Kyoto University
Research Reactor InstituteInternational Symposium on
Nuclear Back-end Issues and the Roleof Nuclear Transmutation Technology
after the accident of TEPCO's
Fukushima Daiichi Nuclear Power
Stations
28 November 2013
The Shiran Hall, Kyoto University
KURRI Research Program for Scientific
Basis of Nuclear Safety

Programme&
Abstracts

Preface

The nuclear accident at the TEPCO’s Fukushima Daiichi NPP in March 2011, which
followed an earthquake and a tsunami, brought serious impacts on society. This accident
inevitably underlined the necessity of establishing new and comprehensive scientificresearch for promoting nuclear safety. Facing this situation, the Kyoto University Research
Reactor Institute (KURRI) developed a new research program called the “KUR Research
Program for Scientific Basis of Nuclear Safety” from 2012. In this program, it is planned
to hold an annual series of international symposia along with identified research activities
for promoting nuclear safety.
The second annual symposium in this series deals with nuclear back-end issues and therole of nuclear transmutation technology after the accident of the TEPCO's Fukushima
Daiichi NPP, following the first one for the radiological effect of the accident on the public.
As well as on the impacts of the accident, the accident has called us to focus our attention
on a large amount of spent nuclear fuels stored in NPPs. In fact, public anxiety regarding
the treatment and disposal of high-level radioactive wastes that require long-term control is
now growing, while the government policy on the back-end of nuclear fuel cycle isunpredictable in the aftermath of the accident. The issues may not be simply technical, but
are critically important not only for settling the accident but also for pursuing nuclear
energy production in the world. It is thus needed to address the current status of the backend issues and to discuss the future direction of research and development on radioactive
waste treatment and disposal.
Approximately 9 invited lectures and 54 scientific papers will be presented in the
symposium. Almost all the presentations have been written as original scientific papers,
peer reviewed by specialists of the relevant research fields, and included in the proceedings
booklet. On behalf of all the participants of the symposium, I sincerely hope that this
symposium and its proceedings will contribute to the reduction of public anxiety and will
promote further progress in the research on nuclear safety.

Hirotake Moriyama

Director, Kyoto University Research Reactor Institute
Chair of the organizing committeeHost Organization
Kyoto University Research Reactor Institute (KURRI)

Support organization
Atomic Energy Society of Japan

Organizing Committee
Chair: Prof. Hirotake Moriyama (Director, KURRI)
Dr. Shinichi Nakayama (JAEA)
Prof. Masayuki Igashira (Tokyo Institute of Technology)
Prof. Toshikazu Takeda (Fukui University)
Prof. Kenji Ishibashi (Kyushu University)
Prof. Yoshiaki Kiyanagi (Nagoya University)
Prof. Tsuyoshi Misawa (KURRI)
Prof. Yoshiharu Mori (KURRI)
Prof. Ken Nakajima (KURRI)
Prof. Sentaro Takahashi (KURRI)
Prof. Hajimu Yamana (KURRI)

International Advisor
Prof. Hajimu Yamana (KURRI)
Dr. Hiroyuki Oigawa （JAEA)

Program Committee
Chair: Prof. Ken Nakajima (KURRI)
Prof. Jun-ichi Katakura (Nagaoka University of Technology)
Prof. Tomohiko Iwasaki (Tohoku University)
Prof. Akio Yamamoto (Nagoya University)
Prof. Kengo Hashimoto (Kinki University)
Prof. Ken Kurosaki (Osaka University)
Dr. Kazufumi Tsujimoto (JAEA)
Prof. Kazuya Idemitsu (Kyushu University)
Prof. Yoshihiro Ishi (KURRI)
Prof. Yasushi Saito (KURRI)
Prof. Cheol Ho Pyeon (KURRI)
Prof. Toshiyuki Fujii (KURRI)

Conference Secretariat
Chair: Prof. Toshihiro Yamamoto (KURRI)
Prof. Tadafumi Sano (KURRI)
Prof. Junichi Hori (KURRI)
Prof. Daisuke Ito (KURRI)
Ms. Maki Nakatani (KURRI)

1

Program
9:00-10:00 Registration, Poster Reception
10:00-10:15 Opening Remarks

Prof. Hirotake Moriyama, Director of KURRI

10:15-10:45 Decommissioning Activity at Fukushima Daiichi NPS
Speaker : Mr. Hideki Masui (TEPCO)
Chair: Prof. Ken Nakajima (KURRI)
10:45-11:15 Issue of the High-Level Radioactive Waste
Speaker : Prof. Kenji Yamaji (RITE)
Chair: Prof. Ken Nakajima (KURRI)
11:15-11:45 Roles of Nuclear Fuel Cycle Technologies on Geological Disposal
Speaker : Prof. Joonhong Ahn (University of California, Berkeley)
Chair: Prof. Yoshiaki Kiyanagi (Nagoya University)
11:45-12:15 Expectation to Nuclear Transmutation
Speaker : Dr. Akito Arima (Musashi Academy)
Chair: Prof. Yoshiaki Kiyanagi (Nagoya University)
12:15-14:00 Lunch and poster session
14:00-14:30 OECD / NEA Activities Related to the Nuclear Fuel Cycle
Speaker : Dr. Thierry Dujardin (OECD)
Chair: Dr. Kazufumi Tsujimoto (JAEA)
14:30-15:00 Contribution of the European Commission to a European Strategy for HLWManagement through Partitionning & Transmutation- Presentation of MYRRHA and its in the European P&T strategy Speaker : Prof. Hamid Aït Abderrahim (SCK･CEN)
Chair: Prof. Hironobu Unesaki (KURRI)
15:00-15:30 ADS Study in JAEA Speaker : Dr. Hiroyuki Oigawa (JAEA)
Chair: Prof. Hironobu Unesaki (KURRI)
15:30-16:00 Coffee break and poster session
16:00-16:30 Accelerator-Driven System (ADS) Study in Kyoto University Research Reactor
Institute (KURRI) Speaker : Prof. Cheol Ho Pyeon (KURRI)
Chair: Prof. Jun-ichi Katakura (Nagaoka University of Technology)
16:30-17:00 Nuclear Data Study for Nuclear Transmutation
Speaker : Prof. Masayuki Igashira (Tokyo Institute of Technology)
Chair: Prof. Jun-ichi Katakura (Nagaoka University of Technology)
17:00-18:00 Panel Discussion

Chair: Prof. Ken Nakajima

18:00-18:15 Closing Remarks

Prof. Hajimu Yamana

18:15-19:40 Reception2

Poster sessions
Topics 1-a Nuclear transmutation - Reactor system –1 Lu ZHANG, Kun FENG, Yucui GAO, Deliang FAN, Yongwei FANG
Physical Analysis of LBE Spallation Target for CIADS
2 Kun FENG, Lu ZHANG, Deliang FAN, Zhuang WU, Yongwei YANG
Analysis of Thermal-hydraulics for the LBE Spallation Target of ADS
3 Takanori KITADA, Vu Thanh MAI, Noboru DOBUCHI
Study on Neutron Spectrum of Pulsed Neutron Reactor
4 Toshihiro YAMAMOTO
Theory of power spectral density and Feynman-a method in Accelerator Driven System and theirhigher order mode effects
5 W.F.G. VAN ROOIJEN
Neutron requirement for transmutation of MA isotopes
6 Tetsushi HINO, Masaya OHTSUKA, Renzo TAKEDA, Junichi MIWA, Kumiaki MORIYA
Application of the Resource-renewable Boiling Water Reactor for TRU Management and LongTerm Energy Supply
7 Takanori SUGAWARA, Kenji NISHIHARA, Kazufumi TSUJIMOTO, Toshinobu SASA
Hironari OBAYASHI, Yuji KURATA, Hiroyuki OIGAWA
Investigations of Lead-Bismuth cooled Accelerator-Driven System
8 Toshinobu SASA, Hayanori TAKEI, Shigeru SAITO, Hironari OBAYASHI, Kenji NISHIHARA,
Takanori SUGAWARA, Hiroki IWAMOTO, Kazufumi TSUJIMOTO, Hiroyuki OIGAWA
J-PARC Transmutation Experimental Facility
9 Kyoko ISHII, Mitsuaki YAMAOKA, Yasuyuki MORIKI, Takashi OOMORI, Yasushi TSUBOI, Kazuo
ARIE, Masatoshi KAWASHIMA
Development of Uranium-Free TRU Metallic Fuel Fast Reactor Core with Enhanced Doppler
Reactivity Feedback
10 Akihiro SASAHARA, Tetsuo MATSUMURA
Chemical Isotopic Analysis and Computational Evaluation of Nuclide Composition in PWR-UO2
and MOX Spent Fuels Using JENDL, ENDF/B, JEF, and JEFF
11 Kenji KONASHI, Tsugio YOKOYAMA
Enhancement of Transmutation of Minor Actinides by Hydride Target
12 Hiroki SONO , Kotaro TONOIKE , Kazuhiko IZAWA , Takashi KIDA , Fuyumi KOBAYASHI ,Masato SUMIYA , Hiroyuki FUKAYA , Miki UMEDA , Kazuhiko OGAWA, Yoshinori MIYOSHI
Project to Modify the STACY Critical Facility for Experimental Study on Fuel Debris Criticality
Control
13 Naoto AIZAWA, Tomohiko Iwasaki, Tatsuro TAKANI, Takashi TANI, Shunsuke KANEMOCHI,
Motomu SUZUKI, Fumito KUBO, Yasuaki WATANABEADS study in Tohoku University
314 Toshikazu TAKEDA, Koji FUJIMURA
Method Development for Calculating Minor Actinide Transmutation in a Fast Reactor
15 Daisuke ITO, Kazuki HASE, and Yasushi SAITOHeat transfer study for ADS solid target-surface wettability and its effect on a boiling heat transfer16 Gen ARIYOSHI, Daisuke ITO, and Yasushi SAITO
Experimental study of flow structure and turbulent characteristics in lead-bismuth two-phase flow
17 Kyoko Mukaida, Naoto Yasumatsu, Masanori Heta, Akira Ohtaki, Hiroki Shiotani, Kiyoshi Ono,
Masaru Hirata
Scenario study for the transition from thermal reactors to fast reactors in the world
18 Tetsuya MOURI, Taira HAZAMA, Hiroshi NISHI
Current Status and Future Prospect of MA Irradiation Tests in Monju
19 Shigeo OHKI, Tsutomu OKUBO, Tomoyuki ABE
Study on Actinide Burning by Fast Reactor
20 Seung-Woo HONG, Sang-In BAK, Jong-Seo CHAI, J. P. Curbelo, Yacine KADI, Claudio TENREIRO
Neutronic analysis and transmutation performance of Th-based Molten Salt Fuels
Topics 1-b Nuclear transmutation - Fuel cycle –1 Kenji NISHIHARA, Kazufumi TSUJIMOTO, Hiroyuki OIGAWA
Transmutation Scenarios after Closing Nuclear Power Plants
2

Kotaro TONOIKE, Hiroki SONO, Miki UMEDA, Yuichi YAMANE, Teruhiko KUGO,
Kenya SUYAMA
Principle Options of Fuel Debris Criticality Control in Fukushima Daiichi Reactors

3 Seung-Woo HONG, Sang-In BAK, Masoud BEHZAD, Jong-Seo CHAI, Yacine KADI,
Vijay K. MANCHANDA, Tae-Sun PARK, Claudio TENREIRO, Chirag K. VYASADS Research Activities in Sungkyunkwan University
Topics 1-c Nuclear transmutation - Material science –1 Shizuka TAKAI, Kouichi HAGINO
Nuclear transmutation of long-lived nuclides with laser Compton scattering : Quantitative analysis
by theoretical approach
2 Kento YAMAMOTO, Keisuke OKUMURA, Kensuke KOJIMA, Tsutomu OKAMOTO
Sensitivity analyses of initial compositions and cross sections for activation products ofin-core structure materials
3 Hiroaki MUTA, Toshiaki KAWANO, Yuji OHISHI, Ken KUROSAKI, Shinsuke YAMANAKA
Thermophysical properties of thorium oxide44 Toshiaki KAWANO, Hiroaki MUTA, Yuji OHISHI, Ken KUROSAKI, Shinsuke YAMANAKA
Characterization and thermophysical properties of RE2Zr2O7 and Nd2Ce2O7 precipitate in ThO2based fuel5 Minoru TAKAHASHI, Yun GAO, Marion GUIHOT, Asril PRAMUTADI
Overview of Studies on Lead-bismuth Technology in Tokyo Institute of Technology
6 Hironari OBAYASHI, Hayanori TAKEI, Hiroki IWAMOTO, Toshinobu SASA
Evaluation of structural integrity of beam window for TEF target by thermal/fluid-structureinteraction analysis
7 K. HIROSE, T. OHTSUKI, Y. SHIBASAKI, N. IWASA, J. HORI, S. SEKIMOTO, K. TAKAMIYA,
H. YASHIMA, K. NISHIO, and Y. KIYANAGI
Fission cross-section measurement of minor actinides using a lead slowing-down neutron
spectrometer KULS at KUR
8 Tsuyoshi NISHI, Yasuo ARAI, Masahide TAKANO, Hirokazu HAYASHI, Masaki KURATA
Development of TRU nitride database for designing ADS fuel9 M. KOIZUMI, F. KITATANI, H. TSUCHIYA, H. HARADA, J. TAKAMINE, M. KURETA, H.
IIMURA, M. SEYA, B. BECKER, S. KOPECKY, W. MONDELAERS, P. SCHILLEBEECKXRecent progress in research and development on the neutron resonance densitometry (NRD) for
quantification of nuclear materials in particle-like debris
10 Marion GUIHOT, Minoru TAKAHASHI
Status of Studies and Future Subjects of Material Corrosion for Lead-bismuth Cooling System
11 Shigeru SAITO, Kenji KIKUCHI, Dai HAMAGUCHI, Shinya ENDO, Kouji USAMI, Naotoshi
SAKURABA, Hiromitsu MIYAI, Katsutoshi ONO, Hiroki MATSUI, Masayoshi KAWAI, Yong DAI
Mechanical properties of beam window materials for ADS irradiated in a spallation environment
12 Jun-ichi HORI, Tadafumi SANO, Yoshiyuki TAKAHASHI, Hironobu UNESAKI, Ken NAKAJIMA
Development of non-destructive assay to fuel debris of Fukushima Daiichi NPP (1) Experimental
validation for the application of a self-indication method
13 Tadafumi SANO, Jun-ichi HORI, Yoshiyuki TAKAHASHI, Hironobu UNESAKI, Ken NAKAJIMA
Development of non-destructive assay to fuel debris of Fukushima Daiichi NPP (2) Numericalvalidation for the application of a self-indication method
14 Shoji NAKAMURA
Precise Measurements of Neutron Capture Cross-Sections for LLFPs and MAs
15 Kenya SUYAMA, Gunzo UCHIYAMA, Hiroyuki FUKAYA, Miki UMEDA, Toru YAMAMOTO,
Motomu SUZUKI
Development of the Method to Assay Hardly Measurable Elements in Spent Nuclear Fuel andApplication to BWR 9×9 fuel16 M. Yamawaki, T. Terai, T. Koyama, Y.Arita1, Y. Sekiguchi, K. Uozumi and M.Kinoshita
Evaluation of volatile FP elements behavior for severe accident analysis of molten salt reactor

5Topics 2-a Back-end cycle issues - Radioactive waste treatment/disposal 1 Nobuyoshi ISHII, Shinichi OGIYAMA, Shinji SAKURAI, Keiko TAGAMI, Shigeo UCHIDA
Environmental parameters for assessing the behavior of radiocarbon in the paddy soil-to-rice
plant system
2 Jian ZHENG, Tagami KEIKO, Wenting BU, and Shigeo UCHIDA
Development of ICP-MS based analytical method for the determination of radioactive Cs
isotopes
3 Asako SHIMADA, Mayumi Ozawa, Yutaka KAMEO, Takuyo YASUMATSU, Koji NEBASHI,
Takuya NIIYAMA, Shuhei SEKI, Masatoshi KAJIO, Kuniaki TAKAHASHI
Development of rapid analytical method of 129I in the contaminated water and woods arising at
Fukushima Daiichi Nuclear Power Station
4 Shigeo UCHIDA, Keiko TAGAMI
Determination of low-level technetium-99 in the environmental samples by inductively coupledplasma mass spectrometry
5 Guosheng YANG, Jian ZHENG, Keiko TAGAMI, Shigeo UCHIDA
A simple and rapid method for separation and preconcentration of Ra in water samples
6 Hiromi TANABE, Kuniyoshi HOSHINO
Consideration on treatment and disposal of secondary wastes generated from treatment ofcontaminated water
7 Takayuki SASAKI, Akira KIRISHIMA, Yuu TAKENO, Kohei FUKUDA, Nobuaki SATO
A simulation of Sr leaching in aqueous solution mixed with seawater and preparation of U/Zr
oxides at high temperatures
8 Yoko FUJIKAWA, Hiroaki OZAKI, Hiroshi TSUNO, Pengfei WEI, Aiichiro FUJINAGA, Ryouhei
TAKANAMI, Shogo TANIGUCHI, Shojiro KIMURA
Volume reduction of municipal solid wastes contaminated with radioactive cesium byferrocyanide coprecipitation technique
9 Takao TSUBOYAA Prescription toward to the Long-term Isolation of High Level Radioactive Waste in Japan
10 Shinji YOSHIKAWA, Kouichi USHIRODA, Michiaki OHKUBO, Yutaka HISHIKI,
Kiyoshi HOMMA
MEXT-Sponsored Research Activities for Environmental Load Reduction
11 Masahiko SAITO, Satoru SUZUKI, Isamu SETO
NUMO’s activities after the Great East Japan earthquake
12 Akira KITAMURA, Yukio TACHI
Migration Parameters and their Evaluation and Estimation Methodologies of Safety-relevantRadionuclides for Performance Assessment of Japanese Geological Disposal of HLW and TRU
Waste

6

13 Takamitsu ISHIDERA, Seiichi KUROSAWA, Masanori HAYASHI, Yasuyuki SUZUKI
Evaluation of Distribution Coefficients for Radionuclides Sorption on Bentonite Colloid
14 Yuya Takahashi, Hitoshi Nakamura, Akira Yamada, Koji Mizuguchi, Reiko Fujita
The treatment process of simulated debris from a severe accident using molten salt system
Topics 2-b Back-end cycle issues - Public acceptance 1 Akemi YOSHIDA
Considering Geological Disposal Program of High-level Radioactive Waste through Classroom
Debate7

Abstracts

Decommissioning Activity at Fukushima Daiichi NPS
Hideki Masui
Safety Research Manager, Nuclear Asset Management Dept.
Tokyo Electric Power Company

The presentation will provide the recovery works from the accident which occurred at Fukushima
Daiichi Nuclear Power Station on March 11, 2011.
First, the focus is given to the countermeasures conducted until the cold shutdown state wasaccomplished in December 2011, which was the condition where the temperature within the reactor was
below 100 degrees Celsius and the radiation level at the station premise was significantly suppressed.
For instance, alternative circulating cooling systems, contaminated water processing facilities andprimary containment vessels (PCVs) gas control system were established to stabilize the plants.
Then, the roadmap toward the decommissioning for Units 1 to 4 will be introduced. The roadmap
divides the term of decommissioning into the following three phases.
1. The beginning of fuel removal from the SFPs
2. The beginning of fuel debris removal
3. The end of decommission activities
To achieve this roadmap, a variety of activities is ongoing. The examples include:- The investigations on internal conditions of PCVs are imperative before fuel debris removal fromdamaged reactors. The visual inspection, radiation dosimetry and measurement of accumulated
water level inside PCVs for Units 1 and 2 are being conducted by inserting an industrial endoscope.
- It was planned to install building covers for Units 1, 3 and 4 where the upper parts of the reactor
buildings were blown during the accident. In preparation for this, dismantling of rubbles from top
floors of reactor buildings is under way in Unit 3, and has been completed in Unit 4.
Above all, the inventory control (contaminated water management) is the most pressing issue. Since
the ground water volume flowing into reactor buildings is significant, the volume of contaminated water
increases steadily. It causes a big challenge to storage. To cope with it, the measures are being
considered and tested, such as decontaminating the incoming water by using multi-nuclide removal
equipment, or constructing a groundwater bypass from the upstream to the sea so that incoming flow
can be reduced. Special attention will be given to the latest situation and countermeasures of water leakincidents.
In addition to these activities, R&D works toward the decommissioning is now being developed.
This presentation will be finally summarized with the issues we are facing and the countermeasures weare considering for future recovery activities.
Keywords
Fukushima, decommissioning, contaminated water8

