JAEA-Conf 2020-001
INDC(JPN)-206
DOI:10.11484/jaea-conf-2020-001

Proceedings of the 2019 Symposium on Nuclear Data
November 28-30, 2019,
Kyushu University,
Chikushi Campus, Fukuoka, Japan(Eds.) Yukinobu WATANABE, Nobuhiro SHIGYO, Tadahiro KIN
and Osamu IWAMOTO

Nuclear Science and Engineering Center
Sector of Nuclear Science Research

December 2020
Japan Atomic Energy Agency

日本原子力研究開発機構

本レポートは国立研究開発法人日本原子力研究開発機構が不定期に発行する成果報告書です。
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なお、本レポートの全文は日本原子力研究開発機構ホームページ（https://www.jaea.go.jp）
より発信されています。
国立研究開発法人日本原子力研究開発機構 研究連携成果展開部 研究成果管理課〒319-1195 茨城県那珂郡東海村大字白方 2 番地4
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JAEA-Conf 2020-001
INDC(JPN)-206

Proceedings of the 2019 Symposium on Nuclear Data
November 28-30, 2019,
Kyushu University,
Chikushi Campus, Fukuoka, Japan(Eds.) Yukinobu WATANABE*, Nobuhiro SHIGYO*, Tadahiro KIN*,
and Osamu IWAMOTO
Nuclear Science and Engineering Center, Sector of Nuclear Science Research,
Japan Atomic Energy Agency,
Tokai-mura, Naka-gun, Ibaraki-ken(Received August 24, 2020)
The 2019 Symposium on Nuclear Data was held at Chikushi Campus Cooperation Building (C-Cube),
Kyushu University, on November 28 to 30, 2019. The symposium was organized by the Nuclear Data
Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Sigma Investigative Advisory
Committee of AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency, Kyushu
Branch of AESJ, and Center for Accelerator and Beam Applied Science of Kyushu University. In the
symposium, there were one tutorial, “From the resonance theory to statistical model”, and five sessions,“Study on Nuclear Data and related topics”, “Reactor physics”, “International Cooperation”, “Nuclear
Physics”, and “High Energy Nuclear Data and their Application”. In addition, recent research progress on
experiments, nuclear theory, evaluation, benchmark and applications was presented in the poster session.
Among 85 participants, all presentations and following discussions were very active and fruitful. This reportconsists of total 42 papers including 13 oral and 29 poster presentations.
Keywords: Nuclear Data Symposium 2019, Experiments, Nuclear Theory, Nuclear Data Evaluation,
Benchmark Test, Nuclear Data Applications*Kyushu University
Organizers: Yukinobu Watanabe (Chair, Kyushu Univ.), Hideaki Otsu (Co-chair, RIKEN), Toshiya Sanami(KEK), Tatsuya Katabuchi (Tokyo Tech.), Atsushi Kimura (JAEA), Katsuhisa Nishio (JAEA), Go Chiba(Hokkaido Univ.), Satoshi Kunieda (JAEA), Takanori Kitada (Osaka Univ.), Shin-ichro Meigo (JAEA),
Futoshi Minato (JAEA) , Satoshi Chiba (Tokyo Tech.)iii

JAEA-Conf 2020-001
INDC(JPN)-206

2019 年度核データ研究会報告集

2019 年 11 月 28 日～30 日
九州大学
日本原子力研究開発機構(編) 渡辺筑紫キャンパス

原子力科学研究部門

福岡県

原子力基礎工学研究センター

幸信*、執行 信寛*、金 政浩*、岩本修(2020 年 8 月 24 日受理)2019 年度核データ研究会は、2019 年 11 月 28 日～30 日に、福岡県春日市にある九州大学筑紫キャンパスの総合研究棟(C-Cube)にて開催された。本研究会は、日本原子力学会核データ部会が主催、日本原子力学会「シグマ」調査専門委員会、日本原子力研究開発機構原子力基礎工学研究センター、日本原子力学会九州支部、九州大学加速器・ビーム応用科学センターが共催した。今回、チュートリアルとして「共鳴理論から統計模型へ」を、講演・議論のセッションとして、「核デー
タ研究及び関連トピックス」、「炉物理研究」、「国際協力」、「原子核物理」、「高エネルギー核デー
タと応用」の 5 セッションを企画し実施した。さらに、ポスターセッションでは、実験、理論、
評価、ベンチマーク、応用等、幅広い研究内容について発表が行われた。参加者総数は 85 名、それぞれの口頭発表及びポスター発表では活発な質疑応答が行われた。本報告集は、本研究会における口頭発表 13 件、ポスター発表 29 件の論文を掲載している。

キーワード：2019 年度核データ研究会、実験、原子核理論、核データ評価、ベンチマークテスト、
核データ応用原子力科学研究所：〒319-1195

茨城県那珂郡東海村大字白方 2-4*九州大学2019 年度核データ研究会実行委員会：
渡辺幸信(委員長、九大)、大津秀暁(副委員長、理研)、佐波俊哉(KEK) 、片渕竜也(東工大)、木村敦(原子力機構) 、西尾勝久(原子力機構) 、千葉豪(北大)、国枝賢(原子力機構)、
北田孝典(阪大)、明午伸一郎(原子力機構) 、湊太志(原子力機構)、千葉敏(東工大)iii

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Contents
1.

2019 Symposium on Nuclear Data - Programme ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 1
Papers Presented at Oral Sessions

2.

Neutron Capture Cross Section Measurement of Minor Actinides in Fast Neutron Energy Region for Studyon Nuclear Transmutation System ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 5
Tatsuya KATABUCHI (Tokyo tech.) et al.

3.

Recent Progress and Future Plan on the JENDL Project ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 11
Osamu IWAMOTO (JAEA)

4.

A Consideration on Nuclear Data Needs for 1F Decommissioningꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 17
Tokio FUKAHORI (JAEA)

5.

Fundamental Physics using Neutron at J-PARC and Accelerator-driven Neutron Source ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 21
Tamaki YOSHIOKA (Kyushu Univ.)

6.

Recent R&D of HTGR and Requirement for Nuclear Dataꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 27
Yuji FUKAYA (JAEA) et al.

7.

Performance of a Gas-cooled Reactor as a Tritium Production Device for Fusion Reactorsꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 33
Hideaki MATSUURA (Kyushu Univ.)

8.

Research and Development of an Innovative Transmutation System of LLFP by Fast Reactorsꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 39
Satoshi CHIBA (Tokyo tech.)

9.

Nuclear Data Activities and Related Database Services at the IAEA Nuclear Data Section ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 45
Shin OKUMURA (IAEA)

10. Status of the Nuclear Energy Agency Data Bank – Services for Nuclear Data and Computer Programs ꞏ 51
Kenya SUYAMA (OECD/NEA) et al.
11. From the Resonance Theory to the Statistical Model ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 55
Toshihiko KAWANO (LANL)
12. Current Status of Search for Element 119 ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 61
Satoshi SAKAGUCHI (Kyushu Univ.) et al.
13. Application of CDCC to Many-body Breakup Reaction ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 63
Takuma MATSUMOTO (Kyushu Univ.)
14. Topics from Radiation Safety Design of J-PARC ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 69
Hiroshi NAKASHIMA (JAEA)

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Poster Session
15. Production Cross Sections of 45Ti via Deuteron-induced Reaction on 45Sc ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 75
Tsoodol ZOLBADRAL (Hokkaido Univ., National Univ. of Mongolia) et al.
16. Production of 169Yb by the Proton-induced Reaction on 169Tm ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 79
Moemi SAITO (Hokkaido Univ.) et al.
17. Measurements of Nuclear Cross Section Data for Proton-induced Reactions on Bi and Pb Targets at
Intermediate Energy ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 83
Mahdi BAKHTIARI (POSTECH) et al.
18. Measurement of Nuclide Production Cross Sections for Z = 26-30 Elements Irradiated with 0.4 - 3.0 GeV
Proton in J-PARC ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 89
Hiroki MATSUDA (JAEA) et al.
19. Study of an Angular Correlation of γ-rays Emitted by 117Sn(n,γ) Reactions for T-violation Search ꞏꞏꞏꞏꞏꞏꞏ 95
Jun KOGA (Kyushu Univ.) et al.
20. Measurement of the Photon Strength Function in 115In at the γELBE Facility ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 101
Ayano MAKINAGA (Teikyo Univ., JIFS, Hokkaido Univ.) et al.
21. Comparison between Experimental and Calculation Neutron Spectra of the 197Au(γ,sn) Reaction for 17 MeVPolarized Photon ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 107
Tran Kim TUYET (SOKENDAI, KEK) et al.
22. Measurement of Neutron Total Cross-section of 93Nb at J-PARC MLF ANNRI ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 113
Shunsuke ENDO (JAEA) et al.
23. High-Energy Measurement of the Neutron Capture Cross Section of 237Np ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 119
Gerard Rovira LEVERONI (Tokyo tech.) et al.
24. New Methods to Reduce Systematic Uncertainties of Capture Cross Section Measurement Using a Sample
Rotation System ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 127
Yu KODAMA (Tokyo tech.) et al.
25. Proton Spectra with Low-Energy-Threshold from 40- and 70-MeV Proton-Induced Reactions ꞏꞏꞏꞏꞏꞏꞏꞏꞏ 131
Yuji YAMAGUCHI (Kyushu Univ.) et al.
26. Charged Particle Emission Reactions Induced by 100-MeV/u 12C ionsꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 137
Kazuhito YOSHIDA (Kyushu Univ.) et al.
27. Measurement of Neutron Energy Spectra of 345 MeV/u 238U Incidence on a Copper Target ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 143
Kenta SUGIHARA (Kyushu Univ., RIKEN) et al.
28. Measurement of the Energy Spectra of Hydrogen Isotopes from Nuclear Muon Capture in natSi ꞏꞏꞏꞏꞏꞏꞏꞏꞏ 147
Seiya MANABE (Kyushu Univ.) et al.
29. Detailed Examination of Benchmark Method for Large Angle Scattering Reaction Cross Section at 14MeV
for a flake target ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 153
Kazuki FUKUI (Osaka Univ.) et al.

