a consistent set of guiding principles for capturing and generating metadata,
recommending a shortlist of selected relevant standards and guidelines on international good practises.
A main intent of RepMet is to provide descriptions of conceptual data models and controlled
vocabularies, forming the so-called “libraries” in the project context, which support data and metadata use
for relevant topics within radioactive waste repository programmes. The data models are logicallystructured collections of relationships between types of data, metadata and real-world objects. They aredeveloped at the conceptual level, providing a high-level description of the logical and semanticorganisation between different library data and metadata elements. The controlled vocabularies enhance thesemantic description and searchability of the library contents, providing important tools and methodologies
for the eventual data system interoperability. Libraries are well defined and can be easily customised andimplemented by a radioactive waste management organization within their data storage system.
RepMet has developed libraries across three disciplines relevant to radioactive waste repositories. This
paper describes the three libraries and discusses their relationship to the overall RepMet initiative. The
topic libraries are reported in the table below, together with the corresponding disciplines and topics.
Disciplines

RepMet Libraries

TopicsGeoscience

Site Characterisation Library

Geological and geophysical characterisation of
the repository site.

Radioactive Waste
Management

Waste Package Library

Packaged waste and spent nuclear fuel ready forfinal disposal at the repository.

Engineering

Repository Library

Repository requirements and structure at closure.

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of
Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy’s National Nuclear Security
Administration under contract DE‐NA0003525. SAND2017‐6741 A

176

Session 6A-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Metadata in Geological Disposal of Radioactive Waste:
The RepMet Initiative
1 CIAMBRELLA, Massimo; 2 MCMAHON, Kevin; 3FEKETE, József I.
1 Nuclear Energy Agency (NEA)

Radioactive Waste Management Division
Paris, France
2 Sandia National Laboratories

Nuclear Waste Disposal Research and Analysis
Albuquerque, New Mexico USA
3Public Limited Company for Radioactive Waste Management

Budaörs, Hungary
Key words; repository, metadata, waste management
Abstract
OECD Nuclear Energy Agency (NEA) launched the Radioactive Waste Repository Metadata
Management initiative (RepMet) in 2014 under the auspices of the Integration Group for the Safety Case
(IGSC) technical body. RepMet’s goal is to recommend sets of metadata that can be used by national
radioactive waste repository programmes to manage their data, information, and records thereof, in a waythat is both harmonised internationally and suitable for long-term management and utilisation, e.g., insafety cases. Furthermore, the initiative that involves over ten different countries’ programmes worked on
the formulation of a consistent set of guiding principles for capturing and generating metadata,
recommending a shortlist of selected relevant standards and guidelines on international good practises.
National radioactive waste repository programmes require large amount of data across multiple
disciplines (e.g. geoscience, radioactive waste management, engineering) that increase as these
programmes proceed in number, type and quality for multiple reasons and goals (e.g.: site characterisation,
licensing, safety case elaboration, etc.). Considering these boundary constraints, the core idea of long-term
data management is that “data are being collected and managed for others to use them”. Next generationsof data-users have to be able to understand and access the information that the preserved data represent.
Individual scientists and research teams, as well as managers and communications specialists, need to beaware of this and document their work accordingly.
RepMet is facilitating their task by bringing about a better understanding of a key aspect of themodern data management within the field of radioactive waste disposal, namely the identification andmanagement of metadata. The initiative has analysed the metadata implementation both from the high-level
point of view (i.e. methodologies, approaches, organisation policies) and from a more technical one (i.e.
recommendation and application of selected metadata standards, data modelling techniques andimplementation of controlled dictionaries).
The RepMet initiative fills a unique and important niche in the broader programmes on data,
information and knowledge management that are conducted nationally and internationally by operators,
regulators and other relevant actors in the radioactive waste management field. This paper provides an
overview of the initiative and the status of the initiative as of the time of this writing.
Sandia National Laboratories is a multimission laboratory managed and operated by National Technology
and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc.
for the U.S. Department of Energy’s National Nuclear Security Administration under contract
DE-NA0003525. SAND2017-6740 A
177