Issue of the High-Level Radioactive Waste
Kenji Yamaji
Research Institute of Innovative Technology for the Earth (RITE)In relation to disposal of high-level radioactive waste (HLW), related government organizations andthe Nuclear Waste Management Organization of Japan (NUMO) have been making efforts aimed at
beginning a literature survey, a first step of HLW disposal according to fundamental policies and final
disposal plan based on the “Designated Radioactive Waste Final Disposal Act”. However, a difficult
situation continues in which responses from municipalities, which are necessary for beginning literature
survey, are not being made.
In September 2010 the Science Council of Japan (SCJ) received a deliberation request from the
Chairman of the Japan Atomic Energy Commission, and SCJ formed a Review Committee for Disposalof High-Level Radioactive Waste. The Review Committee made a Reply on Disposal of High-Level
Radioactive Waste in September 2012. In this paper, an outline of the current HLW disposal policy andthe contents of the Reply are introduced.
Until now, the Final Disposal Act formulated in June 2000 has been the policy framework for Japan’s
HLW problems, and NUMO was established in October 2000 as an organization to implement thosepolicies.
The following is an outline of policies stipulated based on the Final Disposal Act.
•
Conduct final disposal by burying it, after confirming safety, in a geological layer at least 300
meters underground.
•
The Ministry of Economy, Trade and Industry (METI) is in charge of supervising and regulating
disposal projects, and the Minister of METI will stipulate a fundamental policy on final disposal
based on the law and will, every five years, stipulate a 10-year plan for final disposal.
•
Selection of a place for final disposal of high-level radioactive waste will be conducted throughthe following 3-stage procedures: (1) selection of a “preliminary investigation area,” (2) selectionof a “detailed investigation area,” and (3) selection of a “site for construction of final disposal
facilities.”•
NUMO is established as the main party for conducting geological disposal.
•
When selecting preliminary investigation areas, NUMO will accept offers from all cities, towns,
and villages in Japan.
•
A party that constructs a nuclear reactor for generating power must contribute the expenses
required for final disposal of high-level radioactive waste. The Radioactive Waste Management
Funding and Research Center will be in charge of managing those funds as a reserve fund, andwill provide the funds to NUMO after receiving approval from the Minister of METI.
•
If necessary, the law will be reconsidered when 10 years have passed since it was enacted.
On the other hand, the Review Committee of SCJ pointed out the following six proposals in order tosearch for a path toward consensus formation: 1) Fundamental reconsideration of policies related todisposal of high-level radioactive waste; 2) Awareness of the limits of scientific and technical abilities
and securing scientific autonomy; 3) Rebuilding a policy framework centered on temporal safe storageand management of the total amount of HLW; 4) Necessity of persuasive policy decision procedures forfairness of burdens; 5) Necessity of multiple-stage consensus formation by establishing opportunities fordebate; and 6) Awareness that long-term persistent undertakings are necessary for problem resolution.
Keywords
High-level radioactive waste, temporal safe storage, management of total amount9Roles of Nuclear Fuel Cycle Technologies on Geological Disposal
Joonhong AHN
Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA 94720 USA
Spent fuel management has been a challenge in any nuclear country. Even for countries withadvanced technological capability and diverse geological and societal conditions, finding an actual site
for a geological repository has proven very difficult politically and societally. Almost all countries that
tried to advance from their feasibility study stage to siting stage had a major setback, such as Canada’s
Seaborne Commission in 1998, President Obama’s decision in 2009 for the Yucca Mountain Project in
the United States, and most recently the recommendation made by the Science Council’s committee tothe Atomic Energy Commission of Japan.
By learning from numerous failures and a few successful cases, such as Finland’s Olukiluoto,
Sweden’s Forsmark, and US WIPP for TRU wastes, the concept of geological disposal has evolved
significantly. In 1980s and earlier, the concept of geological disposal was just about a mined geologicalrepository. Through discussions in the international community during the last decade, the contemporary
geological disposal concept has expanded significantly. It now includes wide variety of options, ranging
from treatment and interim storage prior to final geological disposal, resulting in great flexibility toaccommodate people’s values, which also range widely and evolve with time. Variety and flexibility are
crucial to implement reversibility in public-participatory, adaptive staged approach for decision-making
process.
To achieve a breakthrough for deadlocked situation in spent nuclear fuel management, historicallyvarious concepts of partitioning and transmutation (P&T) have been proposed and developed. While theeffects of P&T on reduction of toxicity of long-lived minor actinides (MA) have been well recognized
and studied, it has been pointed out mainly by the repository community that such reduction of MA
would not directly affect the safety of a geological repository. This conflict of views on effects of P&T
has resulted from the fact that the repository performance is measured and regulated solely by theradiological safety, i.e., annual dose for the public by intake of radionuclides that are released from therepository.
As the understanding about repository mechanisms and importance of the flexibility in decisionmaking process backed by variety of options has achieved remarkable progress, new roles of P&T seemto have emerged, which can be observed from two points. One is the convergence of the spent fuelmanagement. As far as the nuclear power utilization continues, stockpile of nuclear materials after theirutilization will increase. This ever-increasing, diverging tendency makes the safety and safeguards ofspent fuel management worse, and thus affected the public perception adversely. P&T can contribute toenhance convergence of nuclear material stockpile under sustainable nuclear-power utilization. The
second point is the variety and flexibility. While the geological repository has moderate variety intechnological options, they are limited basically by adjusting the time for cooling and choosing thelocation of the interim storage and repository. P&T enhances the technological variety significantly byaugmenting options of waste forms and radionuclide inventory in waste forms, which would also open a
possibility of making a repository more compact and robust.
Keywords
Convergence, Reversibility, Geological disposal

10

Expectation to Nuclear Transmutation
Akito Arima
Musashi Academy

1. 一次エネルギー、電力の需要はますます伸びる
The demand for primary energy and electricity is increasing year by year.
2. 地球温暖化は一層深刻化する
The global warming is becoming a more serious problem.
3. 再生可能エネルギーは開発せよ
しかし時間と金がかかる
The development of renewable energy has to be promoted. However, it will require sufficient
resources of time and budget.
4. 人類は原子力も使わざるをえない
Human beings cannot help depending on nuclear energy as well as other energy resources.
5-1 原子力の将来のための安全技術を開発せよ
The safety technology of nuclear energy has to be developed for the future.
5-2 バック・エンド技術を強化し、最終処理場を早く決めよ
それは国の責任
The technology for the back-end of nuclear fuel cycle has to be enhanced. The site for final
disposal of nuclear wastes has to be determined as soon as possible, which is a responsibility of
the Government of Japan.
5-3 ADSのような技術を研究開発し、最終処理をやり易いようにせよ
The research and development of innovative technologies such as Accelerator Driven System haveto be promoted in order to encourage the progress of final disposal.
5-4 廃炉、安全技術、バックエンド技術etc. で国際協力を強力に行え
The research and development of nuclear technologies for reactor decommissioning, safety
technology, and back-end, etc., have to be promoted intensively through international cooperation.

11

Contribution of the European Commission to a European Strategy
for HLW Management through Partitionning & Transmutation
Presentation of MYRRHA and its in the European P&T strategy
Hamid AÏT ABDERRAHIMSCK•CEN, Boeretang 200, B-2400 Mol, Belgium
haitabde@sckcen.be or myrrha@sckcen.be
Presently, the European Union relies for 30% of its electric power production on Generation II-III
fission nuclear reactors leading to the annual production of 2500 t/y of used fuel, containing 25 t of
Plutonium, and High Level Wastes (HLW) such as 3.5 t of minor actinides (MA), namely Neptunium(Np), Americium (Am) and Curium (Cm) and 3 t of long-lived fission products (LLFPs). This MA and
LLFP stocks need to be managed in an appropriate way. The used fuel reprocessing (closed fuel cycle)
followed by the geological disposal or the direct geological disposal (open fuel cycle) are today theenvisaged solutions depending on national fuel cycle options and waste management policies. Required
time scale for the geological disposal exceeds our accumulated technological knowledge and thisremains the main concern of the public. The Partitioning and Transmutation (P&T) has been pointed out
in numerous studies as the strategy that can relax constraints on the geological disposal, and reduce themonitoring period to technological and manageable time scales. Therefore a special effort has to bemade to integrate P&T in advanced fuel cycles and advanced options for HLW Management.
Transmutation based on critical or sub-critical fast spectrum transmuters should be evaluated, in order toassess the technical and economical feasibility of this waste management option, which could ease thedevelopment of a deep geological storage.
Despite diverse strategies and policies pursued by European Member States concerning nuclear
power and envisaged fuel cycle policy ranging from the once through without reprocessing to the
double-strata fuel cycle ending with the ADS as the ultimate burner or Gen-IV fast critical reactors
multi-recycling all transuranic (TRUs), P&T require an integrated effort at the European and evenworldwide level. Even when considering the phase out of nuclear energy, the combination of P&T and a
dedicated burner such as ADS technologies, at a European scale, would allow to meet the objectives ofboth types of countries, the ones phasing out the nuclear energy as well as countries favouring thecontinuation of the nuclear energy development towards the deployment of new fast spectrum systems.
After nearly twenty years of basic research funded by national programmes and EURATOM
framework programmes, the research community needs to reach a position of being able to quantify
indicators for decision-makers, such as the proportion of waste to be channelled to this mode ofmanagement, but also issues related to safety, radiation protection, transport, secondary wastes, costs,
and scheduling.
From 2005, the research community on P&T within the EU started structuring its research towards amore integrated approach. This resulted during the FP6 into two large integrated projects namely
EUROPART dealing with partitioning and EUROTRANS dealing with ADS design for transmutation,
development of advanced fuel for transmutation, R&D activities related to the heavy liquid metal
technology, innovative structural materials and nuclear data measurement. This approach resulted in a
European strategy given in introduction based on the so-called “four building blocks” at engineeringlevel for P&T that will result in identification of the costs and the benefits of partitioning andtransmutation for European society.
The MYRRHA project contributes heavily to the third building block of this European strategy andin this paper we will focus on the ADS programme in the EU through the MYRRHA project.
12

OECD / NEA Activities Related to the Nuclear Fuel Cycle
Thierry Dujardin
OECD / NEA
Acting Deputy Director-General
Deputy Director, Science and Development
The OECD Nuclear Energy Agency is an intergovernmental organisation which mission is:- To assist its member countries in maintaining and further developing, through international
co-operation, the scientific, technological and legal bases required for a safe, environmentallyfriendly and economical use of nuclear energy for peaceful purposes, as well as- To provide authoritative assessments and to forge common understandings on key issues, as
input to government decisions on nuclear energy policy and to broader OECD policy analyses
in areas such as energy and sustainable development.
The OCDE/NEA is conducting a series of projects related to the Nuclear Fuel Cycle, mainly throughthree of its standing technical committees: the Nuclear Science Committee, the Nuclear Development
Committee, and the Committee on the Safety of Nuclear Installations which methods of work will bebriefly described.
Scientific activities related to fuels, materials, physics, separation chemistry and fuel cycle scenarios
are being undertaken, mostly focussing on advanced nuclear systems.
The recent study on the Economics of the Back-End of the Fuel Cycle and its policy
recommendations will also be presented.

13

ADS Study in JAEA
Hiroyuki OIGAWA

Japan Atomic Energy Agency
Management of radioactive wastes is one of the most critical issues to use nuclear power. By
transmuting long-lived minor actinides (MA) contained in high-level radioactive wastes to short-lived or
stable nuclides, reduction of the burden of nuclear waste disposal can be foreseen. Aiming at effective
transmutation of MA, the Japan Atomic Energy Agency (JAEA) is conducting research anddevelopment on an Accelerator-Driven System (ADS), which is a subcritical nuclear reactor coupledwith an intense spallation neutron source driven by a proton accelerator. One unit of ADS with thermal
power of 800 MW can transmute 250 kg of MA annually, which corresponds to the amount generated in
10 units of light water reactors with electric power of 1 GW per year. In JAEA, a variety of technicalareas are being studied to verify feasibility of ADS. A superconducting proton LINAC is being
developed, and its reliability is being investigated. Lead-bismuth eutectic (LBE) is considered as a
primary candidate of a spallation target and a core coolant, though we do not have sufficient experienceon it. We are, therefore, developing the LBE technology from viewpoints of material corrosion, thermal
hydraulics, activation products, and so on. For the safety of a subcritical core, we are performing
transient analysis on various accidental conditions. Moreover, construction of the Transmutation
Experimental Facility (TEF) is being planned as a phase-2 program of the J-PARC project to performbasic experiments for transmutation technology. TEF consists of two facilities: the ADS Target Test
Facility (TEF-T) and the Transmutation Physics Experimental Facility (TEF-P). TEF-T is a material
irradiation facility by using 400 MeV – 250 kW proton beam and an LBE spallation target. In TEF-T,
we expect to accumulate important experimental data for beam window of ADS. TEF-P is a critical /subcritical assembly with maximum thermal power of 500 W. TEF-P can accept a 400 MeV – 10W
proton beam. MA-bearing fuel can be loaded in TEF-P. We will investigate physics properties oftransmutation systems in TEF-P.

Keywords
minor actinides, high-level radioactive wastes, transmutation, accelerator driven system, spallation target,
superconducting proton accelerator, subcritical core, Transmutation Experimental Facility, J-PARC
14

Accelerator-Driven System (ADS) Study in
Kyoto University Research Reactor Institute (KURRI)
Cheol Ho PYEON
Nuclear Engineering Science Division, Research Reactor Institute, Kyoto University
The experimental studies on the accelerator-driven system (ADS) are being conducted for an
application to nuclear transmutation techniques with the combined use of the Kyoto University Critical
Assembly (KUCA; A-core: solid-moderated and -reflected core) and the fixed-field alternating gradient(FFAG) accelerator, in the Kyoto University Research Reactor Institute. At KUCA, the spallation
neutrons generated by 100 MeV proton beams from the FFAG accelerator had been successfully
injected into the uranium-loaded core for the first time in the world, and also into the thorium-loaded
core. For the ADS experiments with 100 MeV protons, the static conditions and kinetic behaviors were
examined in aiming at the establishment of measurement techniques of reactor physics parameters,
including the reaction rates, the neutron spectrum, the neutron multiplication, the neutron decay
constants and the subcriticality, and the confirmation of numerical simulation precision by the Monte
Carlo approach.
An upcoming ADS at KUCA could be composed of highly-enriched uranium-fueled and Pb-Bireflected core, in consideration of actual ADS designed by the Japan Atomic Energy Agency.
Especially the neutronic characteristics of Pb-Bi are considered importantly analyzed experimentallyfrom in the viewpoint of reactor physics: neutron yield and neutron spectrum by Pb-Bi target;
uncertainties of Pb-Bi cross sections in the core. Also, Pb-Bi liquid behaviors are expected to be
examined with the use of Pb-Bi liquid loop equipments and test facilities for the investigation of Pb-Bi
liquid characteristics from the viewpoints of heat transfer and thermal hydraulics in two-phase flow. In
the future, as preliminary study on the Pb-Bi characteristics, the sample worth experiments on Pb, Bi
and Pb-Bi could be conducted to investigate the uncertainties of these cross sections with the use of thesolid plates. Furthermore, irradiation experiments of the minor actinides (237Np and 241Am) could be
carried out in hard spectrum core at KUCA to conduct the feasibility study on conversion analyses(237Np/238U and 241Am/235U).
Keywords
ADS, KUCA, FFAG accelerator, Pb-Bi, Nuclear transmutation15

Nuclear Data Study for Nuclear Transmutation
Masayuki IGASHIRA
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology
The current national policy in Japan for the management and disposal of Minor Actinides (MAs:
Np-237, Am-241, Am-242m, Am-243, Cm-244, Cm-245, Cm-246, etc.) and Long-Lived Fission
Products (LLFPs: Se-79, Zr-93, Tc-99, Pd-107, Sn-126, I-129, Cs-135, etc.) generated in fission
reactors is a sequence of vitrification, interim storage, and then depositing underground together withother nuclear wastes. Alternatively, if MAs and LLFPs are extracted and transmuted into stable nuclides,
the environmental loading in the geological repository becomes very small. Moreover, the ethicalproblem that MAs and LLFPs are undesirable property for our far descendants will be solved.
From this viewpoint, the nuclear transmutation of MAs and LLFPs is a very attractive subject, andneutron capture and/or fission reactions are the most promising transmutation reactions. Therefore,
databases on neutron capture and/or fission reaction cross sections are indispensable for developing thetransmutation technology. However, the accuracy of nuclear databases for MAs and LLFPs is quite poor
both in quality and quantity at the present time. Therefore, the improvement of the database accuracy isan urgent task.
In this presentation, the present status of the accuracy of nuclear databases such as JENDL-4.0 and
ENDF/B-VII.1 is explained for some important nuclides of MAs and LLFPs. Then, a general view ofactivities of nuclear data measurements in the world such as those of n_TOF at CERN, LANSCE at
LANL, and ANNRI at J-PARC/JAEA is taken. Finally, a Japanese nuclear data project entitled“Systematic Study on Neutron Capture Reaction Cross Sections for the Technological Development of
Nuclear Transmutation of Long-Lived Nuclear Wastes” is briefly described. The objective of the projectis to contribute to the improvement of the database accuracy, by making the precise measurements ofcapture cross sections of MAs and LLFPs, analyzing the measured results theoretically, elucidating thecapture reaction mechanism of MAs and LLFPs, and supplying reliable calculated capture cross sections
for all MAs and LLFPs in the whole neutron energy region.
This work was supported by JSPS KAKENHI Grant Number 22226016.
Keywords
Nuclear transmutation, Minor actinides, Long-lived fission products, Neutron capture, Neutron fission,
Cross sections

16

Poster session

1a-1

Physical Analysis of LBE Spallation Target for CIADSLu ZHANG1, Kun FENG1, Yucui GAO2, Deliang FAN1, Yongwei YANG1*¹Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000, China²The School of Electronic and Information Engineering, XI’AN JIAOTONG UNIVERSITY, Xi’an 710049,china

For the project of the Chinese Initiative Accelerator Driven Sub-critical system(CIADS), the Lead-Bismuth-Eutectic( LBE)
spallation target is one of the two alternatives, which has high quality of thermal performance, mature technology, and otheradvantages. In the ADS, the high intensity proton beam bombards on the spallation target through the beam window and produces
neutrons, then the neutrons are used to driven the sub-critical system. Physically, the design of the spallation target determines the
production and the utilization of the neurons, as well as the performance of the sub-critical reactors and other key issues.Focusing
on the energy deposition and the neutron flux distribution of the target and the beam window, we got the curve of target withdifferent geometric parameters and beam sizes, through a series of calculations using the Monte Carlo program-MCNPX. Analysisof the target performance coupled to the sub-critical reactor was done, from the point of view of physics. And targets withdifferent locations and different geometries in the core of sub-critical reactor were studied considering the energy deposition rateof the target and windows. In the mean time, the profile of neutron flux and the neutron energy spectrum for different positions
were calculated. The relationship of keff and the power amplification factor of the couple system were also studied. The energy
deposition rate can give a heat source for the thermal-hydraulics analysis. And other parameters can be used for the optimizationof the LBE target design.
Keywords: LBE Target , Energy Deposition Rate , Target coupled with the reactor , keff

17

1a-2
Analysis of Thermal-hydraulics for the LBE Spallation Target of ADS
Kun FENG, Lu ZHANG, Deliang FAN, Zhuang WU, Yongwei YANG*
Institute of Modern Physics Chinese Academy of Sciences, Lanzhou 730000, China

The spallation target is an important part of the Acceleration Driven sub-critical System (ADS). A loop-type Lead-BismuthEutectic(LBE) spallation target was studied through the investigation of the existing LBE target concepts . The code MCNPX wasused for the physics simulation of the spallation process, the CFD code FLUENT was used for the thermal-hydraulics analysis.
The methodology was verified and validated by repeating the work mentioned in some literatures. This paper mainly introduces
to our work related to the analysis of thermal-hydraulics of the spallation target. The thermal-hydraulic behavior of the target has
been studied by using FLUENT. A series of the results were obtained by changing the inlet velocity of LBE, the diameter ofproton spot, and the intensity of proton beam under the condition that the thickness of target window is fixed, and theallowable intensities of beam under the different thickness of target windows were obtained. The operating condition which meetthe criterion of thermal-hydraulics with enough margin was given .The results are valuable for the parameter choice of the LBE
spallation target for the Chinese Initiative Acceleration Driven sub-critical System (CIADS)
Keywords
LBE spallation target, ADS, Thermal-hydraulics

18

1a-3

Study on Neutron Spectrum of Pulsed Neutron Reactor
Takanori KITADA*, Vu Thanh MAI and Noboru DOBUCHI
Osaka University, Graduate School of Engineering, Division of Sustainable Energy and Environmental Engineering

Accelerator Driven Subcritical Reactor (ADSR) is considered as one of the best candidate to annihilate the radioactivity of
nuclear waste, and has been investigated in many institutes for many years. ADSR is operated by the pulsed proton beam as an
ignition of spallation reaction to produce many neutrons. Kyoto University Critical Assembly (KUCA) is one of the facilities todemonstrate ADSR by using accelerated proton for spallation reaction or deuteron for DT reaction.
The experiments at KUCA modeled thermal reactor with DT reaction had been performed in order to clarify the validity ofreactivity measurement through the fitting of neutron decay constant at various positions in subcritical and critical states. For the
analysis of the experiments, continuous energy Monte Carlo code: MVP-II with JENDL-4 library was used in eigenvalue modeand time-dependent mode.
The neutron distribution evaluated by time-dependent mode is changing with time after neutron ignited into the system, and theneutron distribution in energy and space becomes almost stable after 1 micro-second of neutron ignited. This paper focuses on
the neutron spectra of two modes in subcritical state, although there are many specific points to be considered in subcritical system.
There is a remarkable difference in neutron spectra between two results of eigenvalue and time dependent modes. The neutron
spectrum (at 1 micro-second after neutron ignited) evaluated in time dependent mode is softer than that in eigenvalue mode, andthe difference is more remarkable as the subcriticality is larger. This fact can be understood by considering the phenomena causedin the subcritical system described in the following. The ignited neutron causes many types of reactions; fission, radiative capture,scattering, etc. In subcritical system, the magnitude of neutron is decreasing with time but thermal neutron decrease after fastneutron starts to decrease. Thermal neutron needs additional time to be moderated in the system, in other words, although thermal(moderated) neutron decreases after moderation process, fast (un-moderated) neutron decreases without moderation process.
This causes the time-difference of decreasing between fast and thermal neutrons, and thus neutron spectrum is softer in timedependent mode compared to that in eigenvalue mode.
Eigenvalue mode is usually used to evaluate k-eigenvalue, and k-eigenvalue is useful to check the subcriticality of the system.
It is widely known that the neutron distribution in space obtained by k-eigenvalue mode is different from that by fixed-source
mode, therefore there are many suggestions about how to evaluate the effective-subcriticality. In addition to this, this paperemphasizes the necessity of the consideration to evaluate the effective-subcriticality that the neutron distribution in energy obtainedby k-eigenvalue mode is also different from that by time-dependent mode. When the neutron spectrum of pulsed neutron reactor
is evaluated, the alpha-eigenvalue mode instead of k-eigenvalue mode is recommended in order to match the neutron spectrum
during the decreasing in time after pulsed neutron is ignited to the subcritical system.
Keywords
ADSR, Neutron spectrum, k-eigenvalue mode, time-dependent mode, alpha-eigenvalue

19

1a-4
Theory of power spectral density and Feynman-α method in Accelerator
Driven System and their higher order mode effects
Toshihiro YAMAMOTO*
Research Reactor Institute, Kyoto University