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30. Integration Test with a Gaseous Detector and a Solenoidal Magnet for the Precise Neutron Lifetime
Measurementꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 159
Kodai YANO (Kyushu Univ.) et al.
31. HUNS Upgrade; Simulation of the Beam Profile for a Thermal Neutron Source in HUNS-2 ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 165
Kaoru Y. HARA (Hokkaido Univ.) et al.
32. Preliminary Experiment on Characterization of RANS-II Neutron Production via the 7Li(p, n)7Be Reaction
with 2.49 MeV Proton Injection ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 171
Shogo IZUMITANI (Kyushu Univ.) et al.
33. Development of a Neutron Beam Monitor for Nuclear Data Measurement using Spallation Neutron Sourceꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 177
Hideto NAKANO (Tokyo tech.) et al.
34. Model Analysis of Isotope-Production Cross Sections for Proton- and Deuteron-Induced Reactions on 93Zr
ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 181
Keita NAKANO (Kyushu Univ.) et al.
35. Systematical Calculation of Probabilities of Beta-delayed Neutron Emission and Fission in the Entire Regionof Nuclear Chart ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 187
Hiroyuki KOURA (JAEA) et al.
36. Development of Multi-group Neutron Activation Cross-section Library from JENDL/AD-2017 ꞏꞏꞏꞏꞏꞏꞏꞏ 193
Chikara KONNO (JAEA)
37. Quantification of Integral Data Effectiveness using the Concept of Active Sub-Space in Nuclear Data Testingꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 199
Go CHIBA (Hokkaido Univ.) et al.
38. Uncertainty Estimation of Conventional Neutron-spectrum Unfolding Codes with Monte-Carlo Based
Methodꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 205
Katsumi AOKI (Kyushu Univ.) et al.
39. The Influence of B and Li Burnable Poison on Effective Multiplication Factor in the HTGR ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 211
Yuki KOGA (Kyushu Univ.) et al.
40. Evaluation of Radioactive Concentration Produced in Electric Equipments and Materials on theDecommissioning of Nuclear Power Plants ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 217
Hirokuni YAMANISHI (Kindai Univ.) et al.
41. Evaluation of Scattered Radiation on the Contralateral Breast in Breast-conserving Therapy ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 221
Rikuto HARA (Teikyo Univ.) et al.
42. Burnup Calculation with Different Fission Yield Data ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 225
Atsuya ISO (Osaka Univ.) et al.
43. Improvement of Fuel Ion Ratio Diagnostics Performance using Anisotropic Neutron Emission Spectrum in
NBI Heating Deuterium-tritium Plasma ꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏꞏ 231
Tomoki URAKAWA (Kyushu Univ.) et al.

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目次
1.

2019 年度核データ研究会プログラム ············································································ 1
口頭発表論文2.

長寿命 MA 核種の高速中性子捕獲反応データに関する研究 ··············································· 5片渕3.

JENDL プロジェクトの最近の進展と計画 ··································································· 11
岩本

4.

秀明 (九州大学)

高速炉を活用した LLFP 核変換システムの研究開発 ······················································ 39
千葉

9.

裕司 (原子力機構) 他

核融合炉用トリチウム製造に対する高温ガス炉の特性···················································· 33
松浦

8.

瑞樹 (九州大学)

最近の高温ガス炉の研究開発と核データへの要求·························································· 27
深谷7.

智生 (原子力機構)

J-PARC での中性子基礎物理実験と小型中性子源 ·························································· 21
吉岡

6.

修 (原子力機構)

1F 廃炉のための核データニースに関する一考察 ··························································· 17
深堀

5.

竜也 (東工大) 他

敏 (東工大)IAEA 核データセクションにおける核データ活動及びデータベースサービス ······················ 45
奥村

森 (国際原子力機関)10. OECD NEA データバンクの現状 - 核データとコンピュータプログラムサービスを中心に ·· 51
須山賢也 (経済協力開発機構／原子力機関) 他

11. 共鳴理論から統計模型まで ························································································ 55
河野俊彦 (ロスアラモス国立研究所)

12. 新 119 番元素合成実験の現状 ···················································································· 61
坂口

聡志 (九州大学) 他

13. CDCC 理論と多体分解反応への応用 ··········································································· 63
松本琢磨 (九州大学)

14. J-PARC における放射線安全設計からの話題 ································································ 69
中島

宏 (原子力機構)

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JAEA-Conf 2020-001

ポスター発表論文
15. Production cross sections of 45Ti via deuteron-induced reaction on 45Sc··························· 75
Tsoodol ZOLBADRAL (北海道大学 / モンゴル国立大学) 他
16. ツリウム標的への陽子入射反応による 169Yb 生成 ·························································· 79
齋藤萌美 (北海道大学) 他

17. Measurements of nuclear cross section data for proton-induced reactions on Bi and Pb targets
at intermediate energy ···························································································· 83
Mahdi BAKHTIARI (浦項工科大学) 他
18. J-PARC における 0.4 GeV – 3.0 GeV 陽子を用いた Z=26-30 元素に対する核種生成断面積測定···························································································································· 89
松田洋樹 (原子力機構) 他

19. 時間反転対称性の破れ探索に向けた 117Sn(n,γ)反応による γ 線角相関に関する研究 ············· 95
古賀淳 (九州大学) 他

20. γELBE 施設における In-115 の γ 線強度関数測定 ······················································· 101
牧永

あや乃 (帝京大学 / NPO 法人あいんしゅたいん / 北海道大学) 他

21. Comparison between experimental and calculation neutron spectra of the 197Au(γ,sn) reaction
for 17 MeV polarized photon ·················································································· 107
Tran Kim TUYET (総研大 / 高エネ研) 他
22. J-PARC MLF ANNRI を用いた 93Nb の中性子全断面積の測定 ·······································113
遠藤

駿典 (原子力機構) 他

23. High-Energy Measurement of the Neutron Capture Cross Section of 237Np ····················119
Gerard Rovira LEVERONI (東工大) 他
24. 試料回転法を用いた中性子捕獲断面積の高精度化のための新たな手法 ···························· 127
児玉

有 (東工大) 他

25. 40 MeV 及び 70 MeV 陽子入射反応における低エネルギー下限陽子スペクトル ··············· 131
山口雄司 (九州大学) 他

26. 100MeV/u 12C ビーム入射荷電粒子生成反応の研究 ····················································· 137
吉田

和人 (九州大学) 他

27. 銅標的に対する 345 MeV/u ウラン入射により生成する中性子エネルギースペクトルの測定 ·· 143
杉原

健太 (九州大学 / 理研) 他

28. Si に対する負ミューオン原子核捕獲反応からの放出水素イオンエネルギースペクトル測定 ··· 147
真鍋

征也 (九州大学) 他

29. フレーク状ターゲットの 14MeV での後方散乱反応断面積のベンチマーク実験法の詳細検討 · 153
福井

和輝 (大阪大学) 他

30. 中性子の精密寿命測定のためのガス検出器とソレノイド磁石の総合試験 ························· 159
矢野浩大 (九州大学) 他

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31. HUNS アップグレード；HUNS-2 熱中性子源のためのビームプロファイルのシミュレーション·························································································································· 165
かおる (北海道大学) 他

原

32. 2.49 MeV 陽子入射による 7Li(p, n)7Be 反応を用いた RANS-II の中性子生成特性に関する予備実験·························································································································· 171
泉谷

祥伍 (九州大学) 他

33. 核破砕中性子源を利用した核データ測定のための中性子ビームモニターの開発 ················ 177
中野

秀仁 (東工大) 他

34. 93Zr に対する陽子・重陽子入射反応における同位体生成断面積の理論モデル解析 ············· 181
中野

敬太 (九州大学) 他

35. β 崩壊遅発中性子放出および β 崩壊遅延核分裂の核図表全体における系統的な計算 ··········· 187
小浦

寛之 (原子力機構) 他

36. JENDL/AD-2017 の多群中性子放射化断面積ライブラリ開発 ········································ 193
今野

力 (原子力機構)

37. 核データ検証における有効部分空間法による積分データの有効性の定量化 ······················ 199
千葉

豪 (北海道大学) 他

38. アンフォールディングコードにおけるモンテカルロ手法を用いた中性子スペクトルの不確かさ推定·························································································································· 205
青木勝海 (九州大学) 他

39. 高温ガス炉におけるホウ素可燃性毒物と T 製造用 Li ロッドの実効増倍率経時変化に及ぼす影響···························································································································211
古賀

友稀 (九州大学) 他

40. 原子力発電所廃止措置における電機品の放射性物質濃度の評価 ····································· 217
山西弘城 (九州大学) 他

41. 乳房温存療法における対側乳房に影響する散乱線量の評価············································ 221
原

陸人 (九州大学) 他

42. 異なる核分裂収率データを用いた燃焼計算 ································································· 225
磯敦也 (大阪大学) 他

43. NBI 加熱 DT プラズマにおける、中性子の非等方放出スペクトルを用いた燃料イオン比診断の測定性能改善 ··········································································································· 231
浦川

知己 (九州大学) 他

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1 2019 Symposium on Nuclear Data - Programme
Chikushi Hall, Chikushi Campus, Kyushu University, Nov 28–29, 2019.