Session 6A-4
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Spent Nuclear Fuel Final Disposal
Knowhow, Methodology and Technology Transfer
Magnus Holmqvist
SKB International AB
Box 1091, SE-101 39 Stockholm, Sweden
Key words; final disposal, design, performance assessment, safety assessment, competencedevelopment, knowhow transfer, technology transfer, KBS-3
Abstract
Taiwan Power Company (TPC) has been tasked by the government to make a Technical
Feasibility Assessment Report on Spent Fuel Final Disposal (SNFD2017) to show the status
and the technical capability of spent fuel disposal in Taiwan. The starting point has been usinga granitic rock reference case and adopting the KBS-3 concept developed in Sweden and
Finland.
The basis for the SNFD2017 report is a design and a performance and safety assessment
based on Taiwanese conditions. TPC identified the need for competence development throughthe transfer of knowhow, methodology and technology for the successful completion of the
SNFD2017 report.
SKB International has since 2015 provided support and advice to TPC based on the best
knowledge worldwide and the most up to date methodology for performance and safetyassessment. The support has been given as a series of seminars and workshops to transfer
knowhow, methodology and technology gained and developed by SKB since the mid
1970-ties. The Taiwanese team has in parallel to the seminars and workshops developed theperformance and safety assessment for disposal in Taiwan. The work, as well as the report has
been reviewed and commented by KBS-3 experts in an iterative process.
The most evident result of the cooperation has been the completion of the SNFD2017 report,behind which lies a quick learning process and subsequent utilization within TPC and its
suppliers. It is evident that TPC and its suppliers has gained a good understanding of thethinking and reasoning that lies behind assessments, decisions and designs and are hence in amuch better position to continue the work to realize a repository in Taiwan.
It can be concluded that transferring knowhow, methodology and technology in an iterative
process where transfer and support activities are mixed with actual work on a prioritized
activity such as the SNFD2017 report is successful. The SNFD2017 provided a strong driving
force and a clear end goal, which has motivated all involved to excel.

178

Session 6B-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Sulfide Corrosion by Sulphate-Reducing Bacteria in MX-80 Bentonites
Lee, Szu-Wei* and Chang, Ching-Tu
Institute of Nuclear Energy Research, Atomic Energy Council
1000 Wenhua Rd. Jiaan Village, Longtan District, Taoyuan City 32546, Taiwan
Key words; Sulphate-Reducing Bacteria, safety assessment, diffusion coefficients
Abstract
In the safety assessment of the Spent Nuclear Fuel Final Disposal Program, the corrosion
behavior of copper canister caused by the sulfide is an important issue. One of the engineered
barrier materials bentonite adjoining the canisters function is to prevent the harmful
substances from the outside to be in contact with the canisters; however, the corrosive sulfide
could exist in bentonites, and corrosion of the copper canisters could happen. One of the mostimportant sources of sulfide comes from the sulphate-reducing bacteria (SRB) group, whichlive naturally in underground water and clays. When active, the bacteria will reduce sulfate
and produce sulfide; therefore, the knowledge of the SRB sulfide production and canister
corrosion mechanisms, and relevant experiments should be developed.
One of the strong restrictions against the microbial activity is the swelling pressure, which
is dependent on compacted density of the buffer. The investigation of how well the SRB
activity is inhibited by the increasing density of MX-80 bentonite is developed, and related
works are done in progress. In addition, the diffusion experiments of sulfate are set up to
obtain diffusion coefficients in the saturated MX-80 compacted bentonite.

179Session 6B-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Adsorption of Uranium(VI) on the MX-80 Bentonite
Ying-Chieh Lin
Sinotech Engineering Consultant, Ltd..
171, Nanking E. Rd., Sec. 5, Taipei 10570, Taiwan
Key words; Adsorption, Bentonite, HYDROGEOCHEM, Uranium(VI)
Abstract
In present work, a coupled model of groundwater flow, thermal transport, and geochemicaltransport through saturated-unsaturated media, HYDROGEOCHEM, has been employing tosimulate the uranium adsorption on the MX-80 bentonite. The numerical experiments were
performed with different sodium and hydrogen concentrations. Uranium concentrations were
kept below 10-7 mol/L to avoid precipitation of amorphous uranium-hydroxide. Adsorption of
uranium on the bentonite shows significant dependency on sodium concentration, because of
the competition between uranium and sodium ions. The numerical simulations of uraniumtransport in the MX-80 bentonite had also been proposed with fresh groundwater and salt
water conditions. These simulations show that the composition of groundwater could
influence the retarded capacity of bentonite and also affect the balance of mineralogy