It is important to monitor the subcriticality of an accelerator driven system (ADS) for maintaining its nuclear criticality safety.
Reactor noise techniques are promising candidates for its purpose since the techniques can be installed with the minimum
disturbance of the reactor operations. This paper focuses on the methods of power spectral density and Feynman-α as reactor noise
methods for monitoring the subcriticality of an ADS where pulsed spallation neutrons are injected with a constant time-period.
That is, the neutron source in an ADS does not have a Poisson character. Thus, the conventional reactor noise theory, which does
not account for either periodically pulsed neutron source or for its non-Poisson character, cannot be applied to the noise
measurements in ADSs. Many studies have been performed for noise theory in ADSs. Through these studies we know that the
Feynman Y function is composed of the sum of a correlated and uncorrelated component. On the other hand, the power spectral
density is also composed of two components: a correlated component and an uncorrelated component of an infinite number of the
Dirac delta function peaks. Yamamoto very recently developed a Monte Carlo method to calculate a series of the Dirac delta
function peaks by solving the frequency domain neutron transport equation using the complex-valued weight Monte Carlo
method1). In this paper, the formulae of the Feynman Y function and power spectral density are presented for an infinite
homogeneous multiplying system where neither spatial-higher order modes nor energy-higher order modes are excited. These
formulae are verified with comparison to their Monte Carlo simulations.
In a realistic ADS, the spatial or energy- higher order mode effects cannot be neglected because ADSs are operated with acertain level of subcriticality and their dimensions are finite. To obtain accurate information on the subcriticality, higher order mode
effects in the Feynman Y function and power spectral density need to be evaluated and thus the theory on the higher order modesof the reactor noise techniques in an ADS is needed. This paper will present the space- and energy-dependent formulae of the
Feynman Y function and power spectral density. These formulae have been applied to numerical test calculations for an ADS with
one-dimensional infinite slab geometry. The Feynman Y function and power spectral density are obtained by the Monte Carlosimulations for noise measurements in the ADS. The results of the Monte Carlo simulations are compared with the theoreticalformulae. Both results agree very well, which may show verification of the theoretical formulae. The Feynman Y function andpower spectral density are decomposed into the sum of the fundamental mode, which we seek to know, and higher order modes.
Generally speaking, the higher order mode effects are more significant in the uncorrelated component of the Feynman Y function
and power spectral density. Especially, it is found that the first higher order mode is almost comparable to the fundamental modein the uncorrelated component. Thus, an α value obtained by fitting the uncorrelated component of the power spectral density to aconventional formula for power spectral density, A /(α 2 + ω 2 ) , largely deviate from the fundamental mode α value. On theother hand, the correlated component of the power spectral density is less contaminated by the higher order modes. We can obtainan α value that is much closer to the fundamental mode α value by using the correlated component.
1) Toshihiro Yamamoto, “Higher order mode analyses of power spectral density and Feynman-α method in accelerator driven
system with periodically pulsed spallation neutron source,” Ann. Nucl. Energy (in press).
Keywords
Rector noise, power spectral density, Feynman-α, ADS, higher order mode

20

1a-5

Neutron requirement for transmutation of MA isotopes
W.F.G. VAN ROOIJEN*
Research Institute of Nuclear Engineering, University of Fukui, Tsuruga, Japan

In the aftermath of the accident at the Fukushima-1 Nuclear Power Station, a reorientation is occurring in Japan with respect
to the future of nuclear power. One of the topics that is gaining interest is the process of actinide transmutation to reduce the amountof long-lived nuclear waste (LLW). The objective of transmutation is to fission the LLW nuclides, because fission product isotopes
in general have a relatively short half-life.
Minor actinide (MA) nuclides remaining in spent nuclear fuel are by definition not fissile, so one expects to require a minimumof roughly two neutrons for transmutation: one neutron to turn the non-fissile isotope into a fissile isotope, and a second neutron tocause fission. If the first or second neutron does not cause fission, one needs even more neutrons for transmutation. When thefission occurs, ν new neutrons are released into the system. One needs to calculate how many neutrons are needed to fullytransmute a given isotope and all of its possible descendants. If the balance is positive, full transmutation can be achieved with a(series of) critical reactor(s). If the balance is negative, external neutrons are needed – provided for instance by an Accelerator
Driven System (ADS) or a nuclear fusion reactor (fission – fusion hybrid).
The total number of neutrons released in a transmutation chain can be calculated from elementary considerations as asummation over the probability that transmutation reaction x occurs, multiplied by the number of neutrons released in that reaction(-1 for capture, ν-1 for fission, (N-1) for (n,Nn), 0 for radioactive decay, etc). The governing formula is derived in [1]:n� r,I = � PI→j1 �n� r,j1 + � Pj1→j2 �n� r,j2 + � Pj2→j3 ⋯��
j1

j3

j2

This equation is somewhat cumbersome to evaluate on a computer (not in the least because the transmutation chain is infinitely
long in principle). Therefore, we propose to use a somewhat different approach. Suppose one has a system with a constant neutron
flux, to which one isotope is added continuously. After sufficient time, the mixture will reach a steady state composition. The
composition reflects the probabilities of the transmutation reactions. Thus neutron balance for transmutation is related to the infinite
multiplication factor of the hypothetical steady state mixture. This will be detailed in the full paper.
A computer program based on this approach was created, using the transmutation matrix from ORIGEN-S. Some results are
given in the table. It is clear that complete transmutation, in general, cannot be achieved with thermal reactors alone; fast reactors
are needed. The most important isotope is obviously U-238, as this is the source isotope for all MA.
Isotope

ThermalFast

Depending on the future availability of fast reactors, external neutrons (ADS) will be
required for transmutation purposes..

U-238

-0.51

0.72

Pu-2390.23

1.54[1] M. Salvatores, I. Slessarev, M. Uematsu, “A global physics approach to transmutation ofradioactive nuclei”, Nucl. Sc. Eng. 116, 1-18 (1994)

Np-237

-1.44

0.61

Am-241

-1.54

0.66

Cm-244

-1.28

1.45

Keywords
Nuclear transmutation, neutron economy, ADS, transmutation reactor

21

1a-6
Application of the Resource-renewable Boiling Water Reactor for TRU
Management and Long-Term Energy Supply
Tetsushi HINO*¹, Masaya OHTSUKA¹, Renzo TAKEDA¹, Junichi MIWA¹, Kumiaki MORIYA²¹Hitachi, Ltd. Hitachi Research Laboratory²Hitachi-GE Nuclear Energy, Ltd.
An adaptation of the BWR has been conceived that has the potential to breed and consume transuranium elements (TRUs)
using a multi-recycling process. This is the resource-renewable BWR (RBWR), and it can be used as a long-term energy supply,
while greatly reducing the negative environmental impact that TRUs cause as they are otherwise long-lived radioactive wastes.
A major characteristic of the BWR is “boiling” in the core which includes water that has the roles of being a moderator and acoolant. The neutron energy spectrum can be hardened by reducing the hydrogen-to-uranium ratio (H/U) using the two-phase flow
and combining with the hexagonal tight fuel lattice, so that the transmutation of U-238 to fissile plutonium is promoted withincreasing resonance absorption. This enables the multi-recycling process of both breeding and consuming TRUs.
The rated electric power, core pressure and diameter of the pressure vessel of the RBWR are 1350 MWe, 7.2 MPa and 6200
mm, respectively, and these values are identical to those of the latest commercial BWR, the advanced ABWR (ABWR). The coreconfiguration in the horizontal direction is composed of 720 hexagonal fuel bundles and 223 Y-type control rods. The height
configuration uses a parfait core concept in which an internal blanket of depleted uranium oxide is placed between the upper andlower fissile zones of the TRU oxides. Cores with a compatible fuel bundle each other, but for different purposes, have been
proposed so far; they are the TRU burners (RBWR-TB/TB2) and the break-even reactor (RBWR-AC). The RBWR has acapability to make the transition between the TRU burners and the RBWR-AC only by changing fuel bundles.
The RBWR-TB is the TRU burner that is designed to fission almost all the TRUs, leaving only a minimum critical mass ofTRUs, by repeating the recycling of them to remove the concern that TRUs will be long-lived radioactive wastes when theirusefulness has been exhausted. The RBWR-TB2 is another type of the TRU burner, and it is designed to be suitable to burn TRUs
from LWR spent fuels, while the RBWR-TB is designed as a burner for TRUs from the RBWR-TB itself. The fuel bundles of the
RBWR-TB and TB2 have a relatively larger H/U value than that of the RBWR-AC to transmute fertile minor actinides into fissile
fuel elements and fission them. The numerical calculation showed the fission efficiencies of TRUs are more than 50% in the
RBWR-TB and more than 40% in the RBWR-TB2. The fission efficiency is defined as the net decrease in TRUs divided by the
total amount of fissioned actinides for the total fuel life. The fission efficiencies in the RBWR-TB and RBWR-TB2 respectively
correspond to more than twice and nearly twice the amount of TRUs accumulated in the ABWR per generating power.
The RBWR-AC is the break-even reactor, which has a breeding ratio of 1.01. This ratio is defined as the number of atoms offissile plutonium left in the discharged fuel bundles per fissile plutonium in the initial fuel bundles. The RBWR-AC burns depleted
uranium by using TRUs extracted from the spent fuel bundles of LWRs without decreasing the amount of the TRUs. The fuelbundle of the RBWR-AC is composed of 271 fuel rods in a triangular lattice.
All of the above RBWR cores achieve negative void coefficients. In addition the isotopic composition of TRUs will bepreserved through the multi-recycling process by adjusting the neutron energy spectrum. This is necessary to continue the fission
and recycling process of TRUs while maintaining the criticality and meeting the various constraints, such as the negative void
coefficient.
Various fast reactors have already been proposed and developed for the TRU management so far, but it is thought to be alsovaluable to investigate achieving breeding and consuming of TRUs by LWRs with proven technologies.
Keywords
BWR, TRU, burner, break-even, void coefficient, multi-recycle

22

1a-7

Investigations of Lead-Bismuth cooled Accelerator-Driven System
Takanori SUGAWARA*, Kenji NISHIHARA, Kazufumi TSUJIMOTO, Toshinobu SASA, Hironari OBAYASHI,
Yuji KURATA and Hiroyuki OIGAWA
Japan Atomic Energy Agency

The Japan Atomic Energy Agency (JAEA) has been conducting research and development on an accelerator-driven system(ADS) to transmute minor actinides (MAs) included in high-level waste (HLW) into short-lived nuclides. The ADS proposed by
JAEA is a lead bismuth eutectic (LBE) cooled-tank-type ADS. The system consists of a high-intensity proton accelerator with
1.5GeV beam energy, a subcritical core with 800MW thermal power and an LBE spallation target placed at a center of thesubcritical core. Since the ADS is a hybrid system of an accelerator and a nuclear reactor, there are various inherent issues in the
research and development of the ADS. In this study, two important issues are introduced from the viewpoint of the ADS plantinvestigation and the latest results are presented.
The first issue is a design of a beam window. The beam window forms a boundary between the accelerator and the spallation
target. It is obvious that the beam window should be tolerant to the following severe conditions: external pressure by the LBE, heatgeneration by the proton beam, creep deformation at high temperature, corrosion in the LBE and irradiation damage by neutronsand protons. Through the calculations of particle transport, thermal hydraulics and structural analysis, a feasible beam window
concept, which was an ellipse one with 2mm thickness as minimum, was presented with the consideration of the external pressure,the heat generation and the corrosion effect. The consideration of the creep deformation and the irradiation damage will be thefuture work for the beam window design.
Another issue is a safety of the ADS. It is considered that the ADS is safer than critical reactors because the ADS is operated
in a subcritical state. Since the system is operated under the subcritical condition, its shutdown is easy and a possibility of a criticalaccident is smaller than the conventional critical reactors. Considering such characteristics of the ADS, a preliminary investigation
for the safety was performed by the Level 1 PSA (Probabilistic Safety Assessment) and transient analyses were carried out for the
typical accidents for the ADS. In the case of beam window breakage, as a typical ADS accident, it was confirmed that the LBE
flowed into a beam duct and the spallation neutrons in the subcritical core decreased. Then, the core power decreased. In the caseof protected loss of heat sink (PLOHS), as one of the severest accidents, the calculation result indicated that there was a possibilityof the core damage because the cladding tube temperature reached at its melting point after 20h. As a next step, it is required todesign a safety system of the ADS to decrease these frequencies and to ease the accidents.
Keywords
Accelerator-Driven System (ADS), Beam window, Transient analysis, Safety, Lead-bismuth eutectic (LBE)

23

1a-8

J-PARC Transmutation Experimental Facility
Toshinobu SASA*, Hayanori TAKEI, Shigeru SAITO, Hironari OBAYASHI, Kenji NISHIHARA, Takanori SUGAWARA,
Hiroki IWAMOTO, Kazufumi TSUJIMOTO and Hiroyuki OIGAWA
Japan Atomic Energy Agency

After the Fukushima accident, nuclear transmutation got much interested as an effective option of nuclear waste management.
Japan Atomic Energy Agency (JAEA) proposes the transmutation of minor actinides by accelerator-driven system (ADS) using
lead-bismuth eutectic alloy (Pb-Bi) as a spallation target and a coolant of subcritical core. Current design of ADS has 800MWth of
rated power, which is driven by 20MW proton LINAC, to transmute minor actinides generated from 10 units of standard Light
Water Reactors.
To obtain the data required for ADS design including European MYRRHA project, JAEA plans to build a Transmutation
Experimental Facility (TEF) within the framework of J-PARC project. TEF consists of two buildings, one is an ADS target test
facility (TEF-T), which will be installed high power Pb-Bi spallation target, and the other is Transmutation Physics Experimental
Facility (TEF-P), which set up a fast critical/subcritical assembly driven by low power proton beam. TEF will be located at the endof 400 MeV LINAC of J-PARC and accepts 250kW proton beam with repetition rate of 25 Hz.
As a major research and development items of TEF-T, irradiation test for candidate ADS structural materials, engineering tests for
Pb-Bi target operation and experiments to determine the effective lifetime of proton beam window will be performed. Several
kinds of Pb-Bi target will be used according to above mentioned research purposes. Reference design parameter, that considers
operation condition of ADS transmutor, was determined by thermal-hydraulic analyses and structural analyses. When the target
operates with full power beam, fast neutron spectrum field is formed around the target and it is possible to apply multi-purpose
usage. Various research plans have been proposed and layout of the experimental hall surrounding the target are underway. Basicphysics application such as measurements of nuclear reaction data is considered as one of the major purposes.
In the presentation, roadmap to establish the ADS transmutor and design activities for TEF construction will be summarized.
Keywords
Transmutation, Accelerator-driven System, J-PARC, Transmutation Experimental Facility

24

1a-9
Development of Uranium-Free TRU Metallic Fuel Fast Reactor Core with
Enhanced Doppler Reactivity Feedback
Kyoko ISHII¹*, Mitsuaki YAMAOKA¹, Yasuyuki MORIKI1, Takashi OOMORI1, Yasushi TSUBOI1, Kazuo
ARIE1, Masatoshi KAWASHIMA2¹Toshiba Corporation²Toshiba Nuclear Engineering Service Corporation

For sustainable nuclear power deployment, not only ensuring its enhanced safety but also reduction of environmental burden
associated with radioactive waste management is a challenging issue for the international community. History has shown thatgetting public support is difficult for waste management plans which involve mass disposal of radioactive waste with a half-life oftens of thousands of years. Therefore, as one of the solutions Toshiba has been developing a system, taking into account that for
the time being light water reactors (LWRs) play a leading role in commercial nuclear power plants, which enables toxicity andradioactivity of high level waste to be reduced to the same as those of natural uranium within a few hundred years.
This system is characterized by a closed fuel cycle which encircles following main facilities: fuel manufacturing plants which
fabricates uranium-free transuranium (TRU) metallic fuel and long-lived fission products (LLFPs) target from TRU and LLFPs
extracted from LWR spent fuel, fast reactors to burn those fuel, and recycling facilities to reprocess and refabricate the spent fuelfrom fast reactors by pyroprocessing. Use of TRU metallic fuel without uranium makes it possible to maximize TRU transmutation
rate in comparison with uranium and plutonium mixed oxide fuel, because it prevents the fuel itself from producing new plutonium
and minor actinides, and furthermore because metallic fuel has much smaller capture-to-fission ratios of TRU than those of mixed
oxide fuel. Besides, adoption of metallic fuel enables recycling system less challenging even for uranium-free fuel because
conventional scheme of fuel recycling by electrorefining and injection casting is applicable. Although substances remained after
reprocessing are finally to be disposed outside the cycle, their toxicity and radioactivity are diminished to the same level as thoseof natural uranium by establishment of technologies to enhance burning and processing rates and storing them for a few hundredyears within the system.
Safety prospects for uranium-free TRU fuel, however, are yet to be ensured since Doppler reactivity feedback of the core is
anticipated to decrease seriously in the absence of uranium. The purpose of this paper is to analyze the effect of following measurestaken, respectively, to enhance Doppler reactivity feedback of the core: (1) use of the elements with relatively larger resonanceneutron capture cross section under fast neutron spectrum circumstances, e.g. Zr, Nb, Mo, etc., and (2) use of neutron moderators,
e.g. Be, 7Li, 11B, H, etc., to enhance negative Doppler reactivity feedback due to even-plutonium isotopes. Assuming a typical
prototype fast reactor core, analysis was performed using in-house two dimensional neutron diffusion calculation code. The results
show that the equal or much greater negative Doppler coefficient compared to that of the metallic fuel containing uranium can beobtained when using elements such as Nb or Mo as a metal fuel alloy or moderators such as BeO, 7Li2O, 11B4C, or ZrH2.
Consequently, it is confirmed that aforementioned measures have potentiality to enhance Doppler reactivity feedback of theuranium-free TRU metallic fuel core.
Keywords
Uranium-free, Fast Reactor, Metallic Fuel25

1a-10Chemical Isotopic Analysis and Computational Evaluation of
Nuclide Composition in PWR-UO2 and MOX Spent Fuels Using
JENDL, ENDF/B, JEF, and JEFF
Akihiro SASAHARA*¹ and Tetsuo MATSUMURA2¹Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry,
2-11-1, Iwado Kita, Komae-Shi, Tokyo 201-8511, Japan
2
IERE Central Office, 2-11-1, Iwado Kita, Komae-Shi, Tokyo 201-8511, Japan

Isotopic composition in the spent fuels is basic data to evaluate criticality of the spent fuels as well as the source intensity
such as radiation emission and heat generation. Especially, at the accident of the nuclear reactors, the estimation of the inventoryof the nuclides in the fuels loaded into reactors or spent fuel storage pools is quite important. Therefore, the verification and theimprovement of the accuracy for the calculated nuclide compositions in spent fuels are required.
In this study, chemical isotopic analyses of actinides and fission products were carried out on a high burn-up PWR-UO2 anda high burn-up MOX fuel. Furthermore, computational analyses were performed using the integrated burn-up calculation code
SWAT. The differences between the amounts by the chemical isotopic analysis and the SWAT calculation using JENDL-3.2,
JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2 and JEFF-3.0 were evaluated as the ratios of the calculated values to theexperimental ones (C/E ratios). Regarding actinides, the calculated 244Cm amount, which is an important nuclide as a major
neutron source in spent fuels, was underestimated. The main sensitive path to 244Cm was therefore investigated by a simple
depletion calculation for actinides and the cause of its underestimation was discussed. Fission products 88Sr, 90Sr, 89Y, 106Ru,
133Cs, 135Cs and 144Nd in the PWR-UO fuel, and 88Sr, 90Sr, 106Ru, 133Cs, 134Cs and 135Cs in the PWR-MOX fuel, which2
contribute to gamma sources, decay heat, neutron absorption or a burn-up indicator in spent fuels, were further investigated to
improve their C/E values by using simplified burn-up chains of fission products using JENDL-3.3, and consequently, thecorrected value for the fission yield or the capture cross section of fission products was estimated by using the sensitivity
coefficients.
As the results of the analysis for the PWR-UO2 and the PWR-MOX fuels. The calculation accuracy of 244Cm will be quite
improved by the reevaluation for both capture cross sections of 243Am and plutonium isotopes. For 88Sr, 90Sr and 106Ru, thecorrection for their own fission yields improved their C/E ratios. The correction for the 89Sr fission yield improved the C/E ratio
for 89Y, and that for the 133Xe fission yield improved the C/E ratio for 133Cs and 134Cs. For 135Cs, it suggests that theunderestimation of the fission yield or the overestimation of the capture cross section of 135Xe results mainly in theunderestimation of the calculated 135Cs amount, and in the case of 144Nd, it suggests that the underestimation of the 144Ce fission
yield results mainly in the underestimation of the calculated 144Nd amount.
The effect of the power history on nuclide composition was also investigated on the PWR-UO2 fuel. In the comparison of
the calculated amounts with the detailed power history and the constant one, the amounts of 241Am, 242mAm and 155Gd wereslightly affected by the power history.
Keywords
high-burnup UO2 fuel, high-burnup MOX fuel, isotopic analysis, actinides, fission products, sensitivity analysis, JENDL-3.2,
JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2, JEFF-3.0

26

1a-11

Enhancement of Transmutation of Minor Actinides by Hydride Target
Kenji KONASHI*¹ and Tsugio YOKOYAMA2¹Institute for Materials Research, Tohoku University, Oarai, Ibaraki-ken, 311-1313 Japan.
²Toshiba Nuclear Engineering Services Corporation, 8 Shinsugita, Isogo-ku, Yokohama, 235-8523 Japan

High level wastes generated after reprocessing spent nuclear fuels include long-lived radioactive nuclides of minor actinides(MA), i.e. 237Np, 241Am, 243Am, 244Cm and so on. The transmutation of MA by the Fast Breeder Reactor (FBR) has intensively
studied to reduce radioactivity of the wastes. Core design study with a driver fuel containing MA shows limitation on maximum
allowable MA content to avoid the impact of reactivity coefficients. Another loading mode of MA was studied, where MA isseparately located at radial blanket region. The latter mode has no impact on main reactor core parameters but a transmutation rateis small due to lower neutron fluxes. A hydride target including MA is able to enhance the transmutation rate in FBR. Fast neutrons
generated in the core region of FBR are moderated in the hydride target assemblies and then produce high flux of thermal andepithermal neutrons. The neutron reaction cross sections in the thermal energy region are large compared with that in the fast
energy region. In this paper, a design study of core with MA hydride targets shows an advantage in transmutation of MA comparing
with core with a driver fuel containing MA.
Previously, a number of transmutation using moderated subassemblies in FBR were proposed. In the case that Zr hydride andmetal MA are separately located in the transmutation subassemblies, neutrons are moderated in the region of Zr hydride withoutabsorption and cause power peaking in the fuel assemblies adjacent to the targets. The MA hydride target improves the power
peaking problem, since the moderated neutrons are mostly absorbed in the MA hydride target. The MA hydride target has also anadvantage on the increase of the MA loading content, since MA and MA-Zr alloy absorb hydrogen such as Zr.
Transmutation performances of the hydride MA target in fast reactors have been evaluated for the core of Japanese prototype
fast reactor Monju, where the thermal power is 710MWt and the diameter of the active core is about 1800mm and the height is
930mm. A total of 54 MA hydride target assemblies are located at the most inner row of the three radial blanket rows. Eachassembly contains 61 MA hydride target pins, where the diameter of the pellets is 10.4mm and the stack length is 930mm. The
MA hydride target pin has hydride pellets consisting of (MA0.3, Zr0.7)H1.6 ,where the composition of MA is derived from LWR
discharged fuel. The calculations with a metal MA target pin were also done for comparison.
The calculation results show that the transmutation amount of actinides is equivalent to the annual production of that from eight
1000MWe-class LWRs. The reduction ratio, which is the ratio of the transmuted MA to the loaded MA after one year, is 0.266.
The effective half life, which is defined as the time required to be half of the loaded amount of MA is only 2.24 full power years,
which is one fifth times smaller than that of the metal MA case. The hydride target shows excellent performance for transmutationof the wastes. We propose the energy system combining LWRs and FBRs with the hydride targets to enclose the long-lived nuclear
wastes in the fuel cycle.
Keywords
hydride, minor actinide, fast breeder reactor27