Thursday, November 28
Opening Address (13:00–13:10)

Prof. Yukinobu Watanabe (Kyushu University)

Study on Nuclear Data (13:10–15:10)

Chair: Prof. Yukinobu Watanabe
13:10 Neutron Capture Cross Section Measurement of Minor Actinides inFast Neutron Energy Region for Study on Nuclear Transmutation
SystemProf. Tatsuya Katabuchi (Tokyo Institute of Technology)
13:40 Recent progress and future plan on the JENDL project
Dr. Osamu Iwamoto (Japan Atomic Energy Agency)
14:10 A Consideration on Nuclear Data Needs for 1F Decommissioning
Dr. Tokio Fukahori (Japan Atomic Energy Agency)
14:40 Fundamental physics using neutron at J-PARC and accelerator-driven
compact neutron source
Dr. Tamaki Yoshioka (Kyushu University)

Coffee Break (15:10–15:30)

Group photos will be taken at the entrance of the C-Cube building.

Poster Session - Gallery, C-Cube 3rd Floor (15:30–17:00)
Social Gathering - Restaurant Sonne (17:15–19:15)

Friday, November 29
Reactor Physics (9:30–11:00)

Chair: Prof. Go Chiba (Hokkaido University)
9:30 Recent R&D of HTGR and Requirement for Nuclear Data
Dr. Yuji Fukaya (Japan Atomic Energy Agency)
10:00 Performance of a gas-cooled reactor as a tritium production device forfusion reactors
Dr. Hideaki Matsuura (Kyushu University)
10:30 Research and development of an innovative transmutation system of
LLFP by fast reactors
Prof. Satoshi Chiba (Tokyo Institute of Technology)-1-

JAEA-Conf 2020-001

International Cooperation (11:10–12:10)

Chair: Dr. Tokio Fukahori (Japan Atomic Energy Agency)
11:10 Nuclear Data Activities and Related Database Services at the IAEA
Nuclear Data Section
Dr. Shin Okumura (International Atomic Energy Agency)
11:40 Status of the OECD NEA Data Bank – Services for Nuclear Data andComputer Program
Dr. Kenya Suyama (OECD NEA Data Bank)

Tutorial (13:20–14:30)

Chair: Dr. Osamu Iwamoto (Japan Atomic Energy Agency)
13:20 From the resonance theory to the statistical model
Dr. Toshihiko Kawano (Los Alamos National Laboratory)

Nuclear Physics (14:45–16:00)

Chair: Dr. Hideaki Otsu (RIKEN Nishina Center)
14:45 Current status of search for new element 119
Prof. Satoshi Sakaguchi (Kyushu University)
15:15 Application of CDCC to many-body breakup reaction
Dr. Takuma Matsumoto (Kyushu University)

High Energy Nuclear Data and Application (16:10–17:40)
Chair: Dr. Nobuhiro Shigyo (Kyushu University)
16:10 Topics from Radiation Safety Design of J-PARC
Dr. Hiroshi Nakashima (Japan Atomic Energy Agency)

16:40 Present Status and Outlook of IFMIF and A-FNS
Dr. Atsushi Kasugai (National Institutes for Quantum and Radiological
Science and Technology)
17:10 Construction status and future plan for RAON and its nuclear data
production system
Prof. Moses Chung (Ulsan National Institute of Science and Technology)

Poster Award and Closing Address (17:40–18:00)

Prof. Yukinobu Watanabe (Kyushu University), Dr. Hideaki Otsu (RIKEN Nishina
Center)Saturday, November 30
Accelerator Facility Tour (10:00–11:10)
Dr. Nobuhiro Shigyo (Kyushu University)-2-

JAEA-Conf 2020-001

2019 Symposium on Nuclear Data - Poster List
#

Name

Title

1

TSOODOL, Zoldabrai(Hokkaido U)

Production cross sections of 45Ti via deuteron-induced reaction on
45Sc

2

SAITO, Moemi(Hokkaido U)

Production of 169Yb by the proton-induced reaction on 169Tm

3

HWANG, Jongwon(CNS, U Tokyo)

Proton-induced reactions of 93Zr using the OEDO beamline

4

BAKHTIARI, Mahdi(POSTECH)

Measurements of nuclear cross section data for proton-induced
reactions on Bi and Pb targets at intermediate energy

5

MATSUDA, Hiroki(J-PARC/JAEA)

Measurement of nuclide production cross sections for Z = 26-30
elements irradiated with 0.4 - 3.0 GeV proton in J-PARC

6

KOGA, Jun
(Kyushu U)

Study of an angular correlation of γ-rays emitted by 117Sn(n,γ)
reactions for T-violation search

7

MAKISE, So(Kyushu U)

Measurement of γ ray angular distribution to come from 4.53 eV pwave resonance of 111Cd in compound nuclear reaction

8

MAKINAGA, Ayano(Teikyo U)

Measurement of photon strength function in In-115 at gELBE
facility

9

TRAN, Kim Tuyet(SOKENDAI)

Comparison between experimental and calculation neutron spectraof the 197Au (γ,n) reaction for 17 MeV polarized photon

10

ENDO, Shunsuke(JAEA)

Measurement of total neutron cross section of Niobium at J-PARC
MLF ANNRI

11

ROVIRA LEVERONI,
Gerard (Tokyo Tech)

High-Energy Measurement of the Neutron Capture Cross Section of
237Np

12

KODAMA, Yu(Tokyo Tech)

New Methods to Reduce Systematic Uncertainties of Capture Cross
Section Measurement Using a Sample Rotation System

13

YAMAGUCHI, Yuji
(Kyushu U)

Proton Spectra with Low-Energy-Threshold from 40- and 70-MeV
Proton-Induced Reactions

14

YOSHIDA, Kazuhito(Kyushu U)

Charged particle emission reactions induced by 100-MeV/u 12C
ions

15

MOON, Dalho(Sunkyunkwan U)

Neutron Production Double-differential cross section from Carbon
and Niobium targets bombarded with 290 MeV/u 136Xe ions

16

SUGIHARA, Kenta(Kyushu U/RIKEN)

Measurement of Neutron Energy Spectra of 345 MeV/u 238U
Incidence on a Cu Target

17

MANABE, Seiya(Kyushu U)

Measurement of the energy spectra of hydrogen isotopes from
nuclear muon capture in natSi

18

FUKUI, Kazuki(Osaka U)

Detailed examination of benchmark experimental method for large
angle scattering reaction cross section at 14MeV for a flake target

19

YANO, Kodai(Kyushu U)

Integration test with a Gaseous Detector and a Solenoidal Magnet
for the Precise Neutron Lifetime Measurement-3-

JAEA-Conf 2020-001

#

Name

20

HARA, Kaoru(Hokkaido U)

HUNS upgrade; Simulation of the beam profile for a thermal
neutron source in HUNS-2

21

IZUMITANI, Shogo(Kyushu U)

Preliminary experiment on characterization of RANS-II neutron
production via the 7Li(p, n)7Be reaction with 2.49 MeV proton
injection

22

NAKANO, Hideto(Tokyo Tech)

Development of a neutron beam monitor for nuclear datameasurement using spallation neutron source

23

NAKANO, Keita(Kyushu U)

Model analysis of isotope-production cross sections for proton- anddeuteron-induced reactions on Zr-93

24

FUJIO, Kazuki(Tokyo Tech)

Dependence of the potential energy surface of U-236 system on
effective nucleon-nucleon interactions

25

ZHANG, Xuan
(Tokyo Tech)

Fission study using four-dimensional Langevin model - Nuclear
shape of fission fragments

26

KOURA, Hiroyuki(JAEA)

Systematical calculation of probabilities of beta-delayed neutron
emission and fission in the entire region of nuclear chart

27

INAKURA, Tsunenori(Tokyo Tech)

Estimation of covariance of the neutron cross section of the longlived fission product

28

AMITANI, Tatsuki(Tokyo Tech)

Estimation of Covariance of Neutron Cross Sections for
Decommissioning

29

KONNO, Chikara(JAEA)Development of multi-group neutron activation cross-section
library from JENDL/AD-2017

30

CHIBA, Go(Hokkaido U)

Quantification of Integral Data Effectiveness Using the Concept ofActive Sub-Space in Nuclear Data Testing

31

AOKI, Katsumi(Kyushu U)