180

Session 6B-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Am(III)/Nd(III) INTERACTIONS WITH BORATE: EXPERIMENTALINVESTIGATIONS OF Nd(OH)3(micro cr) SOLUBILITY IN NaCl SOLUTIONS IN
EQUILIBRIUM WITH BORAX
Yongliang Xiong1)*, Leslie Kirkes1), Cassie Marrs1) and Jandi Knox1)
1)

Sandia National Laboratories (SNL), Carlsbad Programs Group, 4100 National Parks Highway,
Carlsbad, NM 88220, USA

Key words; Nuclear Waste Management, Waste Isolation Pilot Plant (WIPP), Salt Formation, Rare
Earth Elements

Abstract
In previous studies on Am(III)/Nd(III) interactions with borate, the borate
concentrations were kept at a relatively low level, up to ~0.16 mol•kg–1 as the highest. Such
concentrations of borate are not saturated, in terms of borate, with borate-bearing phases suchas borax (Na2B4O7•10H2O) and tincalconite (Na2B4O7•5H2O) that could be present ingeological repositories. For instance, tincalconite has been observed as a corrosion productfor boroslicate glass for HLW under the repository conditions in China.
In this study, we examine the Nd(III) interactions with borate in equilibrium withborax in a wide range of NaCl solutions at 25oC.
In our experiments, thesolubility-controlling phase for Nd(III) is Nd(OH)3(micro cr).
In synthesis ofNd(OH)3(micro cr), we followed the procedure from literature. First, high purity Nd2O3 wasloaded into Paar® reaction vessels with deaerated DI water, and then the reaction vessels weresealed in a glovebox under a positive pressure of an inert gas. The reaction vessels were
removed from the glovebox and placed into a muffle furnace. Nd(OH)3(micro cr) formed
when the high purity Nd2O3 reacted with the deaerated DI water at 200oC in Paar® reaction
vessels for a period of two weeks. Second, following the synthesis step, the reaction vessels
were then transferred back into the glovebox, and were opened for drying in an atmosphere ofinert gas.
This synthesis method assures the complete conversion of Nd2O3 to
Nd(OH)3(micro cr), as demonstrated by XRD and SEM-EDS characterizations. Note, thedeaerated DI water used in the synthesis was prepared by vigorously bubbling high purity Ar
or N2 gas through the DI water for a minimum of 30 minutes in the glovebox. Thisdeaeration process was intended to remove any dissolved CO2 and therefore ensure thesynthesis process was not contaminated by carbonate.

181

Session 6C-1
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Microbial DNA; A brand-new tracer of groundwater flow
1Ayumi Sugiyama, 2Kenji Kato, 2Kazuyo Nagaosa, 1Tetsuo Ibara, 1Kazuyoshi Takenobu,

and 3Atsunao Marui
1Asano Taiseikiso Engineering Co., Ltd., 2-8-7, Kitaueno, Taitou-ku, 110-0014, Tokyo, Japan,
2Department of Geoscience, Faculty of Sciences, Shizuoka University
3Geological Survey of Japan,

National Institute of Advanced Industrial Science and Technology (AIST)
Key words; Microbial DNA, groundwater flow path, Brand-new tracer
Abstract
To understanding better how the groundwater flow passes and its influence on the waterchemistry we employ a brand-new method to analyze microbes with abundance andcommunity constituents in groundwater. In the previous studies, estimation of transit time andgroundwater flow path by groundwater flow analysis, estimation of averaged residence timeof groundwater by chemical analysis such as various radioactive elements and stable isotopes
have been widely used in the study of hydrology. These methods show an averaged value ofchemistry of the examined water which was blended by various water with different sources
and flow passes in subsurface environment. In order to improve the reliability of these results,
we focus on Microbial DNA. Particles represent the place from which they transported.
Microbes which are living organisms, express the environment appropriate for their growth.
Interaction among rocks - groundwater - microorganisms is studied to evaluate
biogeochemical influences on natural barrier of geological disposal facility forhigh-level-radioactive wastes. This study shows the possibility to reveal the groundwater
history such as groundwater flow and the environment from which groundwater flow throughanalyzing the density of prokaryotes and microbial community structure analysis.