1a-12
Project to Modify the STACY Critical Facility for Experimental Studyon Fuel Debris Criticality Control
Hiroki SONO*, Kotaro TONOIKE , Kazuhiko IZAWA , Takashi KIDA , Fuyumi KOBAYASHI ,Masato SUMIYA , Hiroyuki FUKAYA , Miki UMEDA , Kazuhiko OGAWA and Yoshinori MIYOSHI
Japan Atomic Energy Agency

In the severe accident of Fukushima Daiichi Nuclear Power Stations (NPS), most of the fuel loaded in the Unit
1, 2 and 3 cores was seriously damaged and melted, resulting in a considerable amount of fuel debris. It is believedthat some parts of fuel debris involve molten structural materials such as zircaloy, stainless steel and concrete.
Toward the decommissioning of Fukushima Daiichi NPS, the whole fuel debris should be retrieved from eachpressure vessel and containment vessel. The fuel debris, of which chemical composition and physical state areuncertain, needs to be treated with great care from criticality safety's viewpoint. In particular, large blocks of fueldebris can change in physical state such as size and water content when they are broken into fragments to retrieve
in cooling water. Criticality control measures for the fuel debris, therefore, will have to be established by the time
the fuel debris begins to be retrieved from Fukushima Daiichi NPS (scheduled for 2020 by the governmental council
for the decommissioning of Fukushima Daiichi NPS).
For dealing with such a serious problem on criticality control of fuel debris, Japan Atomic Energy Agency(JAEA) has been planning to modify Static Experiment Critical Facility (STACY) and to pursue critical experiments
on fuel debris. STACY, a facility using solution fuel, is to be converted into a thermal critical assembly using fuelrods and light water moderator. It is notable for the modified STACY that its core has a wide range of neutron energy
spectrum between thermal-reactor spectra and intermediate-reactor ones. The core spectrum can cover a relatively
hard spectrum of the fuel debris likely to become critical. A series of critical experiments will be conducted in themodified STACY using simulated fuel debris samples. The simulated fuel debris samples (sintered pellets) are to
be manufactured by mixing uranium oxide and reactor structural materials (Zr, Fe, Si, Gd, B, etc.) with various
chemical compositions. The manufacturing and analytical equipments for the simulated fuel debris samples are to
be installed in glove boxes in the experimental building adjoining the modified STACY.
The STACY modification is under safety review by the regulatory authority of Japan to comply with new safety
standards for research reactors enforced in December 2013. The first criticality of the modified STACY is scheduled
for 2018. The modified STACY will provide benchmark data for criticality safety of fuel debris to validate thecriticality control measures applicable to Fukushima Daiichi NPS.
Keywords
Fukushima Daiichi, Fuel debris, Criticality control, Criticality safety, Critical facility, STACY, simulated fuel debris
sample28

1a-13

ADS study in Tohoku University
Naoto AIZAWA*, Tomohiko Iwasaki, Tatsuro TAKANI, Takashi TANI, Shunsuke KANEMOCHI,
Motomu SUZUKI, Fumito KUBO, Yasuaki WATANABE
Graduate School of Engineering, Tohoku University

Accelerator-Driven System (ADS) has inherent problems arising from the combination of an accelerator and a subcritical core.
Since ADS is stably-operated by the control of an accelerator, the change of beam profiles influences the core and its components.
For example, if the beam incident position is shifted from the center of the target, the thermal power distribution will be varied.
Moreover, the material strength of a beam duct will be worsened as the burnup proceeds because the beam current is gradually
increased in a cycle to compensate the decrease of keff and maintain the thermal power. On the basis of these issues, we workthrough two researches: a reactivity control for the reduction of beam current, and transient analyses on the variation of beam
condition such as shape and position.
With respect to a reactivity control, the introduction of burnable poison is studied, and the metal hydride is employed as thefirst candidate for burnable poison. We have investigated the use of the burnable poison assembly with Gd-hydride in the past, andthe burnup swing is reduced by 74% in the case of ADS with the initial multiplication factor keff = 0.97 proposed by Japan Atomic
Energy Agency (JAEA) as a result of the optimization of the specifications such as the ratio of hydride and loading position. Safetyparameters such as Doppler coefficient and void reactivity coefficient are almost the same as the original core model.
As for transient analyses, ADS dynamics calculation code ADSE has been developed in Tohoku University to analyze thebeam transient which accompanies with the variation of the beam condition as shape and position. ADSE is neutronicsthermohydraulics coupled code, and can deal with the change of the external neutron source. The variation of neutron flux, thethermal power, and temperatures of fuel, cladding and coolant are obtained from ADSE. The soundness of cladding is also able tobe evaluated by the use of the programs of the thermoelastic analysis and the creep analysis. ADSE is applied to the former designof ADS with keff = 0.95 proposed by JAEA, and beam transient is analyzed. The result of the analysis indicates that the stress acting
on cladding exceeds the failure criteria when the incident beam centroid moves large distance from the center of the target.

Keywords
ADS, reactivity control, hydride, transient, beam behavior,29

1a-14
Method Development for Calculating Minor Actinide Transmutation in a
Fast Reactor
Toshikazu TAKEDA*1 and Koji FUJIMURA2
University of Fukui, Research Institute of Nuclear Engineering
Hitach-GE Nuclear Energy, Ltd., Hitachi Works

1
2

Fast reactors are one of the most promising devices to transmute minor actinides which have long-lived toxic because ofpresence of a large number of neutrons and hard neutron spectrum. Here we first define transmutation of each MA nuclide fuelled
in a reactor core. The transmutation rate is composed of two terms; first is the amount of incineration rate by fission and thesecond is the net transmutation rate to fuel (Uranium and Plutonium). The first fission rates of individual nuclides contain the
direct fission of the relevant nuclide plus the fission of other nuclides transmuted by decays or neutron reactions. It was found thatthe indirect fission contribution by Pu-238 and Pu-239 is remarkably large for nuclides Np-239 and Am-241. The net production
rates of U and Pu are calculated from the difference between the production rate of U and Pu from the relevant MA nuclide andthe MA production from the initial U and Pu.
The calculation method of the nuclide-wise transmutation rate has been developed. In this method we first perform theconventional burnup calculations, and store the burnup-dependent flux in each region, which is used in the second step calculation.
In the second step, we consider only the relevant MA in each region, and perform burnup calculations using the flux obtained in
the first step. To transmute MA we consider the homogeneous loading and the heterogeneous loading of MAs. In the homogeneous
loading we optimize the transmutation rate of MAs and the safety aspect of a FR core. The sodium void reactivity worth and the
Doppler reactivity is considered as the safety parameters. To reduce the sodium void reactivity worth we consider a FR core withsodium plenum on the top of fissile region and an internal blanket region. In the heterogeneous loading of MAs we consider special
fuel assemblies with MAs and moderator materials. The merit of the heterogeneous loading is to decrease the number of special
fuel assemblies with MAs et al.
In addition to the improvement of calculation method, we will develop the estimation method of calculation uncertainty
based on the burnup sensitivity analysis. Also to improve the calculation accuracy of the sodium void worth the cross section
adjustment method and the bias factor method will be improved.
Keywords
Minor actinide, Transmutation, Fast reactor30

1a-15Heat transfer study for ADS solid target- surface wettability and its effect on a boiling heat transferDaisuke ITO*¹, Kazuki HASE2, and Yasushi SAITO1¹Research Reactor Institute, Kyoto University²Graduate School of Energy Science, Kyoto University

Accelerator driven system (ADS) has been developed as a next-generation nuclear energy system. High energy particles andhigh density heat are generated by irradiating a proton beam to a spallation target. Although a lead-bismuth liquid target is theleading candidate as a spallation target for transmutation, however a solid target should be also investigated if ADS is used for aneutron source like a research reactor. Therefore, it is essential to establish a heat removal technique in the high energy radiation
field from the solid target cooled by water. Such high energy radiation may affect the surface wettability and the boiling heat transfer.
Recently some researchers have reported that surface wettability effect on a boiling heat transfer by using ultraviolet or gamma ray,however, there is no research on the relation between the proton beam irradiation and the surface wettability. Thus, in this study,
the effect of the proton beam irradiation on the surface wettability was investigated by comparing with those of ultraviolet and
gamma ray irradiations. Various material samples (titanium dioxide, stainless steel, copper, tungsten and tantalum) were irradiatedby the Fixed Field Alternating Gradient (FFAG) accelerator in KURRI. As a result of the proton beam irradiation, the wettability
enhancement was confirmed by measuring a contact angle on the surface before and after irradiations. This effect shows a similartendency as that for ultraviolet and gamma ray. Additionally, the influence of water radiolysis to the radiation-induced surface
activation was also suggested by comparing between the irradiations in water and air. On the other hand, the effect of the surface
wettability on boiling heat transfer was investigated by applying photo- and radiation-induced surface activations. Copper heating
surface coated with titanium dioxide layer was used and the contact angle was changed by ultraviolet irradiation. The contact angleof another copper surface was changed by gamma ray irradiation. Then, boiling heat transfer experiments in a pool boiling setup
were performed with and without irradiation at various subcooling conditions. The heating surface temperature and heat flux were
estimated from measured temperature distribution by assuming one dimensional heat conduction. The boiling regime transits fromnatural convection to nucleate boiling as the wall temperature increases. And then, microbubble emission boiling (MEB) occurred
after critical heat flux (CHF). MEB shows higher cooling characteristics than CHF and was observed at all of the subcooled
conditions at the present experimental conditions. Experimental results show that a nucleate boiling curve moves to higher wall
superheat side with the irradiated surface, however the heat transfer enhancement in the CHF and MEB was not obvious at present
experimental conditions.

Keywords
Surface wettability, Proton beam, Ultraviolet, Gamma ray, Microbubble emission boiling31

1a-16
Experimental study of flow structure and turbulent characteristics in
lead-bismuth two-phase flow
Gen ARIYOSHI*¹, Daisuke ITO2, and Yasushi SAITO²¹Graduate School of Energy Science, Kyoto University²Research Reactor Institute, Kyoto University

For severe accident in an accelerator driven system, a gas-liquid two-phase flow with a large liquid-to-gas density ratio might
appear, such as a steam leakage into hot lead-bismuth flow. It is still difficult to predict such phenomena because there are noavailable flow models for two-phase flow with large density ratio compared to ordinary two-phase flows such as an air-water twophase flow. Therefore, two-phase flow model should be developed based on the experimental data of two-phase flows with large
density ratio. In this study, a liquid metal two-phase flow was measured by using a four-sensor electrical conductivity probe and aminiature electro-magnetic probe in order to establish the experimental database for lead-bismuth flow structure. A lead bismuth
eutectic (LBE) test loop was used to measure the flow structure and turbulent characteristics of the liquid-metal two-phase flow ina vertical pipe with a diameter of 50 mm and a length of 2,000 mm. Nitrogen gas was injected to the test section. So the density
ratio between LBE and nitrogen was about 10,000, which is about 10 times larger than that of air-water two-phase flow. In themeasurement with the four-sensor probe, the radial profiles of void fraction, bubble frequency, interfacial area concentration, gasvelocity and bubble size were measured at different axial positions. The results show that the void fraction profile changes fromwall-peak to core-peak with an increase of the superficial gas velocity at the present flow conditions. The void fraction could be
predicted by one-dimensional drift-flux model. However, existing correlations overestimate interfacial area concentrations.
Experiments were also performed to understand the turbulent structure in a liquid-metal two-phase flow by using the electromagnetic probe. Radial profiles of the axial liquid velocity and turbulence intensity were measured. The turbulence intensity wasincreased as the superficial gas velocity increases. From the data measured by both four-sensor and electro-magnetic probes, it isshown that the turbulence intensity at the pipe center was proportional to the void fraction to the power of 0.8 for higher void
fraction. This result represented the similar tendency as previous data in air-water two-phase flows.

Keywords
Lead bismuth eutectic, Two-phase flow, Four-sensor probe, Electro-magnetic probe32

1a-17
Scenario study for the transition from thermal reactors to fast reactors in theworld
Kyoko Mukaida*¹, Naoto Yasumatsu2, Masanori Heta ², Akira Ohtaki¹, Hiroki Shiotani ¹, Kiyoshi Ono¹, Masaru Hirata¹¹Japan Atomic Energy Agency²NESI

1. INTORODUCTION
Nuclear energy is expected to continue playing an important role in solution for covering the increasing demand for energy mainly
in developing countries, while curbing greenhouse gas emissions. To achieve global sustainable energy future, deployment of fast
reactor (FR) and its fuel cycle system and transition from thermal reactors will be significant. This paper introduces the results ofevaluation carried out to explore the effects obtained by FR deployment in the world through several assumptions and scenarios
using the comprehensive nuclear system evaluation method named nuclear supply chain model (SCM) developed by Japan
Atomic Energy Agency (JAEA).
2. EVALUATION METHOD
SCM is a comprehensive evaluation system capable of evaluating key indicators, such as uranium resources, radioactive waste
generation, non-proliferation, on long-term scenarios by simulating of nuclear supply chain from uranium supply to waste disposal.
In this study, the world nuclear energy capacity was set up at 2500GW in 2100 and it will be continued at same level after the year
and the share of heavy water reactor in the world was set at 6% in 2100. Moreover, it was assumed that three type world nuclear
policies in homogeneous world, i.e. (a) directly disposal of spent fuels (OT), (b) reprocessing spent fuels and recycling plutonium
in light water reactors (LWRs) (PT), and (c) reprocessing spent fuels and recycling plutonium in LWRs and FRs (FR). In case of(c), it was also assumed that the case of only 40% of the country in the world pursues recycling policy (FR40) in addition toheterogeneous world model. The capacity and the share were decided with reference to the scenario standardized by
IAEA/INPRO/GAINS project, and the designs of FR were based on the Fast Reactor Cycle Technology Development Project(FaCT Project) Phase I by JAEA
12

Disposal Volume (Million m3)

3. Results
OT
PT
FR
FR40
10
The result from the evaluation shows the plutonium recycling in
LWR and LWR/FR are effective in reducing spent fuel stock piles,
8
the disposal volume of high-level radioactive waste, plutonium
6
inventories compared with the direct disposal of spent fuels. In
addition, as more realistic approach, when only 40% of the country
4in the world pursues recycling policy, the results indicates the2
reduction of spent fuel stock piles, the disposal volume of high0level radioactive waste, plutonium inventories by plutonium
2060
2110
2160
2010
2050
2100
2150
2200
recycling in LWR/FR although the effect is smaller thanYearhomogeneous world model. Therefore, this study concludes that
Fig. HLW disposal volume
plutonium recycling in LWR/FR will contribute to achieving
global sustainable energy future, and JAEA will contribute to achieve them through R&D of FRs including MONJU and its fuelcycle while international cooperation.
Keywords
Fast Reactor Nuclear Fuel Cycle, Comprehensive Nuclear System Evaluation Method, Global Scenario Study33

1a-18Current Status and Future Prospect of MA Irradiation Tests in Monju
Tetsuya MOURI¹*, Taira HAZAMA¹, Hiroshi NISHI
Japan Atomic Energy Agency, 2-1,Shiraki Tsuruga-shi, Fukui pref, 919-1279,JAPANI. Introduction
This paper overviews the current status and future prospect of the high-level radioactive wastes (HLWs) reduction
technology by FBRs, especially using Monju. Homogeneous MA loading is assumed from the viewpoint of efficient power
generation and not requiring additional specific fuel fabrication facilities.
II. Current status
Preproduction, characterization and material property measurements, and irradiation tests in Joyo have been performed for
the MA (Np and Am) -bearing fuel pellets, as a basic research for future licensing in Monju. Deterioration of thermal conductivity
and melting point, due to Am blending up to around 3wt%, has been found to be in the same order of Pu. Migration behavior of
Am during irradiation has been identified to be nearly the same as Pu.
A useful finding on the reactor physics for MA (Am)-bearing fuel has been obtained from the System Start-up Tests (SST)in Monju. Monju restarted its operation in 2010, after 14-year’s interruption. The core average content of 241Am in the restart
core was 1.5wt%, which accumulated by the 241Pu decay and three times larger than the core before the interruption. Major
reactor physics parameters, such as criticality, control rod worth, isothermal temperature coefficient, and feedback reactivity were
measured. Precise evaluation of the core characteristics measurement and analysis has been performed. As a result, it was
confirmed that the revision of the 241Am cross section in JENDL-4.0 nuclear data library provides consistent results between the
cores before and after the interruption. This implies the adequacy of the revision for the 241Am cross section from JENDL-3.3 to
JENDL-4.0 libraries, which contributes to the precise prediction of the MA (Am) incineration ratio.
III. Future prospect
The development of the radioactive toxicity reduction technology of HLWs has been identified as one of the most importantmissions for Monju at the Working Group on Monju Research Plan, established in the Japanese government. The GACID(Global Actinide Cycle International Demonstration) Project, which is aiming at conducting MA-bearing fuel irradiation tests in
Monju, from on a pin-scale to a bundle scale under collaboration of CEA, DOE and JAEA, has been also positioned as one of the
major activities to be conducted in the future Monju.
The material property measurements and irradiation behavior modeling, based on the Joyo irradiation test, is ongoing forthe licensing in Monju. The fuel fabrication, transportation, irradiation and post-irradiation examinations are scheduled after the
licensing. The first irradiation test is scheduled from the 5th cycle of the operation after restarting Monju. Irradiation tests with
Cm-bearing fuel and on a bundle-scale are the future plan to be implemented.
Preliminary tests are also being planned prior to the GACID Project. Some of the current driver fuels, which contains
241Am up to around 2 wt% at the maximum, are planned to be examined after irradiation. This amount of Am content can berecognized as significant, when compared with the MA content of 1.1 to 1.2 wt% for the future multiple-recycled fuel in the FBR
system. Moreover a pin-scale MA-bearing fuel irradiation test is being planned to be started from the 2nd cycle for an early startof irradiation under the current licensing. These preliminary tests will also contribute to the MA-baring fuel technology
development. Early restart of Monju is essential to complete its missions at the earliest occasion.
Keywords
Monju, GACID, Fast reactor

34

1a-19Study on Actinide Burning by Fast Reactor
Shigeo OHKI*, Tsutomu OKUBO, Tomoyuki ABE
Japan Atomic Energy Agency (JAEA)
Advanced Nuclear System Research and Development Directorate

After the accident at Fukushima Daiichi Nuclear Power Plant, re-examination of national energy strategy including nuclear
energy has proceeded in Japan. Additionally, there are much more interests about the management of radioactive waste from thenuclear power generation.
Regarding this situation, the Fast Breeder Reactor (FBR) with its fuel cycle system has a possibility of playing the following
important roles:- Acquiring domestic energy resources is still an important issue for Japan which currently depends on most of energy
resources from abroad. Even if the power generating capacity of nuclear energy is reduced, it seems important to keep an optionof the sustainable energy resource by the plutonium (Pu) recycling in FBR cycle system.
- The FBR cycle system can also be utilized for the management of actinide nuclides (i.e. Pu and minor actinides (MAs)) in
the spent fuel. The core design flexibility enables either breeding or burning of Pu, depending on the needs. (In the latter case, thereactor is referred to as Fast Reactor (FR).) MAs can be transmuted and eventually burnt by fission in both FBR and FR systems.
Under the above-mentioned background, “the study for reducing the quantity and toxicity of radioactive waste” has also
recently been discussed in relation to the experimental planning of the prototype FBR “Monju.” Based on the accumulated
knowledge of the conventional FBR research and development, the study on actinide burning FR has been performed in JAEA.
This paper describes the present status of the actinide burning FR core design study in JAEA. The basic characteristics inreactor physics are compared between FBR and FR. Based on the result of parametric survey calculations on FR core specifications,the relationships between core performances and Pu/MA transmutation characteristics are discussed. The foreseen experimental
program using “Monju” etc. in support of the actinide burning FR core design is also explained briefly.