Uncertainty estimation of conventional neutron-spectrum unfolding
codes with Monte Carlo based method

32

KOGA, Yuki
(Kyushu U)

The Influence of B Burnable poison and T Production Li rod on
Effective Multiplication Factor in the HTGR

33

Evaluation of radioactive concentration produced in electric
YAMANISHI, Hirokuni
equipments and materials on the decommissioning of nuclear power(Kindai U)
plants34

HARA, Rikuto(Teikyo U)

35 ISO, Atsuya (Osaka U)
36

URAKAWA, Tomoki(Kyushu U)

Title

Evaluation of scattered radiation on the contralateral breast inbreast-conserving therapy
Burnup calculation with different fission yield data
Improvement of fuel ion ratio diagnostics performance using
anisotropic neutron emission spectrum in NBI heating deuteriumtritium plasma-4-

JAEA-Conf 2020-001

2 Neutron Capture Cross Section Measurement of Minor
Actinides in Fast Neutron Energy Region for Study on
Nuclear Transmutation System
Tatsuya KATABUCHI1)*, Jun-ich HORI2), Nobuyuki IWAMOTO3), Osamu IWAMOTO3),
Atsushi KIMURA3), Shoji NAKAMURA3), Yuji SHIBAHARA2), Kazushi TERADA1),
Kenichi TOSAKA1), Shunsuke ENDO3), Gerard ROVIRA1), Yu KODAMA1), Hideto
NAKANO1)
1) Tokyo Institute of Technology,
2) Kyoto University
3) Japan Atomic Energy Agency*E-mail: buchi@lane.iir.titech.ac.jp
A research project entitled “Study on accuracy improvement of fast-neutron capture

reaction data of long-lived MAs for development of nuclear transmutation systems” started in2017 as a four-year project. The purpose of the project is to improve the neutron capture crosssections of minor actinides in the fast neutron energy region that is particularly important forstudy on a nuclear transmutation system. The outline of the project and the current progress
are reported.

1. Introduction
Disposal of high-level nuclear waste from nuclear power plants has been a long-standing
issue in the nuclear industries. Nuclear waste contains long-lived minor actinides (MAs), someof which keep their radiotoxicities for more than thousand years. Currently planned geologicaldisposal of high-level nuclear waste has been in deadlock due to the difficulty of public
acceptance for many years. In order to solve this issue, nuclear transmutation, by which longlived MAs are converted to stable or shorter-lived nuclides via neutron-induced reactions, hasbeen investigated. In recent years, accelerator-driven systems (ADS) are considered as feasible
candidates of MA burners and several ADS projects are ongoing or planned. Detailed core
design of an ADS requires accurate, reliable nuclear reaction data of MAs but the uncertainties
of the current cross section data in evaluated nuclear data libraries in the fast neutron energy
region that is most relevant for ADS are not small enough to satisfy the requirement [1,2].

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JAEA-Conf 2020-001

A research project entitled “Study on accuracy improvement of fast-neutron capture

reaction data of long-lived MAs for development of nuclear transmutation systems” started in2017. The project aims at improving the accuracies of neutron capture cross sections of MAs(237Np, 241Am, 243Am) in the fast neutron energy region. The target uncertainty of cross section
measurement in this project is 5%. To improve the capture reaction data of MAs, an intense
pulsed neutron beam from a spallation neutron source of the Japan Proton Accelerator
Research Complex (J-PARC) is utilized in time-of-flight (TOF) experiments to measure capture
cross sections. In previous research projects, we worked on building and commissioning the
Accurate Neutron Nucleus Reaction Measurement Instrument (ANNRI), a neutron beam line
for nuclear data measurement in the Materials and Life Science Experimental Facility (MLF)
of J-PARC [3]. ANNRI now becomes one of the leading neutron beam lines for nuclear datameasurement in the world.
One of the major reasons that previous measurements of MA capture cross sections were
not able to achieve high accuracies is that MA samples were radioactive. A radioactive sample
emits decay -rays that become background for the detection of neutron capture -rays. The
large decay -ray background hinders accurate capture -ray measurement. The J-PARC highintensity pulsed neutron beam solves the issue by increasing neutron capture reaction rates
in a sample. The rate of capture events to background events can be improved, consequently
achieving small uncertainties of cross sections.
This project consists of four tasks: (1) development of a neutron beam filter system tosolve the so-called double bunch issue of a neutron beam in J-PARC, (2) neutron capture crosssection measurement, (3) sample characteristic assay, and (4) theoretical reaction model study.
The following sections describe details of the tasks.

2. Neutron Beam Filter System
The neutron beam filter is designed to solve the double bunch issue of a neutron beamfrom the J-PARC spallation neutron source. The spallation neutron source is operated at arepetition rate of 25 Hz. Hence, neutrons are generated every 40 ms. The J-PARC accelerator
adopts a special operational pattern called double bunch operation, in which two proton beampulses with a separation time of 0.6 s are injected into the spallation neutron target for eachneutron burst cycle. The purpose of this operation is to increase the thermal neutron intensity,
important for most of measurements with the neutron beam lines in MLF. In the thermal
neutron TOF range, the time structure of the incident proton beam is negligible. The Doppler
broadening effect and moderation time in a moderator erase the proton beam time structurein the thermal neutron TOF region over 12 ms. On the other hand, this double bunch mode isvery problematic for measurements in the high energy region above 100 eV, where the double
bunch structure appears in TOF spectra. Neutrons having two different energies originating-6-

JAEA-Conf 2020-001

from two incident proton pulses overlap at the same TOF channel in measurement. The
capture yield at each TOF channel is contributed by two different neutron energies and theresults cannot be deconvoluted easily.
A neutron beam filter system solves this issue. The neutron beam filter method is often
used in nuclear reactor experiments [4,5]. A reactor neutron beam which has a continuous
energy distribution can be tailored to be mono or quasi-mono energetic through filter materials
that have sharp resonance dips of total cross section at certain resonance energies. In thisproject, the neutron filter technique is combined with the TOF technique to separate out
coexisting different energy neutrons at the same TOF. The neutron beam filter system is underdevelopment. Filter materials Fe, Bi, Al, Si, Cr and Sc were chosen and tested in 2018. Basedon the test results, the system was designed and installed in the ANNRI beam line of J-PARC
MLF. A measured neutron energy spectrum filtered with Fe is shown in Fig. 1. A neutron
spectrum with no filter is also shown for comparison. As seen in Fig. 1, a prominent peak isclearly observed at an energy of 24 keV.

Fig. 1 Neutron energy spectrum filtered with Fe (red). A neutron spectrum with nofilter (black) is shown for comparison.

3. Neutron Capture Cross Section Measurement
This project focuses on fast-neutron capture cross section data of MA. This requires afast detection and data acquisition systems. Fast neutron events appear in fast TOF region
close to the gamma flash, an intense -ray emission produced at the moment that the incident
proton beam pulse reaches the spallation neutron target. The gamma flash overwhelms thedetection system and detection signals are distorted for s (sometimes ms) after the gamma
flash. To detect neutron capture events in the fast TOF region, the system needs to recoverquickly from the distortion caused by the gamma flash. In addition to the gamma-flash, an-7-

JAEA-Conf 2020-001

intense neutron beam from the J-PARC spallation neutron source increase the detector
counting rate, leading to large count loss due to the system dead time.
We plan to measure the neutron capture cross sections of MAs using NaI(Tl) detectors of
ANNRI. NaI(Tl) detectors are suitable for the measurement in fast TOF region [6].
Scintillation detectors have faster response than semiconductor detectors, and what’s more,
an NaI(Tl) detector can measure a -ray spectrum. The pulse height weighting technique toderive neutron capture cross sections is well established for a NaI(Tl) detector [7].
In addition, we developed a new method to determine the absolute neutron capture yields
using a sample rotation system. The effective sample thickness can be changed by tilting thesample with respect to the beam axis. The self-shielding factor changes with the effective
sample thickness. The absolute neutron capture yield is determined from the self-shielding
factor that is calculated from change of capture yield by tilting the sample.

4. Sample Characteristic Assay
Sample characteristic assay is an important task to improve the accuracy of crosssections. Uncertainties of sample characteristics such as total mass, isotope composition andimpurities can be crucial systematic uncertainties. In this project, we plan to analyze theisotope composition and impurities of MA samples by thermal ionization mass spectrometry(TIMS) at the Institute of Integrated Radiation and Nuclear Science of Kyoto University. The
target accuracy of analysis is set at less than 1% in this project. The key to achieving such ahigh accuracy is the stability of ion emission from a filament of the TIMS ion source. To
stabilize the ion emission, the monitoring method of the filament temperature, most dominantfactor for ion emission, was improved in 2018. A test experiment showed significant
improvement of ion emission control. TIMS analysis of MA samples is planned to conduct in2020.