182Session 6C-2
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Seismic Analysis for the Deposition Tunnel in Fractured Rock by 3DEC
Meng-Hsiu Hsieh and Yuan-Chieh Wu
Institute of Nuclear Energy Research, Atomic Energy Council
No.1000 Wenhua Rd, Jiaan Village, Longtan District, Taoyuan City 32546, Taiwan
Key words; fracture rock, seismic analysis, 3DEC
Abstract
In this study, the seismic stability analysis of the deposition tunnel was carried out during
the operation of the nuclear fountains in the outlying island crystalline rock test area. Even in
the deep intact granite rock mass, there still are some fractures, and these fractures might be
the key to the instability of the tunnel during the earthquake. The existed cracks in the deep
integrity of the granite body and rock cracks occurred in the tunnel during the earthquake may
be the key of instability. In order to ensure the safety of the deposition tunnel, the seismic
analysis of fractured rock tunnel should be carried out. In this study, the 3DEC was used toanalyze the stability after the excavation of tunnel and the intact rock and fractured rockduring the earthquake. The results show that the safety factor after tunnel excavation is
directly affected by the fractures, while the safety factor is higher in the intact rock. The
results of the seismic analysis show that the impact of the earthquake on the safety factor isrelatively small. But the existence of the location of the fractures may still have an impact onstability, and still need to do further study in the future.

183

Session 6C-3
6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanGROUNDWATER FLOW ANALYSIS FOR EVALUATING FACTORS ON WATER
INFLOW TO THE FACILITY DURING THE OPERATION PERIOD
Masataka IMAI*1, Minoru EMORI*1, Tomoko ISHII*1, Masahiro KAWAKUBO*1,
Kazuhiro TSUBONO*1, Masayoshi YAMAURA*2, Tsuneyuki MAEMURA*2
and Eiko SUGIMOTO*2*1 Radioactive Waste Management Funding and Research Center
Nichirei Akashicho Bldg. 12F, 6-4, Akashicho, Chuo-Ku, Tokyo, 104-0044, Japan*2 Dia Consultants Co., Ltd.
2-272-3 Yoshino-cho, Kita-ku, Saitama-shi, Saitama 331-0811, Japan
Key words; Geological disposal, Disposal system, Groundwater flow analysis,
Density-dependent flow
Abstract
In order to ensure the long-term safety of geological disposal system of high level
radioactive waste, it is necessary that each component of the disposal system should bemaintained its initial performance required from the long-term safety assessment taking into
consideration of disturbances caused by construction and operation of the undergroundfacility.
This research is aiming to develop an evaluation method to predict evolution of its state
and/or performance of each component focusing on the disturbances during construction andoperation. For this evaluation, unsteady groundwater analysis is necessary to consider thehydraulic characteristics of the site and the staged shape transition of its underground facility
during construction and operation. For this developing purpose, the coastal site was examined
as a test case.
First, numerical simulations of density-dependent groundwater flow were performed byusing a simple two-dimensional vertical model taking into consideration of the operation
process, and major factors that may affect the underground hydraulic environment weresummerized. Then, we conducted three-dimensional groundwater analysis on different
regional size to evaluate the influence of the operation procedure on water inflow to thefacility.
As a result, it’s confirmed that the hydrostatic pressure of sea water was a dominant factor
on water inflow to the facility. It was also possible to understand the trend of changes in water
inflow to the disposal tunnel according to the operation process.
This research is under a grant from the Japanese Ministry of Economy, Trade and Industry(METI).

184