Keywords
Plutonium, Minor actinide, Transmutation, Fast reactor, Core design35

1a-20
Neutronic analysis and transmutation performance of Th-based Molten Salt
Fuels
Seung-Woo HONG*¹, Sang-In BAK¹, Jong-Seo CHAI¹, J. P. Curbelo4, Yacine KADI¹ ², Claudio TENREIRO¹ ³¹Department of Energy Science, SungKyunKwan University, Suwon 440-746, Korea²Engineering Department, European Organization for Nuclear Research, CERN, CH-1211 Geneva 23, Switzerland³Faculty of Engineering, University of Talca, St 720-747, Talca, Chile
4 Higher Institute of Technologies and Applied Sciences, InSTEC, Havana, Cuba

Thorium fuel and thorium fuel cycles are attractive for the long-term nuclear energy production with low radioactive waste.
The utilization of thorium in various reactors ranging from thermal to fast neutron energies has been successfully demonstrated,but substantial R&D is required before commercialization becomes possible. The molten-salt reactor (MSR) systems present avery special feature of a liquid fuel. MSR concepts, which can be used as efficient burners of TRU from spent LWR fuel, havealso a breeding capability in any kind of neutron spectrum (from thermal to fast), when using the thorium or fast spectrum U-Pu
fuel cycle. In both options, they have an interesting potential for the minimization of radiotoxic nuclear waste. We have worked on
the transmutation performance of a thorium-based Accelerator Driven System with Molten Salt Fuels. In this study, we investigate
characteristics of a subcritical ADS with Molten Salt Fuels by considering three types of fuels based on Thorium and compare theresults on neutronic analysis, transmutation of Minor Actinides, and breeding ratios, etc. The transmutation potential of different
Molten Salt fuels, the breeding potential for Th-based fuels, and the safety characteristics of such systems are discussed.
Keywords
Accelerator Driven System, Thorium, Molten Salt Fuels, Transmutation

36

1b-1

Transmutation Scenarios after Closing Nuclear Power Plants
Kenji NISHIHARA*, Kazufumi TSUJIMOTO, Hiroyuki OIGAWA
Japan Atomic Energy Agency, Japan

When we think of the end of the nuclear power generation in the future of a certain country, there will be considerable
amount of nuclear material that will be no longer burnable in the commercial nuclear power reactors. Some countries may
stop nuclear power generation after they operate only the light water reactor (LWR) with the use of the uranium fuel. They
will have depleted uranium, plutonium (Pu) and small amount of minor actinides (MA) in the spent fuel. Others may stop
after introducing a fleet of the fast breeder reactor (FBR), and they will have larger amount of Pu and MA in the FBR spent
fuel than those in the LWR one. When they stop the nuclear power generation, they have to choose among directlydisposing of the spent fuels, transporting them to the other country and burning them in the dedicated transmutation systems.
The first choice might be superior in economics aspect, but difficult to find and construct the repository owing to large sizeand possibility to recover Pu for nuclear weapon. The second one is reasonable if neighbouring countries need them, but
is not available for the last country that stops the nuclear power generation. The third one is costly but effective to minimize
the burden of repository.
In the transmutation strategy, Pu has to be transmuted prior to MA because of its amount and usability as the nuclear
weapon. For the country without the FBR technology, Pu must be transmuted in the LWR dedicated for the transmutation
in forms of the MOX fuel or the ROX fuel. The ROX fuel that stands for “rock-like oxide” fuel contains mainly Pu withless amount of 238U to reduce Pu production and to transmute about 90% of fissile Pu. The spent ROX fuel is so stable toconfine radio-nuclides and it is difficult to recover Pu after disposal. Both transmutation methods in the LWR can reduce
the emission of radio-nuclides to the environment and the opportunity to use Pu in the repository as the weapon. However,the potential radio-toxicity is not reduced due to remaining Pu and MA. If the transmutation systems such as fast reactor
or accelerator-driven system are introduced in the nuclear fuel cycle, very small amount of Pu and MA are contained in thehigh-level waste, and, the potential radio-toxicity is reduced by one to two orders of magnitude.
In the present study, above mentioned transmutation methods were investigated quantitatively in terms of necessary
number of transmutation systems and period after closing the nuclear power with the use of the analysis code for nuclear
material flow. Moreover, the comparison of repository size and its potential radio-toxicity were made along thetransmutation scenarios.

Keywords
Transmutation, Scenario study, ADS

37

1b-2
Principle Options of Fuel Debris Criticality Control in Fukushima
Daiichi Reactors
Kotaro TONOIKE*, Hiroki SONO, Miki UMEDA, Yuichi YAMANE, Teruhiko KUGO and Kenya SUYAMAJapan Atomic Energy Agency

Since March 2011, Japan Atomic Energy Agency (JAEA) has been conducting studies on criticality control
issues of the fuel debris in the damaged reactors of the Fukushima Daiichi site. The studies are both qualitative,referring the existing criticality control principles for normal nuclear facilities, and quantitative, computing expected
criticality characteristics of the fuel debris. The current situation of the reactors is also considered in the study.
In normal nuclear facilities, the goal of criticality control is to secure a subcritical condition of fissile material.
The goal is achieved by regulating composition, mass, and/or geometry of fissile material. In the accident of Three
Mile Island Unit 2 reactor (TMI-2), heavily damaged and melted fuel assemblies formed fuel debris of a large
amount whose geometry is uncertain, whose mass is larger than the minimum critical mass, and whose compositionis unknown except the possible highest 235U enrichment. Moreover, the fuel debris had to be cooled and shieldedby water. To overcome this highly uncertain situation, the coolant water was borated with a concentration morethan 4350 ppm that was enough to secure the subcritical condition.
The situation of Fukushima Daiichi reactors is severer than TMI-2. Most major difference from the viewpointof criticality control is that the coolant water flow is practically "once through". Boron must be endlessly addedin the coolant water to keep an appropriate concentration, which is not feasible. The water is not borated relying
on the circumstantial evidence that the xenon gas monitoring in the containment vessels does not show a sign ofcriticality. Although the fuel debris will not be touched for a while, its condition may change due to gradual dropof its temperature or geometry change by aftershocks. The condition can be intentionally changed when the
retrieval work of fuel debris starts. Such condition change may lead criticality of the fuel debris.
To avoid the criticality, or the environmental impact due to the criticality, a certain principle of criticality control
must be established. There may be several options: prevention of criticality by the coolant water boration,
prevention of criticality by the neutronic monitoring, prevention of environmental impact due to criticality, etc.
Every option has both merits and demerits.
It is necessary to understand the actual condition of the fuel debris toward its certain criticality control bychoosing an appropriate principle from those options. Observation, sample taking and analysis of the fuel debris
must be conducted adequately.
Keywords
Fukushima Daiichi, Fuel debris, Criticality control

38

1b-3
ADS Research Activities in Sungkyunkwan University
Seung-Woo HONG*¹, Sang-In BAK¹, Masoud BEHZAD¹, Jong-Seo CHAI¹, Yacine KADI¹ ², Vijay K. MANCHANDA¹, TaeSun PARK¹, Claudio TENREIRO¹ ³, Chirag K. VYAS¹¹Department of Energy Science, SungKyunKwan University, Suwon 440-746, Korea²Engineering Department, European Organization for Nuclear Research, CERN, CH-1211 Geneva 23, Switzerland³Faculty of Engineering, University of Talca, St 720-747, Talca, Chile

We present research activities on the Thorium based Accelerator Driven System (ADS) conducted in the Department of Energy
Science at Sungkyunkwan University. The activities include computational modeling, partitioning and transmutation, reactor core
design, and life cycle assessment analysis. For the computational modeling, we have been developing nuclear data library basedon the ENDF, which is a new prototype of a nuclear data manipulation package (TNudy) based on the ROOT system(http://root.cern.ch).. An accelerator-based production of radioisotopes and transmutation has also been studied based on theconcept of the TARC (Transmutation by Adiabatic Resonance Crossing) which is a neutron converter target. Thermodynamic
analyses of a multi-channel helium-cooled device are performed with the Computational Fluid Dynamics (CFD) code CFX.
Designs of compact reactor cores, with and without MA fuel elements have been studied. A new protocol has been developed torecover 90 Sr from irradiated 232Th. As a strategy for the management of high level nuclear waste actinide partitioning studies havebeen conducted using ‘Green’ extractants. For the purpose of assessing the possible acceptance of the ADS technology by oursociety, we have been conducting Life Cycle Assessment analyses for estimating the cost of the conventional and modified fuelcycles of nuclear power. The presentation will describe the current activities in these research fields.
Keywords
ADS, Thorium, Nuclear Data, Partitioning & Transmutation, Life Cycle Assessment

39

1c-1
Nuclear transmutation of long-lived nuclides with laser Compton scattering :Quantitative analysis by theoretical approach
Shizuka TAKAI1, Kouichi HAGINO2
1

Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1115, Japan
Department of Physics, Tohoku University, Sendai, Miyagi 980-8578, Japan2

One of the major problems of fuel cycle is the disposal of high-level radioactive waste which contains long-lived nuclides
such as 129I and high-decay heating nuclides such as 137Cs. After the severe accident of Fukushima Daiichi Nuclear Power Plant,
there is also a problem of 137Cs having been concentrated by treatment of contaminated water. Transmuting such nuclides into
short-lived or stable nuclides is one possible way to resolve this problem. Neutron capture reactions have been proposed fortransmutations of such fission products. However, neutron capture cross sections differ significantly from nuclide to nuclide, andthis transmutation method is not effective for nuclides with small neutron capture cross sections such as 137Cs.
Recently, the photo-neutron (γ,n) reactions with laser Compton scattering γ-rays has been suggested as an alternative
method for nuclear transmutations. The main point of this transmutation is to use γ-rays generated by laser photons
backscattered off GeV electrons as well as photonuclear reactions via electric giant dipole resonance (GDR), which in general
has a large cross section. The GDR is a collective excitation of a nucleus which decays mainly by the emission of neutrons andits total cross section is a smooth function of mass number. Therefore, this method is expected to be effective for transmuting
fission products regardless of isotopes.
In this paper, we investigate in detail the transmutation of 137Cs which is one of high-decay heating nuclides and has a small
neutron capture cross section. We calculate (γ, γ), (γ,n) and (γ,2n) reaction cross sections for 137Cs, 136Cs, 135Cs based on the
Hauser-Feshbach theory using the TALYS code and evaluate quantitatively the effectiveness of this transmutation method.
Since the (γ,2n) reaction transmutes 137Cs (T1/2= 30 y) into 135Cs (T1/2= 2.3×106 y), we optimized the γ-ray energy for thetransmutation to control the (γ,2n) reaction.
Figure 1(a) shows the number of 137Cs as a function of irradiation time under various photon fluxes. The transmutation of
137Cs is effective for the photon flux over 1018/s ; irradiation with photon flux 1018/s for 24 hours results in 10% reduction of theinitial amount of 137Cs. However, achievable photon flux is 2×1012/s at present maximum accelerator systems where thereduction rate of 137Cs is two orders of magnitude smaller than the natural decay rate of 137Cs as shown in Figure 1(b). Although
the transmutation of 137Cs with laser Compton scattering is theoretically possible, it is actually difficult in the present condition.

Fig.1 (a) Dependence of the photon flux 𝑁𝑁𝛾𝛾 =1012, 1018, 1019, 1020/s on the reduction of 137Cs of 1g as a functionof irradiation time. (b) The numbers of isotopes when 137Cs of 1g is irradiated with photon flux 2×1012/s.
Keywords
Transmutation, Photo-neutron reaction, Laser Compton scattering

40

1c-2
Sensitivity analyses of initial compositions and cross sections for activation
products of in-core structure materials
Kento YAMAMOTO1*, Keisuke OKUMURA1, Kensuke KOJIMA1, Tsutomu OKAMOTO1**
1

Japan Atomic Energy Agency
Special topic researcher**
Collaborating engineer*

Improving the accuracy of predicting concentrations of activation products is important in the field of nuclear
back-end for various evaluations. A quantitative investigation about the elements and the nuclear reactions leading
to generation of activation products is of much help to achieve it. To clarify quantitatively the source elements andthe nuclear reactions dominating generation of activation products, sensitivity analyses of initial compositions andcross sections were conducted.
In the analyses, ORIGEN2.2 code was used with ORLIBJ40, which is a set of the 1-group cross section libraries
based on JENDL-4.0. A sensitivity coefficient is defined as the ratio of the amount of the concentration change ofthe target activation product to the amount of the change of an initial composition or a cross section. Activations of
cladding tubes, end-plugs and spacers of fuel assemblies and channel boxes in BWR, whose materials are zirconium
alloy, SUS304 stainless steel, and nickel-chromium-based alloy, were analyzed. The typical BWR irradiation history
and flux level were applied in these analyses. From about 50 representative radioactive nuclides, several nuclides
were selected as the targets of sensitivity analyses because of their large concentrations in the target materials.
The results clarified quantitatively the source elements and the nuclear reactions dominating generation ofactivation products. It was remarkable that the dominant generation pathways were clarified even for the nuclides
generated through complicated pathways. These examples are shown in Figure 1 and Figure 2. In conclusion, theresults of sensitivity analyses could be utilized to select the objective of elements for measurements of impurities in
the materials and of nuclear data for improvement of accuracy.

S-34

(n,γ)

S-35

Fe-54

β- (87.3day)
Cl-35

(n,γ)

(n,γ)
0.986

Cl-36

Figure 1 An example of an activation productgenerated through two nuclear reactions and anuclear decay (Cl-36 in SUS304).

Fe-55
(n,α)
0.011

Ni-58

Figure 2 An example of an activation product withtwo contributing generation pathways (Fe-55 in
SUS304). The values in the figure show thesensitivity coefficients of the cross sections.

Keywords
activation products, sensitivity study, burnup calculation, ORIGEN2.2, ORLIBJ40

41

1c-3

Thermophysical properties of thorium oxide
Hiroaki MUTA*1, Toshiaki KAWANO¹, Yuji OHISHI1, Ken KUROSAKI 1, Shinsuke YAMANAKA1,2
1
2

Graduate School of Engineering, Osaka University
Research Institute of Nuclear Engineering, University of Fukui

Thorium oxide (ThO2) has been considered as a nuclear fuel for various reactors, such as light water reactors, heavy water
reactors, high-temperature gas-cooled reactors, and accelerator-driven systems. Additionally, ThO2 is a matrix candidate forplutonium incineration. Numerous studies have been undertaken on the physical properties of (Th,U,Pu)O2 as loading fuels of
initial state. FP behavior has also been evaluated through the synthesis and observation of ThO2-based SIMFUEL. However, theeffects on the thermopysical properties have been rarely evaluated. In the present study, high-density pellets of FP element–
containing ThO2 were prepared using the Spark plasma sintering (SPS) method, and the effects on lattice parameter and thermal
conductivity were investigated. These effects, including the oxygen defect, were quantitatively evaluated and are applicable toadiabatic FP elements.
The powder of Th1-xMxO2-y (M=Y, La, Ce, Nd, Gd, U; x=0.05, 0.10) were synthesized by solid-state reaction. The raw powders
were mixed by wet ball-milling for 12 h and then cold pressed and heated at 1873 K for 72 h in a 4% hydrogen/argon gas flow.
The powders were then ball-milled for another 12 h, packed to graphite die, and sintered in an SPS apparatus in a flow of argon
gas. The sintering temperature and pressure were 1873 K and 50 MPa, respectively, for all the samples. The sintering temperature
was controlled by the electric current. After the sintering, the samples were heat-treated to adjust the M/O ratio. The sample
compositions are assumed be Th1-xMxO2 for M=Ce, U, and Th1-xMxO2-x/2 for M=Y, La, Nd, and Gd. This means that all Ce, Th,
and U ions are tetravalent and other rare earth elements are trivalent in the samples.
The relative sample density reaches above 90% of the theoretical density. ThO2 is known to be sintering-resistive, it generally
requires a special granulation technique, sintering additives, and a high sintering temperature for fabrication of the high density
pellets. Here, relatively high density pellets are obtained by applying the SPS technique at moderate sintering temperature andwithout any additive.The lattice parameter appears to change linearly with mol fraction of dopants for all the samples.Trivalentelements, such as Y, La, Nd, and Gd additions, significantly decrease the thermal conductivity near room temperature. Thisdecrease is caused by the phonon-point defect scattering by the solute ions in addition to the scattering by the oxygen defect.
Klemens and Callaway’s model is used to estimate the lattice thermal conductivity for solid solutions. From the analysis, we canestimate the thermal conductivity of various ThO2 that have rare earth elements added.
Keywords
Thorium oxide, FP, Thermophysical properties

42

1c-4
Characterization and thermophysical properties of RE2Zr2O7 and Nd2Ce2O7
precipitate in ThO2-based fuelToshiaki KAWANO*1, Hiroaki MUTA1, Yuji OHISHI1, Ken KUROSAKI 1, Shinsuke YAMANAKA1,2
1
2

Graduate School of Engineering, Osaka University
Research Institute of Nuclear Engineering, University of Fukui

Thorium fuel cycle is contributed to reduce radioactive waste in the spent fuel because of the less long lived minor actinides
yielding than Uranium-Plutonium fuel cycle. Additionally the fuel can be applied to accelerator driven system for nuclear
transmutation. In spite of these merits, there are few reports on ThO2-based fuel, especially focusing on the FPs chemical state andinfluences on fuel properties, as compared to UO2. As for precipitate FPs in ThO2-based SIMFUEL, pyrochlore-type compound
Nd2(Zr,Ce)2O7, which is not observed in irradiated UO2 and (U,Pu)O2 fuel, were detected. The existence of precipitates is possible
to affect the fuel properties greatly, therefore, it is essential to comprehend the FPs behavior in nuclear fuel exactly. However, thephysical properties and the precipitate mechanism are not elucidated completely.
In this study, the pyrochlore oxides RE2Zr2O7 (RE=Y, La, Pr, Nd, Eu, Gd, Dy, Er) and Nd2Ce2O7 were synthesized by solidstate reactions and pelletized by spark plasma sintering in order to make the high-density pellets. We examined the phase states
and lattice parameters of them by means of X-ray diffraction and SEM/EDX analysis. The lattice parameters of RE2Zr2O7
depended on the ionic radii of rare earth ion. The heat capacity, thermal conductivity and coefficient of linear thermal expansion
were also measured. It was confirmed that the thermal conductivities were independent of temperature and significantly lower thanthose for ThO2. The elastic constant and Vickers hardness was also investigated. In addition, to reveal the precipitate mechanismof pyrochlore-type compound, solid solutions between ThO2 and Nd2O3, ZrO2, were synthesized by solid-state reactions
respectively and annealed under controlled oxygen potential, -60, -210,-260, -310 kJ. The phase states of the obtained samples
were estimated by SEM/EDX analysis and lattice parameter.
Keywords
Thorium oxide, FP, Pyrochlore-type compound

43

1c-5
Overview of Studies on Lead-bismuth Technology in Tokyo Institute ofTechnology
Minoru TAKAHASHI*¹, Yun GAO2, Marion GUIHOT1, Asril PRAMUTADI2¹Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology²Department of Nuclear Engineering, Tokyo Institute of Technology

The early phase of experimental study using lead alloy in our laboratory was initiated by simulation of thermal-hydraulicphenomena in molten fuel-coolant thermal interaction. Wood’s metal (48Bi-26Pb-13Sn-13Cd) was used for the experiment. It was
performed from 1975 till 1980. One of the experiments was a vapor explosion test in the discharge of a jet of hot liquid metal intoa cold water pool. Another one was a basic test of interfacial heat transfer and boiling in hot Wood’s metal-water direct contact
stratified flow in a horizontal duct. for the development of the accelerator driven system (ADS) and lead alloy-cooled fast reactors(LFR) Since 1999, basic experimental studies on the technology of lead-bismuth eutectic (LBE, 45Pb-55Bi) have been performed
for the development of ADS and lead-bismuth-cooled fast reactor (LFR). Based on the studies of the LBE technology, aninnovative concept of LBE, i. e. PBWFR (150MWe/450MWt) was formulated and proposed. The present paper overviews thebasic studies on the LBE technology that have been performed in our laboratory. The subjects of the studies are classified into thefollowing items:(1) Compatibility of structural and cladding materials with LBE- Corrosion of existing steels and steels containing Si and Al- Corrosion of refractory metal-coated and Fe-Al alloy-surface-coated steels- Corrosion of ceramic materials- Corrosion of steels under bending and tensile stresses- Corrosion of cold worked steels and welded steels- Erosion of steel in LBE flow with low oxygen concentration(2) Oxygen control oxygen sensor for LBE flow circuit- Oxygen control using Ar-H2-H2O gas mixture and solid PbO- Oxygen sensor with air reference and the reference of Bi and Bi oxide mixture(3) Gas-LBE two-phase flows- Natural circulation of LBE in single-phase flow
- 2-d Ar gas bubbly two-phase flow in LBE in a rectangular duct
- Visualization of N2 gas-LBE bubbly two-phase flow with gamma ray radiography- Ar-LBE and N2 gas-LBE two-phase flow
- Removal of LBE particles in gas flow with Chevron dryer and electrostatic precipitator(4) Phase changes in LBE systems- Direct contact boiling two-phase flow of water in hot LBE flow
- Violent boiling in direct-contact of hot LBE droplet in cold water- Freezing of lead-alloy on cold metals(5) Property of LBE- Diffusion coefficients of metal impurities in LBE
Keywords
Lead-bismuth eutectic, Corrosion, Oxygen Sensor, Thermal-hydraulics, Particle Flow, Diffusion Coefficient

44

1c-6
Evaluation of structural integrity of beam window for TEF target bythermal/fluid-structure interaction analysis
Hironari OBAYASHI*, Hayanori TAKEI, Hiroki IWAMOTO, Toshinobu SASA
Japan Atomic Energy Agency

Japan Atomic Energy Agency (JAEA) has been researching and developing an accelerator-driven system (ADS) as a
dedicated system for the transmutation of long-lived radioactive nuclides. The ADS proposed by JAEA uses the lead-bismuth
eutectic (LBE) alloy as a target material and a coolant. In the various R&D for ADS, construction of the Transmutation
Experimental Facility (TEF) is planned under the framework of the J-PARC project as a preceding step before the constructionof demonstrative ADS. In the development of the TEF, the estimation of the thermal-fluid properties of the flowing LBE is an
important issue, because the LBE has the tendency to cause the corrosion/erosion to the structural materials. And the beam
window of ADS, which separates the proton accelerator and the LBE subcritical core vessel, is exposed to the high temperatureenvironment induced by the incidence of proton beam, and the embitterment of materials by the irradiation occurs. Therefore,
the feasibility of the beam window is the most important factor for the realization of TEF. The objective of this study is toevaluate the feasibility of a designed beam window of TEF target by the numerical analysis with a 3D model. The analysis was
performed by considering (1) the current density and shape of the proton beam, (2) the thermal-fluid behavior of LBE aroundthe beam window as a function of the flow rate and inlet temperature, (3) the material and the thickness of the beam window,(4) the structural strength of the beam window.
In the reference case, the current density and the profile of the proton beam were 20 μA/cm2 and a Gaussian shape,
respectively. The flow rate of LBE and temperature at the inlet were 1 ℓ /sec and 350 ºC. The material of a beam window wastype 316 stainless steel with the 2 mm thick. In this case, the maximum velocity of LBE and the maximum temperature locatedat the top of the beam window were about 1.2 m/sec and 477 ºC. By increasing the flow rate of LBE up to 4 ℓ /sec, themaximum temperature of a beam window was reduced around 420 ºC. The maximum shear stress was 190 MPa, which wasobserved at the center on the outside surface of a beam window. The value of analyzed stress in the reference case was lower
than the tolerance level of the stress strength of the material given by the legal limitation which is applied to the fast reactor, andhence the feasibility of a designed beam window was confirmed.
Keywords
Accelerator-Driven System (ADS), Lead Bismuth, Beam window, Star-CD, ABAQUS,

45

1c-7
Fission cross-section measurement of minor actinides using a leadslowing-down neutron spectrometer KULS at KUR
K. HIROSE*¹,a, T. OHTSUKI¹,b, Y. SHIBASAKI¹, N. IWASA2, J. HORI3, S. SEKIMOTO3, K. TAKAMIYA3,
H. YASHIMA3, K. NISHIO4, and Y. KIYANAGI5¹Research Center for Electron Photon Science, Tohoku University²Department of Physics, Tohoku University³Research Reactor Institute, Kyoto University
4 Advanced Science Research Center, Japan Atomic Energy Agency
5 Graduate School of Engineering, Hokkaido University
a Present address: Advanced Science Research Center, Japan Atomic Energy Agency
b Present address: Research Reactor Institute, Kyoto University

Management of minor actinides (MAs) and long-lived fission products is one of the most important problems for the use of
nuclear power. Accurate data for neutron-induced reactions are required for nuclear transmutation technique of minor actinides
using an accelerator-driven subcritical system or other innovative reactors.
We have performed a cross-section measurement of fission of MAs (Np, Am and Cm isotopes) using a lead slowing-downneutron spectrometer (KULS) at Kyoto University Research Reactor Institute. KULS is a neutron spectrometer assembled withlead blocks (1.5x1.5x1.5 m3) with a neutron production target in the center. Neutrons produced by irradiating the target withelectron beams are slowed down in the lead material and finally reach to a sample placed in a hole provided in KULS. Theenergy of the slowed neutron is determined by a relation 𝐸𝐸𝑛𝑛 = 𝐾𝐾⁄(𝑡𝑡 − 𝑡𝑡0 )2, where K is a slowing-down parameter andcalibrated as 178 keVµs2. Because of a short distance between the production target and the sample, typically 40 cm, KULS
gives a high flux neutron field and enables us to measure a small cross-section with small amount of sample.
The fission cross-sections were measured using a
multi-layered parallel plate avalanche counter [1] in which
two samples can be set back-to-back. Fission fragments froma sample of MA and a 235U sample were simultaneously
detected. Then, the fission cross-section of MA can beobtained relatively to that of 235U. In Fig.1, the results for
241Am obtained in the present work is compared with theother works [2,3]. The solid line is the evaluation of
JENDL-4.0 [4] which is broadened according to the neutronenergy resolution of KULS.
In the symposium, the results for 237Np, 241, 242m, 243Am
and 245, 248Cm will be presented.
[1] K.Hirose et al., Nucl. Instr Meth A621 (2010) 379.
[2] S.Yamamoto et al., J. Nucl. Sci. Eng. 126 (1997) 201.
[3] V.F.Gerasimov et al., Dubna Reports, No.97 (1997) 348.
[4] K.Shibata et al., J. Nucl. Sci. Technol. 48 (2011) 1.