5. Theoretical Reaction Model Study
Theoretical nuclear reaction models can predict neutron capture cross sections.
Combined with experimental data, theoretical reaction models become powerful tools in
nuclear data evaluation. In this project, neutron capture -ray spectra measured with the
ANNRI-NaI(Tl) detectors are used to refine theoretical model prediction. Capture -ray spectra
give more information on reaction mechanisms than only from capture cross section.
Comparing theoretical calculations with the measured -ray spectra, model parameters suchas gamma-ray strength function and level density can be determined. However, direct output
spectra from experiments cannot be compared with theoretical capture -ray spectra because
measured spectra convolute detector response. Unfolding measured spectra with detector
response function is often performed to compare with theoretical spectra [7] but the unfolding-8-

JAEA-Conf 2020-001

process adds uncertainties to experimental data. Instead, we adopted the reverse process forcomparison. We fold theoretical spectra with detector response and then compare them toexperimental data. Folding process is less ambiguous than unfolding process. We built ageometrical model of the ANNRI-NaI(Tl) detectors for the Monte Carlo simulation code PHITS[8] and calculated the detector response matrix. For a benchmark calculation, the capture ray spectrum of 197Au was calculated with the theoretical reaction model code CCONE [9] andthen folded the calculated spectrum with the detector response matrix

6. Summary and Future Prospect
The project entitled “Study on accuracy improvement of fast-neutron capture reaction

data of long-lived MAs for development of nuclear transmutation systems” started in 2017 asa four-year project. The first two years were spent for development of the neutron beam filter
system, fast data acquisition method for the ANNRI-NaI(Tl) detectors and MA sample
characteristic assay. The neutron filter system and the sample rotation system were installed
in ANNRI. Now the project is in the actual experimental phase to measure the capture crosssections and capture -ray spectra of MAs.

Acknowledgements
This work is supported by the Innovative Nuclear Research and Development Programfrom the Ministry of Education, Culture, Sports, Science and Technology of Japan. The neutron
experiment at the MLF of the J-PARC was performed under the user program (Proposal No.
2018A0213 and 2018B0011).

References[1] M. Salvatores et al ., OECD/NEA Report No.6410 (2008).
[2] H. Iwamoto et al., J. Nucl. Sci. Technol., 50, 856 (2013).
[3] K. Kino et al ., Nucl. Instr. Meth. A, 736, 66 (2014).
[4] O. D. Simpson and L. G. Miller, Nucl. Instr. Meth, 61, 245 (1968).
[5] A. T. Tran et al ., Ann. Nucl. En., 113, 420 (2018).
[6] T. Katabuchi et al., Nucl. Instr. Meth. A, 764, 369 (2014).
[7] S. Mizuno et al., J. Nucl. Sci Technol., 36, 493 (1999).
[8] T. Sato et al., J. Nucl. Sci Technol., 55, 684 (2018).
[9] O. Iwamoto et al., Nucl. Data Sheets, 131, 259 (2016).

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JAEA-Conf 2020-001

3

Recent progress and future plan on the JENDL project
Osamu IWAMOTO
Nuclear Data Center, Japan Atomic Energy Agency
Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 Japane-mail: iwamoto.osamu@jaea.go.jp

The JENDL project has been developing evaluated nuclear data libraries to meet the needs ofnuclear science and engineering. JENDL provides general and special purpose files generally for nuclear
reactor applications and topically for other applications that cannot be covered by the general purpose file,
respectively. Owing to the broadened needs of the nuclear data, many special purpose files have been released
since 2010 in which JENDL-4.0 was released. Concerning the general purpose file, development of the next
version JENDL-5 is in progress. In addition to the revision of the data ranging from light to heavy nuclei, the
new evaluation for whole stable isotopes lacked in JENDL-4.0 are being undertaken to meet the needs forvarious applications. The test libraries of JENDL-5 are produced to get feedbacks from integral experimental
data. The recent progress of JENDL is summarized in this report.

1. Introduction
The latest version of JENDL general purpose file, i.e. JENDL-4.0, was released in 2010 [1]. It wasdeveloped intending to improve the data of fission products and minor actinides that would be important for
research and development of light water reactors with high burn-up and MOX fuels as well as of innovative
reactors such as accelerator driven system. Since then, many special purpose files have been released: JENDLFP Decay Data File 2011 (JENDL/FPD-2011), JENDL FP Fission Yields Data File 2011 (JENDL/FPY-2011),
JENDL-4.0 High Energy File (JENDL-4.0/HE), JENDL Decay Data File 2015 (JENDL/DDF-2015), JENDL
Photonuclear Data File 2016 (JENDL/PD-2016), JENDL Activation Cross Section File for Nuclear
Decommissioning 2017 (JENDL/AD-2017), JENDL LLFP Transmutation Cross Section File 2018(JENDL/ImPACT-2018) [2]. These files were developed to meet needs from expanding fields of applications
such as backends of nuclear energy and accelerator utilizations.
As well as the special purpose files, development of the next version of general-purpose file
JENDL-5 is in progress. JENDL-5 is aiming at increasing completeness and reliability from JENDL-4.0 in
viewpoints of target nuclide species and data uncertainties. The data will be evaluated with reflecting current
knowledge of cross section measurements, nuclear theory and integral benchmark tests. A new R-matrix
theory code AMUR has been developed and applied to the resonance analysis for light nuclides [3]. Structural
materials and medium-heavy nuclides have been evaluated using the modern nuclear reaction model code
CCONE [4,5]. The first test library JENDL-51 was created last year and its integral benchmark tests were

- 11 -

JAEA-Conf 2020-001

performed. The next version of the test file JENDL-52 is being prepared. JENDL-5 is planned to be released
in FY 2021.

2. Special purpose file
2.1 JENDL/PD-2016
Photonuclear data are important for shielding design of electron accelerator as well as gamma-ray
therapy. Since the number of nuclei in the previous version of photonuclear data library JENDL/PD-2004 [6]was limited to 68, the new version JENDL/PD-2016 [7] was developed with increasing the number of target
nuclides applying the current theoretical models and available experimental data. JENDL/PD-2016 provides
2681 nuclides in total from Z=1 to 93 covering unstable isotopes in the expanded version, and the standard
version with the data for 181 nuclides along the beta-stability line is also prepared for convenience. The
secondary particle productions with energy-angle distributions have been evaluated for nucleons and lightions. The data of production cross sections of residual nuclides are also stored. The incident photon energy
ranges from 1 to 140 MeV.
In connection with the Coordinated Research Project about update of photonuclear data library of
IAEA, new evaluations of photonuclear data have been performed using newly measured data. Updating thedata of JENDL/PD-2016 with these new evaluations, the revised version JENDL/PD-2016.1 was released in
February 2020 [2].

2.2 JENDL/AD-2017
For decommissioning of nuclear installations that is expected to increase remarkably in near future,reliable evaluation of activation inventories of the facilities is essential to implement reasonable plan
according to levels of the activities. The activation cross sections of the constituent materials are needed forthis evaluation. However, more than 20 years have already passed since JENDL released the previous versionof the activation file JENDL/A-96 [8] in 1996. We started development of a new activation cross section
library focusing on the decommissioning of light water reactors with the framework of the joint research withJapan Atomic Power Company in 2011. The 221 radioactive isotopes were selected in terms of importancefor dose and clearance evaluations having half-lives longer than 30 days. The cross sections for 311 nuclides
which would produce those isotopes were compiled into the ENDF-6 format and was released as JENDL/AD2017 [9]. While some data were updated based on the latest version of general-purpose file JENDL-4.0 and
supplemented with JEFF-3.1/A, a large part of the data were newly evaluated with the available
measurements and the current theoretical model. The library provides the point-wise cross sections at 0 K
and 293.6 K for energy range from 10-5 eV to 20 MeV for production of radioactive nuclei including isomers.

2.3 JENDL/ImPACT-2018
Fission product nuclear data library JENDL/ImPACT-2018 [10] was developed under the program“Reduction and Resource Recycling of High-level Radioactive Wastes through Nuclear Transmutation”- 12 -

JAEA-Conf 2020-001

funded by the ImPACT Program of the Council for Science, Technology and Innovation (Cabinet Office,
Government of Japan). This ImPACT program aimed at producing innovations on issues of high-level
radioactive wastes by reducing and recycling long-lived fission products by transmutation using accelerators.
In addition to the data for LLFPs of 79Se, 93Zr, 107Pd and 135Cs, JENDL/ImPACT-2018 covers the data forsecondary products that would be produced via transmutation of LLFP using reactions in hundreds MeV
region. The library provides the data of energy-angle distributions for nucleon and light particle emissions aswell as cross sections of residual nuclei productions up to 200 MeV. The data was evaluated using thecomprehensive nuclear reaction evaluation code CCONE with enhancing prediction accuracies for residual
nuclide productions of proton induced reactions around a few hundred MeV that were measured by inverse
kinematics at RIKEN under the same program [11, 12].