Fig. 1 The cross-section of 241Am(n,f) obtained in thepresent work.

Keywords
Minor actinides, fission cross section, lead slowing-down neutron spectrometer

46

1c-8

Development of TRU nitride database for designing ADS fuelTsuyoshi NISHI*, Yasuo ARAI, Masahide TAKANO, Hirokazu HAYASHI, Masaki KURATA
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency

Spent fuels arising from nuclear power plant contain significant amounts of long-lived minor actinides (MA: Np, Am, Cm)
and thus, proper management of MA is one of the key issues for sustainable nuclear energy use. Nuclear transmutation
technology using sub-critical Accelerator Driven System (ADS) potentially offers a promising option to overcome the concern
in the future nuclear fuel cycle [1]. Nitride solid solutions containing MA and Pu, hereafter called as transuranium (TRU)
nitrides, are proposed as a candidate material of ADS fuel, in which TRU nitrides are used by diluting with ZrN. It is necessary
to develop the property database of nitrides for designing ADS fuel. Experimental data or estimated values of extensive
properties for TRU nitrides are being summarized in Japan Atomic Energy Agency (JAEA). Also, from a view point ofpracticality for fuel design, the properties are being formulated, in which error evaluation is also made as much as possible.
Several important properties on TRU nitrides have been measured recently in JAEA. Nevertheless, some property data for
TRU nitrides are still often lacking. In our database, the properties of uranium nitride, UN, and uranium-plutonium mixed
nitride, (U,Pu)N, are substitutionally given in such cases. The data for UN are mostly cited from the efforts in USA [2-7], inwhich UN had been examined as a potential fuel for space reactors. The data for (U,Pu)N are mostly cited from two
distinguished monographs by Matzke [8] and Blank [9] of Institute for Transuranium Elements (ITU), in which (U,Pu)N as
well as (U,Pu)C had been investigated in detail as an advance fuel for fast reactors. Besides, a few property data on TRU
nitrides had been collected under the Advanced Accelerator Assisted (AAA) and the Advance Fuel Cycle Initiative (AFCI)
programs of USA [10-11].
At present, most of the thermophysical properties on TRU nitrides, such as thermal conductivity [12] and lattice thermal
expansion [13], have been available with enough accuracy for practical use in fuel design of ADS. On the other hand, most of
the mechanical properties and irradiation behavior, which are still limited to those of UN and (U,Pu)N, should be examinedhereafter by experiments or reliable simulations. Further, variation in properties with burnup progressing also should be
evaluated, because high burnup (~20 at.%) is necessary for TRU-transmutation in ADS.
[1] K. Tsujimoto, et al., J. Nucl. Sci. Technol., 41 (2004) 21.
[2] S.L. Hayes, et al., J. Nucl. Mater., 171 (1990) 262.
[3] S.L. Hayes, et al., J. Nucl. Mater., 171 (1990) 271.
[4] S.L. Hayes, et al., J. Nucl. Mater., 171 (1990) 289.
[5] S.L.Hayes, et al., J. Nucl. Mater., 171 (1990) 300.
[6] E.K. Storms, J. Nucl. Mater., 158 (1988) 119.
[7] S.B. Ross, et al., J. Nucl. Mater., 170 (1990) 169.
[8] Hj. Matzke, Science of Advanced LMFBR Fuels, North-Holland, Amsterdam (1986).
[9] H. Blank, in: Materials Science and Technology, vol. 10A, Weinheim, New York (1994) p.191.
[10] Y.S. Kim, et al., US-Report, ANL-AAA-068 (2003).
[11] B. Feng, et al., J. Nucl. Mater., 427 (2012) 30.
[12] T. Nishi, et al., J. Nucl. Sci. Technol., 48 (2011) 359.
[13] M. Takano, et al., J. Nucl. Mater., 376 (2008) 114.
Keywords
Accelerator-Driven-System (ADS), TRU nitride database, Thermophysical property

47

1c-9
Recent progress in research and development on the neutron resonancedensitometry (NRD) for quantification of nuclear materials in particle-likedebrisM. KOIZUMI1, F. KITATANI1, H. TSUCHIYA1, H. HARADA1, J. TAKAMINE1, M. KURETA1, H. IIMURA1, M. SEYA2,
B. BECKER3, S. KOPECKY3, W. MONDELAERS3, P. SCHILLEBEECKX3
1

Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki, 319-1195
JAPAN
2Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai-mura,Naka-gun, Ibaraki, 319-1118 JAPAN
3
European Commission, Joint Research Centre, Institute for Reference Materials and Measurements, Retieseweg 111, B - 2440
Geel, BELGIUM

Quantifying nuclear materials (NM) in stock of debris of melted fuel (MF) formed in a severe accident is considered to bedifficult because of their variety of size, shape, unknown compositions and their strong radio activity. Although, techniques of
non-destructive assay (NDA) are indispensable for evaluation of NM in debris, quantification methods have not been
established so far. In the cases of TMI-2 or Chernobyl-4, nuclear material was accounted based on some estimation.
We have proposed a technique called neutron resonance densitometry (NRD) to quantify NM in particle-like debris of MF.
We assume particle-like debris is produced in rapid cooling processes of severe accidents. Small pieces are alto produced when
MF are cut or broken down to be taken out of the damaged reactors.
NRD is a method of a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture
analysis (NRCA) or prompt gamma-ray analysis (PGA). In NRTA, neutron transmission rate is measured as a function ofneutron energy with a time-of-flight (TOF) technique. Characteristic neutron absorption peaks of Pu and U isotopes areobserved in the neutron energy range of 1-50 eV. Measurements of these transmission spectra, therefore, can be carried out witha short-flight-path TOF system. The NRTA measurements are not interfered by strong gamma-ray radiation from MF samples.
However, composing isotopes with large total cross-section (H, B, Cl, Fe, etc.) distort the obtained neutron absorption spectra.
Accordingly, accurate measurement could not be performed. The quantity of these containing isotopes could not be determined
by NRTA measurements, because these isotopes do not resonantly absorb neutrons in this energy range. To identify andquantity the composing isotopes, NRCA/PGA method is introduced. In NRCA/PGA, prompt-gamma rays are measured with ahigh resolution and S/N gamma detection system. Development of a special gamma-ray spectrometer for NRCA/PGA is inprogress. Information from the NRCA/PGA is used to optimize the sample thickness and measurement time, and to support
NRTA analysis.
In order to examine the NRD method, studies have started under the agreement between JAEA and EURATOM in the fieldof nuclear materials safeguards research and development. Experiments studying the effects of sample properties, such as target
thickness, particle size, temperature and the achieved accuracy, are going on at the TOF-facility GELINA of EC-JRC-IRMM.
In this presentation, we briefly describe the concept of NRD, overview the research project of development of NRD, andexplain the recent progress of the study. This research was supported by JSGO/MEXT.
Keywords
Neutron resonance densitometry, NRD, transmission, capture, Fukushima, severe accident, nuclear security48

1c-10Status of Studies and Future Subjects of Material Corrosion for Leadbismuth Cooling System
Marion GUIHOT*, Minoru TAKAHASHI
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

One of the most crucial issues in the development of lead alloy-cooled reactors and accelerator-driven transmutation system isthe compatibility of structural materials with corrosive lead-bismuth eutectic at high temperature up to 650°C. In this paper, thecurrent studies about static corrosion in lead-bismuth on various types of steel, in which different stresses are applied (heattreatments, strengths…) and under different experimental conditions (temperature, oxygen concentration…) are reviewed.
Corrosion kinetics and mechanisms in lead-bismuth are explored. We first examine the importance of oxygen concentration andthe different means of measure and control. We then discuss the influence of oxygen concentration and temperature on corrosion
and the effect of stresses on the surface of corrosion. Current apparatus are described and existing experimental results are
summarized. Finally, based on the present results, some recommendations for future experiments are proposed to complete theknowledge about corrosion in lead-bismuth coolant system before its industrial applications.
Keywords
Lead-bismuth eutectic, Corrosion, Oxygen control

49

1c-11
Mechanical properties of beam window materials for ADS irradiated in a
spallation environment
Shigeru SAITO¹, Kenji KIKUCHI², Dai HAMAGUCHI1, Shinya ENDO³, Kouji USAMI3, Naotoshi SAKURABA3, Hiromitsu
MIYAI3, Katsutoshi ONO3, Hiroki MATSUI3, Masayoshi KAWAI4, Yong DAI5
¹Transmutation section, J-PARC Center, JAEA
²Frontier Research Center, Ibaraki Univ.
³Dept. of Fukushima Technology Development, JAEA
4Institute of Materials Structure Science, KEK
5Spallation Source Division, PSI

In several institutes, the research and development for an accelerator-driven transmutation system (ADS) to transmute minor
actinide (MA) have been progressed. ADS is composed of a high-intensity proton accelerator, a spallation target and a sub-critical
reactor core loaded with MA fuel. As a coolant and a target material, LBE (Led-bismuth eutectic) is selected. The spallation target
is placed at the center of the sub-critical reactor core. The proton beam accelerated at the high-intensity accelerator is injected into
the target. The sub-critical core is driven by the spallation neutrons produced in the target and MA is transmuted to short-lived
isotopes. The bean window of ADS will be used as the boundary between the proton accelerator and the reactor core. The beam
window submerged in the reactor core will be subjected to high-energy proton and spallation neutron irradiation. To evaluate thelifetime of the beam window, post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss
spallation neutron source) specimens has been carried out.
The specimens tested in this study were made from the austenitic steel JPCA (Japan Primary Candidate Alloy). JPCA is a
modified austenitic stainless steel by the addition of Ti to AISI type 316 SS in order to improve swelling resistance. The specimens
were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as
follows: the proton energy was 580 MeV, irradiation temperatures ranged from 100 to 430℃and displacement damage levels
ranged from 7.0 to 19.5 dpa. Tensile tests were performed in air at room temperature (R.T.), 250℃and 350℃. Bend-fatigue tests
were performed in air at room temperature under deflection control mode. The wave form of the control signal was a sine curvewith a frequency of 26 Hz. Fracture surface observation after the tests was done by SEM.
Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission
neutron irradiated data. At the higher temperatures, 250℃and 350℃, the extra hardening was not observed. Degradation of ductility
bottomed around 10 dpa, and specimens kept their ductility until 19.5 dpa. All specimens fractured in ductile manner. Results ofthe bend-fatigue tests showed that the numbers of cycles to failure (Nf) were not changed by irradiation. Dpa dependence of Nf
was not clearly seen in the irradiation conditions. In spallation environment, He atom production ratio is very high and most of He
atoms are retained in the materials. In spite of large number of He atoms, all JPCA specimens show transgranular fracture surface.

Keywords
ADS, Beam window, Austenitic steel, JPCA, Proton irradiation, Tensile property, Bend-fatigue property501c-12Development of non-destructive assay to fuel debris of Fukushima Daiichi
NPP (1) Experimental validation for the application of a self-indication
methodJun-ichi HORI*, Tadafumi SANO, Yoshiyuki TAKAHASHI, Hironobu UNESAKI, Ken NAKAJIMA
Research Reactor Institute, Kyoto University

It is surmised that melted fuel debris are existing in the core at units 1, 2 and 3 of Fukushima Daiichi NPP. Identifying thefuel debris status in the reactors is one of the most important issues for decommissioning. Therefore, we need to study how toanalyze the properties of actual debris collected in advance of removal work. As debris contain melted fuel, cladding tube andstructure materials in addition to a salt content heterogeneously, non-destructive assaying the distribution of nuclear material indebris is absolutely essential for nuclear material accountancy and critical safety.
Neutron Resonance Densitometry (NRD) based on Neutron Resonance Transmission Analysis (NRTA) and Neutron
Resonance Capture Analysis (NRCA) with the Time-of-Flight (TOF) technique is a promising way to characterize debris.
However, there are two difficulties to apply the method to fuel debris. In the NRTA, many resonances of other nuclides which
are contained in debris may make it difficult to identify and quantify the target nuclide. In the NRCA, it is expected that theintense decayed gamma rays from debris give rise to high background and large dead time of gamma-ray detector. In this work,
we proposed a new concept of “self-indication method” as a complementary assay to overcome those difficulties. In the
self-indication method, we set an indicator consisted of target nuclide with a high purity beam-downstream from a sample. By
detecting the reaction products such as neutron capture gamma rays or fission products from the indicator with the TOF method,
the transmission neutron can be measured indirectly. The self-indicator is a transmission neutron detector which has a highefficiency around the objective neutron resonance energies of target nuclide, so that it enables us to quantify effectively theamount of resonance absorption of the target nuclide. Moreover, it is hard to be affected by the decayed gamma rays from thedebris.
In order to verify the method, we performed the experiments using a 46-MeV electron linear accelerator at the Kyoto
University Research Reactor Institute. A sample and an indicator were set at distances of 11.0 and 12.7 m from the neutron
source, respectively. Pulsed neutron beam was well-collimated to 24 mm in diameter. The indicator was surrounded by adetection assembly Bi12Ge3O12 (BGO) scintillators, which consists of 12 scintillation bricks of each 5×5×7.5 cm3. At first, wemeasured the densities of gold foils with 10, 20, 30, 40, 50 µm in thickness by area analysis at the 4.9-eV resonance region. A
10B plug with 8 mm in thickness or a gold foil with 50 µm in thickness was used as an indicator. Since the former thickindicator can absorb almost neutrons with energies below epi-thermal region, it was equal to the conventional NRTA. In thelatter case, it was the self-indication measurement. It was confirmed that the densities of the target nuclide can be determined byboth methods within 3% accuracy. At the next step, we added a silver foil with 50 µm to a gold foil with 10 µm as a sample. It
is worth noting that silver has a large resonance at 5.0 eV close by 4.9-eV resonance of 197Au. It was experimentally shown thatthe contribution from the other nuclide can be remarkably suppressed by applying the self-indication method. Finally, we havedemonstrated non-destructive assay to nuclear material using a mixture composed of natural uranium foils, sealed minor
actinide samples of 237Np and 243Am. In the symposium, the results of those experiments will be presented.
This work is supported by JSPS KAKENHI 24760714.
Keywords
fuel debris, pulsed-neutron source, TOF, self-indication method, neutron resonance absorption

51

1c-13Development of non-destructive assay to fuel debris of Fukushima Daiichi
NPP (2) Numerical validation for the application of a self-indication method
Tadafumi SANO*, Jun-ichi HORI, Yoshiyuki TAKAHASHI, Hironobu UNESAKI, Ken NAKAJIMA
Kyoto University Research Reactor Institute

In order to perform decommissioning of Fukushima daiitchi NPP safely, it is very important to measure the component of fueldebris. Therefore, a new non-destructive assay to identify and quantify the target nuclide in fuel debris using a pulsed neutron
source is under development in Kyoto University Research Reactor Institute.
We use the self-indication method for the non-destructive assay. This method is one of neutron transmission method. The
neutron transmission method is focused on resonance reactions (i.e. capture, fission) at the target nuclide. In the conventional
neutron transmission method, a sample is irradiated by a pulsed neutron beam and the energy distribution of transmitted neutrons
from the sample is measured by the Time-of-Flight technique. Then, the target nuclide in the sample is identified and quantified
by using transmitted neutrons in resonance energy region. This is a remarkably effective method to identify and quantify the target
nuclide. However, if the energy spectrum of transmitted neutron has many dips caused by resonance reaction of other nuclides, it
is difficult to identify and quantify the target nuclide in the sample.
In the self-indication method, the transmitted neutrons from the sample are injected into an indicator. The indicator is consistedof a high purity target nuclide. The transmitted neutrons are obtained via resonance reactions in the indicator. The self-indication
method has high signal to noise (S/N) ratio than the conventional method.
In this study, numerical validation for the self-indication method to identify and quantify nuclides in a BWR-MOX pellet isdescribed. The burnup of MOX pellet is 0GWd/t, 10GWd/t, 20GWd/t, 30GWd/t, 40GWd/t and 50GWd/t. The 12m measurement
line in KUR-LINAC is simulated as a calculational geometry. Numerical calculations are carried out by continuous energy MonteCarlo code MVP2[1] with JENDL-4.0[2] as the nuclear data library. The burnup calculations of BWR-MOX pellet are performed
by deterministic neutronics code SARC2006 [3] with JENDL-4.0.
For example, a numerical result to identify I-129 in the MOX pellet is described. The bunup of MOX pellet is 20GWd/t. I-129
has only 4 resonances in the energy region of 0.1eV to 100eV. The resonance peaks are 41eV, 73eV, 75eV and 97eV. The numericalresult shows that the transmitted neutrons are easily obtained via I-129 resonance absorption reactions in the indicator.
Numerical validation for application of self-indication method is carried out. As the results, it is described that the self-indication
method has good S/N than the neutron transmission method to quantify amount of target nuclides in the fuel debris.
A part of this study is supported by JSPS KAKENHI 24760714.
[1] Y. Nagaya, et. al., MVP/GMVP II: General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations
based on Continuous Energy and Multigroup Methods, JAERI 1348, Japan Atomic Energy Resarch Institute (2005),[2] K. Shibata, et. al., "JENDL-4.0: A New Library for Nuclear Science and Engineering," J. Nucl. Sci. Technol. 48(1), 1-30
(2011),[3] K. Okumura, et al, “SRAC2006: A Comprehensive Neutronics Calculation Code System,” JAEA-Data/Code, 2007-004,(2007).
Keywords Self-indication method, Non-destructive assay, Resonance, MOX pellet, Numerical validation52

1c-14
Precise Measurements of Neutron Capture Cross-Sections for LLFPs andMAs
Shoji NAKAMURA et al.
Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195

To evaluate the feasibility of development of nuclear transmutation technology and advanced nuclear system, precise nuclear
data of neutron capture cross-sections for long-lived fission products (LLFPs) and minor actinides (MAs) are indispensable.
Precise nuclear data for stable isotopes of LLFPs are also needed because of their relatively large fission yields.
In the poster session, we will present our research activities for the measurements of the neutron capture cross-sections for LLFPs
and MAs.

Keywords
neutron capture cross-section, minor actinides, long-lived fission product53

1c-15Development of the Method to Assay Hardly Measurable Elements in Spent
Nuclear Fuel and Application to BWR 9×9 fuelKenya SUYAMA¹, Gunzo UCHIYAMA¹, Hiroyuki FUKAYA1, Miki UMEDA1, Toru YAMAMOTO2, Motomu SUZUKI2¹Japan Atomic Energy Agency²Japan Nuclear Energy Safety Organization

In fission products in used nuclear fuel, there are several stable isotopes which have large neutron absorption effect. For theevaluation of the neutronics characteristics of the nuclear reactor, amount of such isotopes should be evaluated by using burnup
calculation codes. To confirm the correctness of such data obtained by calculation codes is important to assure the precision of
the evaluation of the neutron multiplication factor of used nuclear fuel. However, it is known that there are several hardly
measurable elements in such important fission products. The data of the amount of the hardly measurable elements in usednuclear fuel are scarce in the world.
Japan Atomic Energy Agency (JAEA) had been developed the method to assess the amount of fission products which are
hardly measurable and have large neutron capture cross section, under the auspices of the Japan Nuclear Energy Safety
Organization. In this development, the measurement method was developed, combining a simple and effective chemical
separation scheme of fission products from used nuclear fuel and an Induced Coupled Plasma Mass Spectrometry with highsensitivity and high-precision. This method was applied to the measurement program for used BWR 9×9 fuel assembly. Thismeasurement method is applicable to the required measurement for the countermeasure to the accident of the Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power Company (TEPCO). JAEA has a measurement plan for not only BWR but
also PWR fuel.
This presentation describes the measurement method developed in the study as well as the future measurement plan in
JAEA.
Keywords
Post Irradiation Examinations, Fission Products, Isotopic Composition54

1c-16
Evaluation of volatile FP elements behavior for severe accident analysis ofmolten salt reactor
M. Yamawaki 1 , T. Terai *2, T. Koyama3, Y.Arita1, Y. Sekiguchi2, K. Uozumi3 and M.Kinoshita3
Research Institute of Nuclear Engineering, University of Fukui
Department of Nuclear Engineering and Management, University of Tokyo
3
Central Research Institute of Electric Power Industry
1
2

The recent interests on molten salt reactor stem from the very high operation temperature, the passive safety feature against
unexpected core accident and possible applicability for nuclear transmutation of radioactive waste, while the retention of fission
products (FPs) during the accident event is not well understood. Especially for the volatile FPs, such as iodine and cesium, thevapor pressure nor the chemical form of the vapor species during the accident condition have not been studied yet. Hence, theauthors have started to study high temperature vaporization behaviors from the molten salt containing FPs under moisture and air.
First, the vapor pressures of compounds in gaseous phase in equilibrium with LiF-NaF-KF (46.5-11.5-42.0 mol%) molten salt
containing 3.26 mol% of CsF and 1.0 mol% of CsI were calculated using a thermodynamic database “MALT for Windows”. The
ratio of CsF to CsI was determined according to the fission yield of 233U. At the elevated temperature of 900K under 0.1MPa ofargon gas containing neither air nor moisture, the vapor pressures of the dominant gaseous compounds containing the FP were
calculated as 36 Pa for CsF and 27 Pa for CsI. In addition to these compounds, it was estimated that some other compounds also
exhibit high vapor pressures under air. Actually, the vapor pressures in air without moisture were calculated as 78 Pa for CsNO3,30 Pa for I (iodine in atomic form) and 24 Pa for I2, while those in air containing 1 mol% H2O were 177 Pa for CsNO3, 12 Pa for
I and 4 Pa for I2. Therefore, it was suggested that the vaporization of FPs from the molten fluoride salt is significantly affected by
the ambient atmosphere condition.
According to the above estimation, experiments to measure the actual vaporized amounts of iodine and cesium from the molten
fluoride salt have been planned. The molten fluoride salt containing CsF and CsI, whose composition is the same as describedabove, is to be loaded in a nickel boat. The nickel boat is placed in a 1-inch diameter tube made of nickel, in which the experimental
gas flows. The nickel tube is to be horizontally placed in an electric furnace so that the inside of the tube can be heated up to 1100K.
The vaporized material caused by the heating will flow to the downstream with the argon gas flow and be captured in scrubbers
filled with water. The content of the vaporized material is to be estimated by analyzing the amount of the material collected in the
scrubber solutions as well as in washed solutions of the apparatus, such as inside of the nickel tube and connecting tubes. Bychanging the experimental temperature (800 K though 1100 K), the experimental gas (pure argon, dry air, or wet air), and the gas
flow rate (up to 200 sccm), the behaviors of volatile FPs on severe accidents of molten salt reactors, such as leakage of the molten
fluoride salt, are to be be evaluated.
In addition, equilibrium vapor pressure measurement is to be undertaken by means of the Knudsen cell mass spectrometry toevaluate the thermodynamic activities of the volatile elements, Cs and I, in the pseudo-binary systems such as LiF-CsF and LiFLiI. The obtained activity data will be applied to the analysis of the behaviors of volatile FPs as described above.

Part of this work was supported by Chubu Electric Power Company through “Nuclear Safety Research Laboratory-Research
Grant”.
Keywords
Molten salt reactor, Fission product, Vaporization, Molten fluoride salt,

55

2a-1
Environmental parameters for assessing the behavior of radiocarbon in the
paddy soil-to-rice plant system
Nobuyoshi ISHII*, Shinichi OGIYAMA, Shinji SAKURAI, Keiko TAGAMI, Shigeo UCHIDA
Research Center for Radiation Protection, National Institute of Radiological Sciences

Transuranic (TRU) waste contains a variety of radionuclides, e.g., Np, Pu and long-lived radionuclides like 14C and 129I. In
Japan, TRU waste is categorized into four groups in accordance with physical properties and the concentration of radioactive
materials. Group 2 waste of these groups includes hull and end piece wastes with relatively high amounts of 14C, and leachingof low molecular weight 14C-organic materials from simulated hull wastes has been reported. The 14C-organic materials havevery little sorption properties to clay and rock, and thus the possible migration of the 14C from a TRU repository site to the biospherethrough groundwater presents some concern. Indeed, potential wastes from the Fukushima Daiichi nuclear power plants
expected to have 14C. Therefore, it is necessary to understand the behavior of 14C in the human habitation sphere for the safe
disposal of radioactive wastes.
In our research project, soil-soil solution distribution coefficients (Kd) and soil-to-rice plant transfer factors (TFs) of 14C havebeen determined to clarify the behavior of 14C in Japanese rice paddy fields. In addition, 14C partitioning in solid, liquid, and gas
phases was determined. As a source of 14C, 14C-sodium acetate was used because the chemical form would be occurred thehighest amounts among the potential 14C-organic materials. These environmental parameters commonly used for mathematicaldose assessment models.
The Kd values were obtained by tracer experiments and then calculated from activities of the 14C directly measured in solid andliquid phases at the end of 7 days incubation period. From the results, the mean of Kd value was 140 ± 77 L Kg-1. This value
was higher than expected from its chemical form. The degradation of the 14C-sodium acetate and the accumulation of the 14C bysoil microbes caused the relatively high value of Kd, and also the 14C gas was emitted from the culture (approximately 60% of total
added) due to the microbial activity.
TFs were determined by laboratory and field experiments. In the field experiments, TF of carbon was estimated by using
stable carbon isotope ratios, and the average TF value was 0.11 ± 0.04 for the edible part. On the other hand, the average TF of
14C, which was obtained by tracer experiments, was 6.8 ± 2.4; however, it seemed that the value was overestimated. In the tracer
experiments, rice plants also cultivated in the soil without 14C contamination as a negative control, and interestingly a portion of
the 14C added to other pots was detected in the negative control white rice. The results suggest that the 14C gas, which was emitted
from the contaminated soil as observed in the Kd experiments, was assimilated by the negative control plants throughphotosynthesis.
From these results, it was assumed that acetic acid as a 14C source in irrigated paddy soils was rapidly incorporated anddecomposed by soil microbes. Although the most of the 14C was released into the air as gas forms, some of the 14C was
assimilated into the microbial cells. The main chemical species of 14C gas released into the air was 14CO2 according to our
additional study, and a part of the 14CO2 could be used by photosynthesis of the rice plants. Only a small amount of the 14C wasabsorbed through the roots: however, the soil origin 14C to the rice plants was detected. Human expose to 14C through rice intake
must be considered in the safety assessment of TRU waste disposal.
This work has been partially supported by the Agency for Natural Resources and Energy, the Ministry of Economy, Trade andIndustry (METI), Japan.
Keywords
TRU waste, C-14, rice paddy, Kd, TF

56

2a-2
Development of ICP-MS based analytical method for the determination ofradioactive Cs isotopes
Jian ZHENG*, Tagami KEIKO, Wenting BU, and Shigeo UCHIDA
Research Center of Radiological Protection, National Institute of Radiological Sciences

Radioactive Cs isotopes are high yield fission products up to 6.535% and 6.236% for 135Cs and 137Cs, respectively, from thermal
neutron fission of 235U. They can be released into the environment in nuclear events, such as the detonation of a nuclear weapon
or nuclear reactor accident. In the fission chains for 135Cs and 137Cs, a shielding of 135Cs occurs due to neutron capture of its
precursor, 135Xe, to form 136Xe, whereas production of 137Cs is unaffected. This process causes a high degree of variance of
135Cs/137Cs isotope ratio with source. Thus, the ratio of 135Cs/137Cs will be characteristic of the reactor operation and shutdown
conditions. The determination of 135Cs by radiometric methods is very difficult and quite insensitive because of its long half-lifeand low-energy β-decay. Neutron activation analysis (NAA) and thermal ionization mass spectrometer (TIMS) have been
investigated for 135Cs activity and 135Cs/137Cs isotope ratio analysis in environmental samples, but these techniques are expensive
and not widely available. Recently, the analytical potential of ICP-MS for 135Cs, 137Cs activities and 135Cs/137Cs isotope ratio havebeen discussed in the studies of environmental radioactivity and nuclear forensics, but the analysis of environmental samples
remains to be a great analytical challenge. In Japan, there has been no report on the analysis of 135Cs/137Cs isotope ratio inenvironmental samples since no method has been established for 135Cs analysis in the MEXT environmental radioactivity
analytical method.
Here we report a newly established ICP-MS based analytical method for the determination of 135Cs, 137Cs activities and
135Cs/137Cs isotope ratio in environment samples with an aim to obtain isotopic composition of radioactive Cs isotopes released
from the Fukushima Daiichi nuclear power plant (FDNPP) accident. We propose that 135Cs/137Cs can be used as a powerful tracer
for long-term estimation of the environmental mobility of the FDNPP accident related fission products since the currently used
134Cs/137Cs activity ratio will become unavailable in the future for identification of the FDNPP contamination due to the short halflife of 134Cs. The developed analytical method can be applied to the analysis of radioactive wastes for the accurate radiation dose
assessment by 135Cs, and to nuclear fuel analysis to estimate nuclear fuel burn-up in the damaged reactors for reactor
decommissioning in the future.
Acknowledgment: This study was supported by Grant for Environmental Research Projects from the Sumitomo Foundation and
the Agency for Natural Resources and Energy, METI, Japan.
Keywords
Radioactive wastes, Radioactive Cs isotopes, ICP-MS

57

2a-3
Development of rapid analytical method of 129I in the contaminated water
and woods arising at Fukushima Daiichi Nuclear Power Station
Asako SHIMADA*¹,², Mayumi Ozawa², Yutaka KAMEO², Takuyo YASUMATSU³, Koji NEBASHI³, Takuya NIIYAMA²,
Shuhei SEKI², Masatoshi KAJIO², Kuniaki TAKAHASHI¹¹Nuclear Cycle Backend Directorate, Japan Atomic Energy Agency²Fukushima Project Team, Japan Atomic Energy Agency³Tokyo Power Engineering Co. Inc.
Introduction
At the moment, contaminated water has been arising in the reactor and turbine buildings at Fukushima Daiichi Nuclear
Power Station. The accumulated water processing facility was setup to decontaminate and to desalinate. The number of thetank storing the treated water has been increasing. Woods in the site were cut to get the space to set the tanks. These woods also
became radioactive waste. Iodine-129 is one of the important nuclides of which radioactivity has to be evaluated for the safety
disposal of the waste. Although I‒ is considered major species of 129I generated in the reactor, IO3‒ and I2 are possibly generated
depending on the condition. In addition to this, isotope exchange reaction between natural 127IO3‒ and reactor generated 129I‒
could be taken place. Therefore, analytical condition to determine each species of I in a solution was investigated. Furthermore,conventional analytical method of 129I in concrete was optimized to apply it to woods.
Experimental
129I‒ (0.1 Bq) and 127IO ‒ (1 μg) were added into 3 M NaOH solution or HCl solution (pH=2) with and without reductant
3(NaHSO3) to study the isotope exchange reaction and behavior of these iodine species in the analysis using Anion SR. The
operation of Anion SR was based on reference [1].
Known amount of I‒ or IO3‒ was added into the bark of pine, representative of woody material taken at the establishment ofJapan Atomic Energy Agency, and it was put in a wet oxygen gas line set in an electric furnace. The rate of temperature
increase was optimized for woody sample. The vaporized I was trapped in an alkaline solution. Because insoluble organic
material was deposited in the gas line and the trap, oxidant (hopcalite II) was set in a downstream of the sample to decompose it.
Iodine was measured by Inductively Coupled Plasma Mass Spectrometry measurement with Dynamic Reaction Cell.
Results and discussion
Table 1 recovery of I (%)
Table 1 shows recovery of I from examined
Without NaHSO3
With NaHSO3
diluent
solutions. From 3 M NaOH solution, 129I was
127I
129I
127I
129I
127
recovered and I wasn’t recovered regardless the
3 M NaOH
2.3±1.971±2.92.1±0.2
86±4.3
existence of the reductant. That is, 127I and 129I
HCl (pH2)
8.6±6.3
19±4.864±7.360±3.1
showed different behavior. This result supports thatI‒ is extracted and IO3‒ isn’t extracted by the Anion SR, Reductant doesn’t work, and reaction between 127IO3‒ and 129I‒ isnegligible. On the other hands, 129I and 127I were recovered with the reductant and were not recovered without the reductant
from the HCl solution. Similar behavior of 127I and 129I was observed in this case. This implies that IO3‒ is dominant species in
HCl solution at pH 2, IO3‒ is reduced to I‒ by the reductant in this experimental condition. Since the oxidation process from I‒ to
IO3‒ is including the reaction with I‒, I2 and IO3‒, isotope exchange would be taken place. Consequently, both of I‒ and IO3‒ areanalyzed at pH 2 condition and I‒ is analyzed at 3 M NaOH condition. Accordingly, speciation of I‒ and IO3‒ is available.
For woody material, approximately 90% recovery was obtained for both of I‒ and IO3‒ in the alkaline trap of combustion
method. Anomalous combustion was avoided by the stepwise and slow rate of temperature increase.
[1] Y. Kameo, et al., JAEA-Data/Code, (2010) 2010-028.
Keywords 129I, Accumulated Water, Anion SR

58

2a-4
Determination of low-level technetium-99 in the environmental samples by
inductively coupled plasma mass spectrometry
Shigeo UCHIDA*, Keiko TAGAMI
Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences,

Technetium-99 (half-life: 2.11x105 y) is a pure beta emitter and is produced in the fission of 235U and 239Pu with a relatively
high fission yield of ca. 6% like 137Cs. The amount of 99Tc has been increasing due to the operation of nuclear power plants. For
waste management of spent nuclear fuels and high level radioactive wastes after reprocessing, it is necessary to consider the fateof this radionuclide in the environment. Indeed, internationally, the long-lived isotopes 129I, 79Se and 99Tc have been identified as
key radionuclides for waste repositories. The dominant Tc species in natural aqueous solution in equilibrium with theatmosphere is the highly soluble chemical form pertechnetate, TcO4-. In the terrestrial environment, 99TcO4- has a highgeochemical mobility and availability for plants. We have also carried out 99TcO4- uptake study by plant roots grew inhydroculture, and found fairly high Tc bioavailability (unpublished data). However, it was not clear whether 99Tc was also
mobile in the actual environment or not.
To understand the fate in the environment, global-fallout 99Tc can be a good tracer. However, the concentrations were in low
levels in environmental samples, therefore, it was necessary to develop more sensitive methods. In this study, we focused on theuse of inductively coupled plasma mass spectrometry (ICP-MS) to measure 99Tc. The important points to measure 99Tc by
ICP-MS were (1) to remove interferences (elements having isobars and polyatomic ions) at mass of 99, (2) to concentrate 99Tc
in the separation processes, and (3) to use suitable yield tracers. For (1), removal of Ru and potential 98MoH+ ion interferences
needed to be clarified. For (2), liquid-liquid extractions with cyclohexanone and chromatographic extraction with TEVA resin(Eichrom) were compared. The suitable yield tracers would be 99mTc, 95mTc and Re (chemically similar to Tc) were considered.
Finally, we have developed several methods, i.e. soil, plants, river water, seawater, and seaweeds. In this presentation, we willsummarize the information for 99Tc analysis by ICP-MS. The method is applicable to measure ground water from monitoring
wells near radioactive waste disposal sites.

Keywords
Technetium-99, Analytical methods, Environmental samples, ICP-MS

59

2a-5
A simple and rapid method for separation and preconcentration of Ra inwater samples
Guosheng YANG, Jian ZHENG, Keiko TAGAMI, Shigeo UCHIDA
Office of Biospheric Assessment for waste disposal, Research Center of Radiological Protection,

Radium-226 is one of the important radionuclides for dose assessment of radioactive waste disposal, particularly, foruranium waste. For this uranium waste, shallow land disposal will be considered and it is necessary to carry out environmental
monitoring around the site, for example, well water (and ground water) monitoring. Since Ra-226 is a progeny of U-238,
continuous monitoring of its concentration in water samples is of great importance. It is difficult, however, to detect low
concentration of Ra-226 in water samples. Therefore, preconcentration of Ra in water samples is necessary. Recently,
application of sector field inductively coupled plasma mass spectrometry (SF-ICP-MS) for determination of long-lived
radionuclides present at ultraltrace level in the environment is considered as one of the popular and reliable methods; thismethod has several advantages, i.e., smaller sample size, higher precision, and time saving compared to those of theconventional radiometric methods. However, in order to measure ultratrace-levels of Ra-226 by SF-ICP-MS, some separation
steps for Ra from alkaline and alkaline earth element in the sample matrix is needed. Furthermore, the removal of any potential
sources of interference (molecular ions and isobaric interferences) that may result in incorrect results is of crucial importance,
especially the polyatomic interference 88Sr138Ba+ because of the relatively high concentrations of Sr and Ba in water samples.
Due to their similar chemical property, the separation of Ba-Ra is difficult; the effective removal of Sr from Ba (Ra) is morepractical for 226Ra determination by SF-ICP-MS. In this study, several ion-exchange chromatographic separation methods werecompared and the most suitable condition for Sr separation from Ba (Ra) in simulated river water will be reported.
Keywords
Ra-226, SF-ICP-MS, chromatographic separation, preconcentration, water sample

60

2a-6
Consideration on treatment and disposal of secondary wastes generated
from treatment of contaminated water
Hiromi TANABE1, Kuniyoshi HOSHINORadioactive Waste Management Funding and Research Center

Background
There is a significant volume of highly contaminated water generated from the accident of the Fukushima Daiichi Nuclear
Power Station. Several methods have been applied to decontaminate radioactivity of the water, such as the Areva’s device Curion’s
device, the SARRY, the ALPS, and so on. After the decontamination treatments using these devices, various secondary wastes
such as sludge and spent adsorbents have been generated. These wastes are now tentatively stored at the site but the further
treatment shall be applied to produce appropriate waste forms to be stored and disposed of in a repository finally.
Author has been studying about how to manage Hull wastes, i.e. Zircaloy, Stainless-steel and high Ni alloy, generated from a
reprocessing plant. For this purpose, author has been pursuing waste characterization work, has developed various treatment
technology options and has developed a disposal concept (i.e. the first TRU report). According to these experiences, authors
consider how to promote the treatment and disposal technology in safety and efficient way.
Needs of inventory list and its online management system
Establishing waste inventory including chemical and physical form, radionuclide inventory, source term, location of waste is
the fundamental issue and the first step for pursuing further consideration of treatment and disposal. However it is very difficult tofully establish it from the beginning. So the stepwise development and evaluation are important. A management system of thewaste inventory should be established, and newly obtained information should be shared through the management system. There
have been several and intensive waste characterization works and these results should be added in the waste inventory as soon as
possible.
Development strategy of waste treatment, storage, transport and disposal technique
Establishment of the criteria for development of each technology is the fundamental issue and the first step. For example, fortreatment technology, simple system, applicability for various waste, volume reduction factor, economy, minimized secondary
waste, and difficulty of residual R&D are important. The total cost of treatment and disposal should be also evaluated.
Regarding the stability of waste form, low leaching rate is needed especially for the waste containing long-lived radionuclides.
If identification of radionuclides of waste through the treatment process becomes easier, it will be also considered to be advantage.
Technology options for treatment, storage, transportation and disposal have been proposed from domestic and internationalorganizations. These options should be integrated and managed in a data base of technology options. R&D result should be also
included in the data base as soon as possible.
Organization of evaluation teamTo evaluate the feasibility of each technology and waste inventory information needed for each technology, it is important toorganize a team of system engineers, researchers, implementer and other related stakeholder. The team should evaluate thefeasibility of each technology together as stepwise manner.
Needs of long-term knowledge management
Because the actual treatment and disposal of wastes will be pursued after a few tens of years, it should be also considered aboutknowledge management, training and education of next generation.