3. General purpose file
The next version of general purpose file JENDL-5 is under development. The data in wide rangeof nuclides from light to heavy nuclides have been being evaluated.
Regarding light nuclides, the neutron resonance data of 16O, 15N and 19F were analyzed using the
newly developed R-matrix code system AMUR. Taking account of the data for inverse channels creating thesame compound nucleus, the reliability of the data has been increased. From the resonance analysis, theresonance cross sections as well as their covariances were deduced. JENDL-4.0 contains old data forstructural materials that were evaluated in 1980’s in spite of several revisions from the first evaluation. The
CCONE code has been used to update those data taking account of available experimental data of isotopesof Mn, Cu, Zr and Nb [13-15].
Increase of reliability of the data for fission products and minor actinides is still important becauseof their large uncertainties in evaluated data and needs of managements of those nuclear wastes. Due to thelarge number of nuclides in the fission products, a part of them were not yet revised in the release of JENDL4.0 in spite of making much efforts on the revision of them. Remaining nuclides of fission products of Ga,
Tc, Sb, Te, I and Er isotopes have been evaluated for JENDL-5 [16-19]. Concerning minor actinides,
intensive works to improve accuracy of nuclear data especially on 241,243Am were conducted under the
AIMAC project [20]. The accurate data taken by ANNRI at J-PARC under this project have been used torevise the resonance parameters of Am isotopes.
New experimental data recently becomes available not only for Am but also for other actinides.
Fission cross section for 242Pu, which is listed in NEA Nuclear Data High Priority Request List, was measured
at several facilities in the world after around 2010. Some of those measurements suggest the current evaluateddata are overestimated. The integral experiments of fission reaction rate with fast reactors also suggest
overestimation of 242Pu fission cross section in JENDL-4.0. Taking into account the recent measurements,
the fission cross section has been reevaluated with the least squares fitting code SOK [21]. The evaluatedresults are compared with recent experimental data and the evaluation of JENDL-4.0 in Fig. 1. While theevaluated result with all experimental data is very close to that of JENDL-4.0, the cross section deduced from- 13 -

JAEA-Conf 2020-001

recent experimental data significantly decreases around 1 MeV. The fission reaction rate ratio of 242Pu to239

Pu is estimated using evaluated data from recent measurement with the sensitivity matrix of fission crosssection of 242Pu prepared for ADJ2017 and cross section differences from JENDL-4.0. The deduced results
are shown in Fig. 2. The overestimations seen in JENDL-4.0 clearly become small. As some overestimations
still exist for FCA experiments, further study including contribution from other nuclear data would be needed.
The first test library of JENDL-5α1 was created in FY 2018 and the second one JENDL-5α2 isbeing prepared in FY 2019. They include the updated and newly-evaluated data for more than 90 isotopes.
The preliminary evaluation of fission cross sections of major actinides i.e. uranium and plutonium isotopes
for the fast neutrons are included. The results of the recent international collaboration CIELO [22] are taken
for these test versions. The thermal scattering law data evaluated by Abe et al. [23] is adopted. Feedbacks
from benchmark results for reactors will be taken into accounts in the revision of the JENDL-5 evaluations.

4. Conclusion
Recent progress and future plan of JENDL are summarized. The JENDL project has released many
special purpose files in this decade. Outlines of the most recent 3 files, JENDL/PD-2016, JENDL/AD-2017
and JENDL/ImPACT-2018 are reported. The next version of general purpose file, JENDL-5, is underdevelopment with the plan of the release in FY 2021. Evaluations from light to heavy nuclides are in progress.
Test libraries are being produced with including new evaluations for light nuclei, structure materials, fission
products, actinides etc.

References[1] K. Shibata, O. Iwamoto, T. Nakagawa, N. Iwamoto, A. Ichihara, S. Kunieda, S. Chiba, K. Furutaka, N.
Otuka, T. Ohsawa, T. Murata, H. Matsunobu, A. Zukeran, S. Kamada, J. Katakura, “JENDL-4.0: A New
Library for Nuclear Science and Engineering”, J. Nucl. Sci. Technol., 48(1), 1-30 (2011).
[2] https://wwwndc.jaea.go.jp/jendl/jendl.html#jendl-sp (accessed 2020/09/23).
[3] S. Kunieda, “Status of the R-matrix Code AMUR toward a consistent cross-section evaluation andcovariance analysis for the light nuclei”, EPJ Web of Conferences, 146, 12029 (2017).
[4] O. Iwamoto, “Development of a Comprehensive Code for Nuclear Data Evaluation, CCONE, andValidation Using Neutron-Induced Cross Sections for Uranium Isotopes”, J. Nucl. Sci. Technol., 44(5),687-698 (2007).
[5] O. Iwamoto, N. Iwamoto, S. Kunieda, F. Minato, K. Shibata, “The CCONE Code System and its
Application to Nuclear Data Evaluation for Fission and Other Reactions”, Nucl. Data Sheets 131, 259288 (2016).
[6] N. Kishida et al., “JENDL Photonuclear Data File”, Proceedings of International Conference on Nuclear
Data for Science and Technology, Santa Fe, New Mexico, USA, Sep. 26 - Oct. 1, 2004, 1, 199 (2004).
[7] N. Iwamoto, K. Kosako, T. Murata, “Photonuclear Data File”, JAEA-Conf 2016-004, 53-58 (2016).
[8] Y. Nakajima, JNDC WG on Activation Cross Section Data, “JENDL Activation Cross Section File”,- 14 -

JAEA-Conf 2020-001

Proc. the 1990 Symposium on Nuclear Data, JAERI-M 91-032, 43 (1991).
[9] K. Shibata, N. Iwamoto, S. Kunieda, F. Minato, O. Iwamoto, “Activation Cross-section File for
Decommissioning of LWRs”, JAEA-Conf 2016-004, 47-52 (2016).
[10] S. Kunieda, N. Furutachi, F. Minato, N. Iwamoto, O. Iwamoto, S. Nakayama, S. Ebata, T. Yoshida, K.
Nishihara, Y. Watanabe, K. Niita, “JENDL/ImPACT-2018: A New Nuclear Data Library for Innovative
Studies on Transmutation of Long-lived Fission Products”, J. Nucl. Sci. Technol., 56(12), 1073-1091(2019).
[11] H. Wang, H. Otsu, H. Sakurai et al., “Spallation reaction study for the long-lived fission product 107Pd”,
Prog. Theor. Exp. Phys., 2017(2), 021D01 (2017).
[12] S. Kawase, K. Nakano, Y. Watanabe et al., “Study of proton- and deuteron-induced spallation reactions
on the long-lived fission product 93Zr at 105 MeV/nucleon in inverse kinematics”, Prog. Theor. Exp.
Phys., 2017, 093D03 (2017).
[13] N. Nakayama, “Calculation of neutron cross-sections on copper-63 and -65”, J. Nucl. Sci. Technol.,55(6), 614-622 (2018).
[14] A. Ichihara, “Theoretical calculation of neutron cross sections for 90,91,92,94,96Zr in the incident energy
range between 200 keV and 20 MeV”, J. Nucl. Sci. Technol., 55(9), 1087-1098 (2018).
[15] A. Ichihara, “Calculation of cross sections for metastable state production in the (n,γ), (n,n'), (n,2n) and(n,3n) reactions of 93Nb”, J. Nucl. Sci. Technol., 53(12), 2049-2055 (2016).
[16] K. Shibata, “Evaluation of neutron nuclear data for gallium”, J. Nucl. Sci. Technol., 50(3), 277-286(2013).
[17] K. Shibata, “Evaluation of neutron nuclear data on antimony isotopes”, J. Nucl. Sci. Technol., 51(4),425-436 (2014).
[18] K. Shibata, “Evaluation of neutron nuclear data on tellurium isotopes”, J. Nucl. Sci. Technol., 52(4),490-502 (2015).
[19] K. Shibata, “Evaluation of neutron nuclear data on iodine isotopes”, J. Nucl. Sci. Technol., 52(9), 11741185 (2015).
[20] H. Harada, O. Iwamoto, N. Iwamoto, A. Kimura et al., “Research and development for accuracy
improvement of neutron nuclear data on minor actinides”, EPJ Web of Conferences, 146, 11001 (2017).
[21] T. Kawano, H. Matsunobu, T. Murata et al., “Evaluation of Fission Cross Sections and Covariances for
233

U, 235U, 238U, 239Pu, 240Pu, and 241Pu”, JAERI-Research 2000-004 (2000).

[22] M.B. Chadwick et al., “CIELO Collaboration Summary Results: International Evaluations of Neutron
Reactions on Uranium, Plutonium, Iron, Oxygen and Hydrogen”, Nucl. Data Sheets, 148, 189 (2018).
[23] Y. Abe, “Evaluation of the neutron scattering cross-section for light water by molecular dynamics”, Nucl.
Instrum. methods Phys. Res. A, 735, 568 (2014).

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JAEA-Conf 2020-001

Fig. 1 Neutron induced fission cross section of 242Pu. Blue and red curves with circles shows the evaluatedresults with all experimental data and ones limited after 2000, respectively. Black line indicates JENDL4.0. Recent experimental data are shown by gray symbols.

Fig. 2 Fission reaction rate ratio of 242Pu to 239Pu for fast reactors. C/E values for JENDL-4.0 is shown by
black squares and estimation for new evaluation with sensitivities is indicated by red circles.

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JAEA-Conf 2020-001

4

A Consideration on Nuclear Data Needs for 1F Decommissioning
Tokio FUKAHORIFukushima Research Institute, Japan Atomic Energy Agency
Taira Central Building, Iwaki, Fukushima, 970-8026 Japane-mail: fukahori.tokio@jaea.go.jp

After 8 years from the accident at TEPCO Fukushima Daiichi Nuclear Power Station,
the “on-site” research and development (R&D) situation has been changed into new phase. It
can be considered from understanding to predicting by many direct and indirect measurements
and analyses. In this presentation, nuclear data needs for 1F decommissioning are revisited
according to these situation. The needs for accident evolution characteristics, criticality
management, debris storage, waste management, etc. are considered with the view point of
R&D on prediction what is going on and will be happened in each stage of decommissioning.