61

2a-7
A simulation of Sr leaching in aqueous solution mixed with seawater andpreparation of U/Zr oxides at high temperatures
Takayuki SASAKI*1, Akira KIRISHIMA2, Yuu TAKENO1, Kohei FUKUDA2, Nobuaki SATO2¹Department of Nuclear Engineering, Kyoto University, Kyoto daigaku-Katsura C3, Nishikyo-ku, Kyoto 615-8540, Japan²Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, Katahira, Aoba-ku, Sendai 980-8577, Japan

After the station black out of Fukushima Dai-ichi NPP in March, 2011, thousands of tons of seawater (SW) for cooling thehigh temperature reactor cores was injected. The highly water-soluble radionuclides such as cesium and iodine were released into
the cooling water and the air from the melted core and debris which might be reacted with a certain amount of zirconium alloy ofthe cladding of the fuel and/or its oxide. Additionally, the other FPs and TRU were possibly transferred to the cooling water mixed
with SW under high temperature condition in the pressure vessel. As reported by TEPCO, most of elements have been and will berecovered by several water treatment systems. However, the real SW, which is different from the ideal and clear artificial solution
in lab, contains many kinds of matrix ion, suspended matters and impurities. We have no enough knowledge of the kind andamount of radionuclides (except cesium and iodine) presented at the bottom of RPV/PCV etc. and stacked in water treatment
facilities such as tanks, adsorption towers and waste containers, so far. The simulation of the radionuclide behavior in the real SW
will be helpful to explain the forthcoming analysis data about the contents of minor FPs and TRU in contaminated water and solid
waste, and to suggest more effective and safer decommissioning and waste management. In the present study, 1) the fuel debris
composed of oxides and metals of uranium and zirconium as main constituents in the reactor was prepared and the phase relations
analyzed, and in parallel, 2) the apparent chemical behavior of strontium in SW was preliminary investigated prior to the other FPs.
After the heat treatment of the mixture of UO2 and ZrO2 in vacuum, the UO2 solid solution phase such as ZryU1-yO2+x seemed
to be formed as well as the small amount of the monoclinic ZrO2 phase at temperatures lower than 1273 K. However, phases withhigher oxidation state such as, U3O7 and U3O8 seemed to appear at 1073 - 1473 K with significant amount of the ZrO2 phase. Over
this temperature, a new phase similar to the UO2 phase seemed to be formed showing the decomposition of U3O8 phase at hightemperature. Also, the tetragonal ZrO2 was also observed in the same sample since phase transition from monoclinic to tetragonal
occurred at high temperature. When the mixture of zirconium to uranium oxides was treated by mechanochemical method under
reducing condition, the X-ray pattern of the products showed that they were the same fluorite structure and the lattice parameter ofthis phase was linearly decreased with increasing zirconium ratio. This suggests that the UO2 and ZrO2 form the solid solution
from low Zr/U ratio to high one. These results were also discussed with the phase diagram.
Sr-90 in contaminated water has been detected at a quite low molar concentration level but high radioactivity, while natural
non-radioactive strontium ion exists in SW at around 10-4 mol/dm3 (M). In the thermodynamic point of view, the “additional”
tracer strontium must be dissolved in the cooling water completely. However, in the recovery test of Sr-85 (ca. 10-10 M) from SW(sampled at the seashore in Minamisoma) at the neutral pH and 25/90 °C, 5 % of Sr-85 was filtered out by 3 kDa molecular weight
cut-off filter, while 100% Sr-85 was recovered through 10 kDa to 0.45μm filters. In the SEM-EDX analysis, the filtrate on the 3
kDa filter was tiny, but a sort of silica colloid, aluminosilicate, and organic matter were assigned. The interaction with thesematerials may imply a different behavior of Sr from the ideal solution chemistry.
This work was supported by Grant-in-Aid for Scientific Research (S) 24226021.
Keywords
Fuel debris, Uranium/Zirconium oxides, Fission product, Strontium, Seawater

62

2a-8
Volume reduction of municipal solid wastes contaminated with radioactive
cesium by ferrocyanide coprecipitation technique
Yoko FUJIKAWA*¹, Hiroaki OZAKI 2, Hiroshi TSUNO 2, Pengfei WEI 2, Aiichiro FUJINAGA 2, Ryouhei TAKANAMI ²,
Shogo TANIGUCHI ²,Shojiro KIMURA³¹Kyoto University²Osaka Sangyo University³Osaka University of Pharmaceutical Sciences
The estimated remaining sustainable years of the final disposal sites for municipal solid waste (MSW) in Japan were only 18
years as of the end of FY2008. The waste avoidance, waste volume reduction and recycling of the MSW, therefore, had been thenational policy. However, the pollution of the environment from the TEPCO (Tokyo Electric Power Company) Fukushima Daiichi
nuclear reactor accident created an entirely new problem. Since waste incineration and water treatment are by their nature theprocesses that concentrate pollutants like radioactive cesium (rad-cesium hereafter) in the ashes and treatment sludge, MSW
containing high concentration of rad-cesium started to be produced in some areas. The recycling of MSW as concrete material andcompost, etc. has become difficult, and when the rad-cesium concentration in the waste was below the clearance level, rumors
often prevented its reuse. Also most citizens are in hard opposition to the disposal of rad-cesium-containing wastes even if theradioactivity of the waste is below the governmental limit to the disposal in the landfills with leachate collection systems (i.e., 8,000
Bq/kg of Cs-134+Cs-137). Inevitably, undisposed treatment residues are now piling up in many treatment facilities in Fukushima
Prefecture, which may eventually jeopardize the treatment itself and exert a serious effect to everyday life.
We have studied the applicability of the technique that uses insoluble ferrocyanide (Fer hereafter) compounds to selectively
remove rad-cesium from the extracts of MSW, which are the complex mixture of various organic and inorganic compounds. For
the present study, we collected waste materials (both sludge and fly ash from sewage plants, drinking water treatment plants, refuse
incinerators, and melting furnaces), some of them contaminated with rad-cesium, and some not contaminated. The samples wereextracted with water or oxalic acid. As for the samples originally not contaminated, we first added stable cesium (100 μg/L) or
Cs-137 (8,000 Bq/L), added metal salt if necessary, adjusted pH to the desired value, and then added soluble ferrocyanide salt to
produce insoluble Fer-metal compounds that co-precipitate cesium. Samples with rad-cesium contamination were extracted andtested likewise but without addition of Cs tracer. The extract before and after the coprecipitation were analyzed for rad-cesium andother constituents using a Ge semiconductor detector, ICP-MS and ion chromatography. The precipitate was subjected to X-ray
fluorescence analysis and FT-IR spectroscopy.
In most of the samples, indigenous metals (mixtures of Fe(III), Zn(II), Cu(II), and Mn(II)) in the extract formed precipitates with
Fer and effectively removed Cs. However, Cs removal was reduced when high concentration of Zn (e.g., 10 mM) was containedin the extract because Fer-Zn precipitate has low Cs distribution factor compared with other Fer-metal precipitates. Concerning theextracts of waste materials that had been treated to reduce the leaching of heavy metals, it was necessary to add Fe(II), Fe(III),
Cu(II) or Ni(II) salt to produce the Fer-metal precipitate in them. When Fer -Ni was precipitated in the extract of rad-cesiumcontaminated fly ash from a refuse incinerator, the removal of rad-cesium was low but it was improved when excess Fe(II) wasused. On the other hand, rad-cesium removal was high for Fer-Ni and Fer-Fe(III) co-precipitation when the extract of rad-cesiumcontaminated fly ash from a melting furnace was tested. The difference may be explained by postulating the existence of colloidal,
non-ionic rad-cesium (e.g., sorbed on suspended particles) in the incinerator fly ash extract. In case of oxalic acid extracts, oxalic
acid had to be diluted to 0.05M or lower to produce the Fer precipitate with 0.1 mM of Fer. On the whole, if the type of metal salts
and pH in the waste extract was optimized, Cs removal was higher than 95 percents when Fer concentration was 0.1 mM (c.a. 30
mg/L of precipitate in the extract).
Keywords ferrocyanide co-precipitation, cesium removal, municipal solid waste, decontamination

63

2a-9
A Prescription toward to the Long-term Isolation of High Level Radioactive
Waste in Japan.
Takao TSUBOYAFormer Managing Director, Radioactive Waste Management Funding and Research Center

After the Fukushima-Daiichi Nuclear Power Plants accident, existing spent fuels at reactor sites cause a new arguments
regarding with the high level radioactive waste issue together with vitrified wastes. It seems the high level radioactive waste issues
is one of the main reason of public concern on nuclear utilization.
The technical basis for the long-term isolation of high level radioactive waste (HLW) have been established for geologicaldisposal in Japanese islands by year 2000 based on 25 years intensive research as the H12 Project. On the other hand, a report wascompiled focus to the social and institutional aspect on HLW disposal by 1998. Then, the Final Disposal Law including provisions
e.g., role of the implement body as non-governmental organization, the funding system based on “polluter pays” principle, andconsent base site selection processes was enacted in 2000. It is to say Japan enters into the social implementation phase ofgeological disposal of HLW since 2000.
Although the implement body makes every endeavor to select survey sites, the beginning step for implementation phase,
however, the candidate sites has not yet been realized up to now from difficulty of gaining public confidence.
This paper will try to discuss on those issues from both technical view points and institutional ones of the final disposal of
HLW in Japan, then will try to show key factors may exist in governance of the implement body as well as machinery of site
selection system.
This paper proposes to reform the framework of the Final Disposal Law, e.g., the establish of implement body as thegovernmental organization, the introduction of a new decision-making process grounded on public participation.
Keywords
High Level Radioactive Waste, HLW, Long-term Isolation, Geological Disposal, Final Disposal Law, Governance, Implement
Organization, Decision-making Process, Public Participation64

2a-10

MEXT-Sponsored Research Activities for Environmental Load Reduction
Shinji YOSHIKAWA*, Kouichi USHIRODA, Michiaki OHKUBO, Yutaka HISHIKI, Kiyoshi HOMMA
Department for Nuclear Research and Development, Japan Science and Technology Agency

Ministry of Education, Culture, Sports, Science and Technology in JAPAN(MEXT) sponsors research
activities relevant to innovative nuclear systems, which aim at overcoming problems of existing commercial
nuclear power plants in safety, economy, proliferation-resistance and other issues. This sponsorship is
implemented as a framework of competitive research fund, entitled “Innovative Nuclear Research andDevelopment Program.” Within this framework, a new public recruitment for research proposals began in
FY2013 with specific purpose of environmental load reduction of nuclear spent fuel.
The framework of the competitive research fund and the concept of this research recruitment are explained,
as well as the contents of the 6 adopted projects.

Keywords
spent nuclear fuel, environmental load reduction65

2a-11
NUMO’s activities after the Great East Japan earthquake
Masahiko Saito, Satoru Suzuki, Isamu Seto*
Nuclear Waste Management Organization of Japan (NUMO)

NUMO (Nuclear Waste Management Organization of Japan) is the organization responsible for implementing geologicaldisposal of HLW and TRU waste. A three-stage site selection process was defined by the Specified Radioactive Waste Final
Disposal Act in June 2000 and open solicitation of volunteer municipalities commenced in December 2002. However, to date,no local municipality has come forward as a candidate for an initial literature survey. Against this backdrop, “Radioactive Waste
Working Group” organized by Japanese Government is currently addressing issues related to geological disposal.
NUMO takes various consensus building measures including hosting symposia, workshops as well as supporting classroom
debates at universities. In addition, it conducts nationwide researches on the public attitude on the geological disposal twice a year.
The latest research in February 2013 resulted that 72% of the people understood the need for geological disposal but only 17%
favored to construct a geological disposal facility in their home town. Their major concern was that safe disposal is not assured
considering the large earthquakes such as the March 2011 Great East Japan earthquake and the subsequent accidents at the
Fukushima Daiichi nuclear power plant. In order to improve this situation, NUMO has started studies on the potential impacts ofsuch catastrophic events on the operational and post-closure safety of a geological disposal system.
With respect to the operational safety, it has to be ensured that, despite various incidents which could occur during the repository
construction and operation phases, the overpack has sufficient physical strength and heat resistance to prevent the release of any
radionuclides. The study concluded that １even though the accidental conditions beyond the design-basis were assumed, thephysical and thermal durability of metal overpack would be warranted.
The identification and characterization of faults is an important issue to be addressed for ensuring post-closure safety. In fact,
previously undiscovered faults have caused M7-class earthquakes in Japan during the last decade. It is, however, expected that theapplication of the combined techniques of surface exploration and seismic prospecting would identify such faults during
preliminary and detailed site investigations. In addition, the possible impact of groundwater level changes associated with a major
earthquake on radionuclide migration was also assessed. Although a radionuclide release spike appears in response to such anearthquake, the calculated maximum dose is still significantly smaller than the safety limits of international dose standards.
The above studies have re-confirmed the safety in both the operational and post-closure phases. NUMO, however, willcontinue to increase the confidence of safe geological disposal by applying state-of the art technologies and lessons learned fromthe great earthquake, as well as to communicate with public to solve their concerns.
Keywords
geological disposal, HLW, earthquake66

2a-12Migration Parameters and their Evaluation and Estimation Methodologies
of Safety-relevant Radionuclides for Performance Assessment of Japanese
Geological Disposal of HLW and TRU Waste
Akira KITAMURA*¹, Yukio TACHI¹¹Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency

Many radionuclides have been selected as safety-relevant for performance assessment of Japanese geological disposal system
of HLW and TRU waste as shown in Fig. 1. Reliable migration parameters of these radionuclides, e.g. solubility in porewater inbuffer material, distribution coefficient onto buffer material and rocks, and effective diffusion coefficient in buffer material androcks, should be evaluated and estimated. Evaluation and estimation of solubility are based on thermodynamic calculation using
originally-developed thermodynamic database. Distribution coefficient and effective diffusion coefficient are evaluated andestimated based on accumulating many appropriate experimental values and using sorption and/or diffusion models, though there
are various methodologies. Sorption and diffusion databases have been developed by accumulating experimental values withvarious conditions as many as possible, assigning their reliability based on some criteria, and selecting appropriate values from theaccumulated values. Examples of the proposed methodologies on evaluation and estimation of migration parameters will bepresented. Tremendous efforts have been made to evaluate and estimate the reliable parameters as described above. Required
resources generally increase with increasing the number of elements of interest. Therefore, reduction of safety-relevantradionuclides using nuclear transmutation may be effective to reduce the efforts.
1

2

H
Li

3

Na

4

K

5

Rb

1

6

7

2

3

4

5

6

7

8

9

10

11

12

Be

13

B

14

15

16

17

C

N

O

F

He
Ne

18

P

S

Cl

Ar

C-14

Mg

Al

Si

Cl-36

Cs

Ca
Sr

Sc

Ti

V

Cr

Mn

Zr

Nb

Mo

Tc

Sr-90

Zr-93

Nb-93m,94

Mo-93

Tc-99

Ba

Hf

Ta

W

Re

Y

Fe
Ru

Co

Ni

Co-60

Ni-59,63

Rh

Pd

Cu Zn Ga Ge

Ir

Pt

Ag Cd

In

Sn Sb

Te

Sn-126

Au Hg

Tl

Kr

I

Xe

Pb

I-129

Po

At

Er Tm YbLu

Es Fm Md No

Lr

Cs-135,137

Fr

Br

Se-79

Pd-107

Os

As Se

Bi

Rn

Pb-210

Ra

Rf

Db

Sg

Bh

Hs

Mt

Ds

Rg Cn

Ce

Pr

Nd

Pm

Sm

Eu

Gd

Tb Dy Ho

Am

Cm

Bk

U-233,234, Np-237 Pu-238,239, Am-241,
235,236,238
240,241,242 242m,243

Cm-244,245,246

Ra-226,228

La

Sm-151

Ac

Th

Ac-227 Th-228,229,
230,232

Pa
Pa-231

U

Np

Pu

Cf

Fig. 1 Safety-relevant radionuclides for performance assessment of Japanese geological disposal of HLW and TRU waste
Keywords
High-level radioactive waste (HLW), TRU waste, geological disposal, performance assessment, safety-relevant radionuclide,solubility, distribution coefficient, effective diffusion coefficient672a-13
Evaluation of Distribution Coefficients for Radionuclides Sorption on
Bentonite Colloid
Takamitsu ISHIDERA*¹, Seiichi KUROSAWA², Masanori HAYASHI², Yasuyuki SUZUKI ²¹Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency²Inspection Development Co. Ltd.

In the geological disposal of high-level radioactive waste (HLW) in Japan, the vitrified HLW surrounded by overpack andcompacted bentonite is disposed of in deep underground to ensure the long-term safety. Compacted bentonite saturates withgroundwater after repository closure. Radionuclides contained in HLW may dissolve into porewater of compacted bentonite andmigrate through the compacted bentonite and host rock. Compacted bentonite effectively retards the radionuclides by sorption
with low permeability. However, colloidal particle of bentonite (bentonite colloid) might be generated by the erosion of thecompacted bentonite. Bentonite colloid sorbs less onto the minerals in host rock, so migration of radionuclides is possibly facilitatedby radionuclide sorption on bentonite colloid. Therefore, potential impact of bentonite colloid on the migration of radionuclides
needs to be evaluated in the performance assessment of geological disposal.
Japan Atomic Energy Agency (JAEA) has been investigating sorption behavior of radionuclides on the bentonite colloid anddeveloping the evaluation model for the radionuclide sorption behavior. In our previous studies, the distribution coefficients ofradionuclides on bentonite colloid were found to be higher than those on non-colloidal bentonite particle. This difference wasinterpreted to be due to the higher surface complexation site capacities in bentonite colloid. Considering this site capacities, thesorption behavior of Am was well interpreted based on a previously reported sorption model for non-colloidal bentonite1). The
calculated distribution coefficients by this evaluation method were also good agreement with the experimental results of Cs andAm sorption on bentonite colloid in 'solution - bentonite colloid - granite' ternary system, which is a simulated condition ofradionuclides migration in host rock 2).
To apply the evaluation model for the performance assessment, selectivity coefficients of safety-relevant radionuclides arenecessary for each surface complexation site. Therefore, the applicability of evaluation model should be validated systematicallyfor the safety-relevant radionuclides. In this context, we have been carrying out the sorption experiment of Cs, Am and Np on
bentonite colloid to verify the applicability. In the presentation, the evaluation model for distribution coefficient of radionuclides
sorption on bentonite colloid will be introduced in detail and the application result of the evaluation model to the sorption behaviorof Np on bentonite colloid will be shown.
1) K. Iijima et al., (2008). Radiochim. Acta 96, pp.721-730.
2) K. Iijima et al., (2010). Radiochim. Acta 98, pp.729-736.

Keywords
Geological Disposal, Performance Assessment, Distribution Coefficient, Bentonite Colloid, Sorption Model68

2a-14
The treatment process of simulated debris from a severe accident using
molten salt system
Yuya Takahashi*, Hitoshi Nakamura, Akira Yamada, Koji Mizuguchi, Reiko Fujita
Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862, Japan

In a severe nuclear accident, there is a risk of the nuclear fuel melting and forming debris. The debris contains uranium andhighly radioactive nuclides, and therefore needs to be handled with great care in terms of both critical safety and high radiation
safety. The debris also contains a large amount of weakly radioactive Fe and Zr, and it would be beneficial to separate the debris
into each material group. However, the composition of the debris differs from that of spent fuel, and it is not possible to employ
conventional reprocessing technology involving dissolution with nitric acid, because the debris contains a UO2-ZrO2 stable
insoluble oxide mixture. In addition, it is difficult to employ metallurgical processes because of the high melting points and because
the reactivity of Zr with water becomes higher above 900 °C. On the other hand, some molten salts are known to exhibit highsolubility, good Zr recovery, and the ability to decontaminate Fe components. We speculated that this approach would be usefulfor separating Fe and Zr from the debris to produce low-level radioactive waste. In this study, therefore, we propose the concept of
the treatment process and demonstrate the approach of the treatment process of debris. The simulated debris was prepared andpyrochemical electrolysis of Fe and Zr in NaCl-KCl molten salt was employed to separate Fe and Zr from debris.
The simulated debris was made fromZr, Fe, and CeO2. The CeO2was used for simulating stable UO2-ZrO2. The simulated debris was
prepared by using an arc furnace. The composition of the simulated debris was 20 wt% Zr, 20 wt% Fe, and 60 wt% CeO2. The
reagents of the simulated debris were placed in a crucible and melted with the arc furnace in an Ar atmosphere. This simulated themelting debris at the severe accident nuclear power plant. The cross-section of the simulated debris was analyzed by SEM and
EDX, and the oxide phase was analyzed by XRD. After the simulated debris was melted in the arc furnace and solidified, it was
separated into a metal phase and an oxide phase. The metal phase formed as spheres in the oxide phase. The two phases did notexhibit wettability with each other. From EDX analysis, the metal phase was a uniform alloy of Zr and Fe, and the oxide phase
was Ce oxide. From the XRD analysis, the organic phase was identified as CeO2.
Electrorefining of the simulated debris was conducted in molten NaCl-KCl containing 6 mol% ZrCl4. In this study, NaClKCl molten salt was selected from the viewpoint of the wide potential window and properties of this salt, which is not adeliquescent salt. NaCl-KCl-6mol%ZrCl4 (melting point of NaCl-KCl is 658 °C) was heated to 700 °C in an alumina cell to carry out
electrorefining. The simulated debris was placed on a Ni plate, serving as the anode. Stainless steel was used as the cathode. In theelectrorefining process, metal species in the simulated debris were dissolved electrically from the anode and deposited on thecathode. After electrorefining, 89% of the metal phase was separated from the simulated debris. The metal species in the simulated
debris was separated as a deposit on the cathode.
In this study, we successfully separated Zr and Fe from simulated debris using NaCl-KCl molten salt electrolysis, and weselectively recovered the Zr and Fe. With this approach, it should be possible to reduce the volume of the debris by recovering
metals, which can then be treated as low-radioactivity groups.
Keywords
severe accident, debris,pyrochemical, electrolysis, molten salt

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2b-1
Considering Geological Disposal Program of High-level Radioactive Waste
through Classroom Debate
Akemi YOSHIDA
Sugiyama Jogakuen University
1. Introduction
Debate, in a narrow sense, i.e. “classroom debate”, is conducted for the purpose of education and training. Perceived generally
as a means of enhancing communication skills, debate also provides other benefits. Debate is an excellent way to develop students’ability to collect, organize and understand information since they need to deal with a vast amount of information in order to grasp
issues associated with the resolution. In addition, students are allowed to take multiple,objective and critical perspectives throughthe experiences of taking both affirmative and negative positions, as well as refuting arguments in debate games.
Against this backdrop, “Introduction to Debate” courses offered at Sugiyama Jogakuen University and Aichi Shukutoku
University in the first semester of 2013, focused on “High-level Radioactive Waste Management Program”, a topic of current
importance. Yet students majoring in liberal arts, in particular, had poor knowledge of this topic. This paper will describe andanalyze the process and results of the coursework. It will also examine the advantages of debate in understanding and sharing thesaid topic.
2. Process and results of the coursework
At the outset, manners, techniques and rules commonly used in debate were introduced to students. Then, students repeatedly
practiced model debates, which led them to get accustomed to debating. After wards, explanation about the resolution was given.
“High-level Radioactive Waste Management Program” was set as the subject for debate this year. In order to help students
deepen their understanding of the issues and prepare for debate games, Professor Yamana of Kyoto University was invited to givelectures to students. With the cooperation of the Nuclear Waste Management Organization of Japan (NUMO), students also had
the opportunity to visit Chubu Electric Power Company’s Hamaoka Nuclear Power Station Exhibition Center as well as Japan
Atomic Energy Agency’s Underground Research Laboratory in Tono Geoscience Center.
Debate games were conducted under the resolution: “Should Japan abolish the geological disposal program and mandate
permanent waste management?” Carefully written and debated last year in Professor Fujikawa’s course at Chiba University, theresolution was appropriate for use in this year’s debate at Sugiyama and Aichi Shukutoku. In each class of the two universities,
two or three groups were formed to conduct the research necessary for debate games. As students would take both affirmative andnegative positions, they explored arguments on both sides of the issue.
In debate games, some groups took full advantage of their preparation efforts while others did not. In the end, all seemed torealize the importance of preparation.
3. Evaluating effects of the coursework.
Through lectures by experts and tour to relevant facilities, students successfully gained a deep understanding of the topic which,at the outset, was almost unknown to them.
Participating in debate games also motivated students to fully grasp detailed issues since superficial knowledge is notenough. As students took both affirmative and negative positions, debate was effective in dealing with the controversial issue of
nuclear waste management.
This year, students before joining the course, were almost indifferent to the resolution, which must be examined regardless ofJapan’s policy over nuclear power - , either it will continue or it will be abolished. They became interested in the topic at theend. As a result, the goal of the course was accomplished.
Keywords
Debate, Radioactive Waste

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