1. Introduction
The accident at TEPCO Fukushima Daiichi Nuclear Power Station (1F) was triggeredby the earthquake named “Great East Japan Earthquake” on March 11, 2011. The research anddevelopment (R&D) situation for 1F decommissioning and environment remediation has beenchanged into new phase and weight of the R&D is getting to shift into the “on-site”, since thesituation related to “off-site” is getting more stable. For example, the dose rate at many places
has been down to the normal level compared with other cities[1], the number of evacuating
people is decreased to 1/4[2], and the evacuation area is reduced 1/3[2,3].
On the other hand, the locations of debris in the unit 1-3 are believed now to beunderstood much better than before. The fuels in spent fuel pools have been started to moveout. The piece being seemed debris has been directly touched by the new equipment. The
condition inside the reactor buildings can be considered to change from understanding topredicting by many direct and indirect measurements and analyses.
In this presentation, nuclear data needs for 1F decommissioning are revisited according
to the situation mentioned above. The needs for understanding accident development,
understanding current situation in the reactor building and waste management of rubble andfuel debris are tried to be considered with the view point of R&D on prediction what is goingon and will be happened in each stage of decommissioning. A consideration by the author about
above items with seeing criticality management, debris storage, waste management, heat
reduction, radiation shielding, hydrogen production by radiolysis, etc.

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JAEA-Conf 2020-001

2. Understanding Accident Progression
For understanding accident progression, the calculation starts at the point of final burnup condition. The decay heat makes cooling water evaporating and fuel rods and structural
materials melting. To estimate final burn-up condition, ordinal burn-up calculation is enough.
For this purpose, quantities related to fission reaction for reactor operation are needed as wellas fission yields and decay data for fission product (FP) accumulation which is source of decay
heat. It seems to be considered these nuclear data have enough accuracy, while the informationof reactor operation, fuel combination and shuffling is more important for more precise
calculation.
The fuel debris composition can be estimated by melted materials’ transfer. This
calculation strongly depends on the accident progression calculation results mentioned above.
It is also important to consider debris re-melting after solidification. For this calculation, fission
yields and decay data are also important, because of heating source estimation. These nuclear
data and nuclear structure data are also useful for the heat reduction (cooling) calculation.
3. Understanding Current Situation inside Reactor Building
To estimate current situation inside the reactor buildings, simulation of source term,
criticality, decay heat and radiation distribution are necessary. For source term calculation,
information of fuel debris and FP nuclide locations is needed to be estimated. For this purpose,
data of quantities related to fission reaction, fission yields and decay data are needed.
For the criticality management, amount and location of fissionable nuclides are needed
to be obtained as well as FP nuclides for burn-up credit calculation. Quantities related to fission
reaction and fission yields are important information. To monitor criticality, high energy
gamma-ray from capture state is useful, since it has lower background comparing to the low
energy one. It is also used for “Capture Credit” estimation[4,5]. To estimate the high energy
gamma-ray production, capture cross section and its gamma-ray emission spectra are necessary.
Unfortunately, it is hard to say the gamma-ray spectrum data are enough both qualitatively andquantitatively at this moment.
Heat reduction is important to control the temperature of cooling water. To calculate
heat source, the data of fission yields, decay and nuclear structure data are necessary. They arealso used for radiation shielding calculation with total cross sections of structural materials.
4. Waste Management of Rubble and Fuel Debris
The waste management is important to be considered before starting the rubble andfuel debris bring-out procedure. Considered are criticality management of debris storage,
radiation shielding, heat reduction, radiolysis, etc. For the criticality management, nuclear dataneeds are same as mentioned in “3. Understanding Current Situation inside Reactor Building”.

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For estimation of radiation production, the composition of fuel debris and FP
production data are necessary as radioactive sources. Considering activation cross sections forstructural materials are also important for radiation shielding calculation. In the case of theradiation shielding and heat reduction needs similar nuclear data as well as total cross sectionof structural materials.
Radiolysis[6], especially hydrogen-gas production, must become important for safety
storage, since the waste at 1F has significant radioactivity. For radiation source estimation,
fission yields, decay and nuclear structure data are needed. In addition, emitted particle spectra,
especially for alpha-particles, are necessary to calculate hydrogen-gas production rate in thewater included inside the storage. The alpha-particle spectra are also used to detect U/Pu
particles[7] which are alpha emitter. It is important for Pu-signal separation from those of radon
progeny to identify the higher energy alpha-particles. The spectra can be used for calculationsof alpha-particle transport both in the debris and water.
5. Summary
One of considerations on nuclear data needs for 1F decommissioning is introduced forunderstanding accident development, understanding current situation inside reactor building
and waste management of rubble and fuel debris. As the summary of nuclear data need,following physical quantities are picked up.
 Quantities Related to Fission Reaction
 Fission Yields
 Decay and Nuclear Structure Data
 Total Cross Section for Structural Materials
 Activation Cross Section for Structural Materials
 High Energy Gamma Production from Capture State
 Emitted Particle Spectra (especially for Alpha-particle)
Those above are one of considerations by the author. More needs might be hidden according todevelopment of new technology for decommissioning.

References[1] Fukushima Prefecture Radioactivity Measurement Map, http://fukushima-radioactivity.jp/pc/(those of foreign cities https://www.env.go.jp/chemi/rhm/h30kisoshiryo/h30kiso-02-05-07.html)(accessed 2020/07/03).

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JAEA-Conf 2020-001[2] Fukushima Prefecture, “Steps for Revitalization in Fukushima 26th edition <August 5,
2019>” and “Damage Report <October 7, 2019>”,https://www.pref.fukushima.lg.jp/site/portal-english/ayumi-en-15.html (accessed 2020/07/03).
[3] Ministry of Economy, Trade and Industry of Japan, “Evacuation Areas”,
https://www.meti.go.jp/english/earthquake/nuclear/roadmap/evacuation_areas.html(accessed 2020/07/03).
[4] Y. Nauchi, “Nuclear Data Required for Measurements of Reactivity and Nuclear Material
Composition”, JAEA-Conf 2019-001, p.41-46 (2019).
[5] Y. Nauchi, “Nuclear Data Needs Related to Estimation of Fuel Debris Nuclear
Characteristics -Measurement of Gamma-ray Spectrum by Neutron Irradiated Nuclear
Reaction -”, Nuclear Data News (ISSN 0385-4876), No.122, p.54-66 (2019) in Japanese.
[6] R. Nagaishi, “Hydrogen Production in Storage of Fuel Debris”, private communication.
[7] Y. Morishita, T. Torii, H. Usami, H. Kikuchi, W. Utsugi and S. Takahahira, “Detection of
Alpha Particle Emitters Originating from Nuclear Fuel inside Reactor Building ofFukushima
DaiichiNuclearPower
Plant”,
Scientific
Reports
www.nature.com/scientificreports), DOI: 10.1038/s41598-018-36962-4 (2019).

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5 Fundamental physics using neutron at J-PARC andaccelerator-driven neutron source
Tamaki Yoshioka1
1

Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
E-mail: yoshioka@phys.kyushu-u.ac.jp

Precise measurements using low energy neutron give complementary information to the collider
experiments using high energy accelerators. Moreover, it is also sensitive to a new physics in highenergy regions where the collider experiments may not reach. A high-quality neutron beam can beprovided to an experimental apparatus by using the neutron optics, which has been greatly advanced
in recent years. A beamline BL05, which is intended for the fundamental physics using low energy
neutron, has been constructed at the J-PARC/MLF, and successfully started the operation at the end of
2008. In this article, the experiments using low energy neutron conducted at the J-PARC/MLF/BL05
will be given. In addition, the recent situation of an accelerator-driven compact neutron source will
also be presented in this article.

1. Introduction
The Higgs boson, which is the last piece predicted in the Standard Model (SM) of the particle
physics, was discovered in 2012 [1, 2]. Some phenomena are known that can not be explained by theSM, such as the existence of the dark matter. There is, however, no evidence for a physics beyond theSM (BSM) despite the intensive search at Large Hadron Collider (LHC). So-called the energy frontier
experiment is aiming to explore the BSM directly by using the high energy accelerator such as the LHC,
as stated above. In contrast, the search for the BSM by the precision experiment using high-intensity secondary particles such as the neutron is an indirect approach. They are complementary with each other,
but the indirect search sometimes more sensitive than the direct approach for some specific BSM models.
The Japan Proton Accelerator Research Complex (J-PARC) is a joint project of the Japan Atomic
Energy Agency (JAEA) and High Energy Accelerator Research Organization (KEK) and is constructedat Tokai-village, Ibaraki-prefecture [3]. Figure 1 is a schematic view of the J-PARC, and a fundamental
physics experiment using low energy neutron is conducted at the Material and Life science Facility(MLF) located at the center of the figure.

2.

Fundamental physics using low energy neutron

The neutron was discovered by Chadwick in 1932 and is a composite particle consisting of two up
quarks and one down quark. The neutron has a spin of 1/2 and a mass of about 1 GeV/c2 , and decays
to an electron and two neutrinos via weak interaction with a lifetime of 880.3 seconds [4]. The classification of neutrons, according to their energy, is shown in Fig. 2. Cold neutrons, which are often used infundamental physics experiments at J-PARC, have an energy of about meV. Fundamental physics experiments also use very cold and ultracold neutrons in the even lower energy range, and such low energy
neutrons have unique features. Low-energy neutrons can be confined to a material bottle because they
are totally reflected by the nuclear potential called the Fermi potential which is about several hundred
neV for various materials. As mentioned earlier, the neutron has a spin 1/2, and the potential diﬀerence
between the spin-up and spin-down states is about 120 neV in a magnetic field of 1 T, which is easily
achievable in the laboratory. The gravitational interaction is usually ignored in particle physics, but the- 21 -

JAEA-Conf 2020-001

Rapid Cycle Synchrotron

LINAC

Main Ring

Hadron Experimental
Facility

Materials and Life
science Facility(MLF)

Fig. 1. A schematic view of the J-PARC. A building located at the center of the figure is called Materials andLife science Facility (MLF) where fundamental physics experiment using low energy neutron is conducted.

positional energy acquired by neutrons is about 100 neV/m, which is the same order of magnitude as theinteractions described above. Moreover, as also mentioned earlier, neutrons decay through weak interactions but have an extremely long lifetime for an unstable particle. We prove the high-energy phenomenain the quantum loop via the precise measurement by utilizing these unique features.

Fig. 2. Classification of neutrons according to their energy. Cold neutrons, which are often used in fundamental
physics experiments at J-PARC, have an energy of about meV.

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JAEA-Conf 2020-001

In the J-PARC/MLF, a proton beam from the Rapid Cycle Synchrotron shown in Fig. 1 is irradiatedto a mercury target to produce neutrons, which are moderated by a hydrogen moderator and transported
to each beamline. There are 23 beamlines in total in MLF, among which beamline number 5 (BL05)
named Neutron Optics and Physics (NOP) is a beamline dedicated to fundamental physics experiments
using low energy neutrons [5]. The BL05 has a feature of dividing a neutron beam from a mercury
target to three diﬀerent beam branches by a neutron optics device called a super mirror bender. Figure 3
shows a picture of the super mirror bender before being installed inside the concrete shield. From theleft, there are polarization beam branch, un-polarized beam branch, and low-divergence beam branch,and the neutron beam from the mercury target is injected from the front of the photograph. As can beseen from the picture, each beam branch is curved to reflect low energy neutrons and transport themto the experimental area. A number of physics experiments and device R&D have been carried out at
BL05 [6–9]. Among them, the flagship experiment, precision measurement of the neutron lifetime usinga polarized beam branch, is described in the next section.

Polarized

Low-divergence
Un-polarized

Fig. 3. A picture of the super mirror bender before being installed inside the concrete shield. From the left, there
are polarized beam branch, un-polarized beam branch, and low-divergence beam branch, and the neutron beamfrom the mercury target is injected from the front of the photograph.

3.

Precise measurement of neutron lifetime

The neutron lifetime has been measured for more than 50 years since its discovery. The neutron lifetime is an important input parameter for the Big Bang Nucleosynthesis (BBN). It can also be used for atest of the unitarity of the Cabbibo-Kobayashi-Makawa (CKM) matrix. There are historically two typesof methods to measure the neutron lifetime; storage method and beam method. The former confines theultracold neutron in a material bottle as described earlier for a certain period of time and counts the number of the survived neutron. The latter utilizes the cold neutron beam and counts the number of incident
neutrons and decay protons. Each method has a measurement precision of less than 1%, however, there isa long-standing significant discrepancy corresponds to 8.5 sec. or 4.0 standard deviation. Recently, there
is an attempt to explain this discrepancy by the BSM. It cannot be, however, ruled out that the diﬀerenceis due to some unknown systematics, a new type of experiment is therefore desired in this regard.

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JAEA-Conf 2020-001

The precise measurement of the neutron lifetime at J-PARC is a kind of beam method, but we countelectrons instead of decay protons. In addition, in the previous beam method, the incident neutrons anddecay protons were counted with diﬀerent detectors, and the resulting uncertainty was considered to be
dominant. In our experiment, the incident neutrons and decay electrons are measured simultaneouslywith a single gaseous detector called a Time-Projection Chamber (TPC). Originally, this type of experiment was developed by Kossakowski et al. [10], the following cares are employed in our experimentin order to suppress background events. The TPC is made of a material with small radioactive material
contamination [11]. The neutron beam provided by the polarization branch is further chopped by using
a special device called Spin Flip Chopper (SFC) so that the neutron bunches are fully contained in the
TPC. We started data taking in 2014, and the first result was obtained with an accuracy of a couple ofpercents. In the future, we are aiming to achieve 0.1% accuracy with an improvement of the SFC and a
suppression of the dominant background caused by the TPC operation gas.
We are also planning yet another neutron lifetime measurement using a superconducting magnet. In
the course of the data analysis in the current experiment, we found the dominant background source isoriginated from the TPC operation gas, as mentioned above. Geant4 simulation shows that this background can be significantly suppressed to achieve 0.1% accuracy with a 600 mT magnetic field. The
idea of this experiment can be found in Ref. [12] in detail. A dedicated TPC with non-magnetism materials was constructed at Kyushu University for this experiment. An integration test with the TPC andsuperconducting magnet was conducted at KEK using cosmic rays and a radioactive source. Figure 4
shows the integration test at the KEK. The TPC is installed on the dedicated stand shown in the centerof the photograph. Behind the TPC, the superconducting magnet can be seen, and a vacuum chamber
is installed inside of the magnet. In this test, we confirmed that the TPC has the expected performancein a magnetic field. As of this written, the TPC and superconducting magnet are installed at the BL05
beamline. The magnet excitation test was successfully done, and we took a first commissioning data with
the neutron beam.

Superconducting
MagnetVacuumChamber

TPC

Fig. 4. Integration test at KEK. The TPC is installed on the dedicated stand shown in the center of the photograph. Behind the TPC, the superconducting magnet can be seen and a vacuum chamber is installed inside of themagnet.

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JAEA-Conf 2020-001

4.

Accelerator-driven compact neutron source at Kyushu University

Neutron beam has been used not only in particle and nuclear physics experiments but also for industrial and medical applications, recently. As described earlier, a high-intensity spallation neutron source
has been in operation at J-PARC since 2008. Although such a large facility enables ultra-precise measurements with high-intensity beams, it takes several months from the proposal submission to the actual
experiment, and the allocated machine time is rather limited. Comparisons between large and small facilities are summarized in the Table I. As shown in the table, they are complementary to each other,which is similar to the relationship between SPring-8, a large synchrotron radiation facility, and an Xray device in a university laboratory as a small facility. Although the development of compact neutron
sources is widely spreading, it is ineﬃcient to develop in one institute/laboratory since there are a numberof R&D items. The Japan Collaboration of Accelerator-driven Neutron Sources (JCANS) was consolidated to activate a nation-wide network of individual research activity on neutron source and moderators.
Accelerator-driven compact neutron sources are already in operation at Hokkaido University, Kyoto University, and RIKEN, and are now being developed at Nagoya University [13].
Fast neutrons have been produced so far by using the tandem accelerator at Kyushu University. Fastneutrons are useful for testing a soft error of semiconductor, but lower energy (thermal and cold) neutrons
are preferred in particle physics and materials science. We have firstly demonstrated to produce thethermal neutrons at Kyushu University in July 2019. An aluminum duct was connected to downstream ofthe scattering chamber in the experimental hall. A blank flange was attached to downstream of the duct,
and a neutron production target was placed inside of the flange. Fast neutrons produced by the proton
beam irradiation from the tandem accelerator are slowed down by a polyethylene moderator installed
downstream of the target, and the thermal neutrons are detected by a neutron detector located at themost downstream of the beamline. We chose a Berylium metal as a neutron production target. We expect
4 × 107 neutrons/sec by assuming proton energy of 13 MeV and the beam current of 1 nA of the tandem
accelerator. Figure 5 is an image of the two-dimensional detector, in which a cadmium plate with a "Q"
shape is attached to the center of the surface of the detector, and a boron gum is attached to the bottom of
the detector. Left figure of Fig. 5 shows a detector image with the polyethylene moderator, and the
shadows of cadmium and boron gum can be seen due to the absorption of thermal neutrons. Right figureof Fig. 5 is the case without the moderator, and the shadows can not be seen because fast neutronstransmitted the cadmium and boron gum. This indicates that the produced fast neutrons weresuccessfully slowed down in the moderator and became thermal neutrons. In the next step, we will installa mechanical chopper to measure the time of flight. We are also aiming to produce the cold neutrons byfarther cooling the thermal neutrons in the future.

Table I. Pros and cons of the accelerator-driven compact neutron source.

Large Facility

Small Facility

Pros• Precision measurement thanks to highintensity beam
• Well maintained beamline• Relatively plenty/flexible beam time
• Education use• Test experimentCons• Limited resource/beam time/operationmode• Take long time from proposal stage• Weak beam intensity• Need setup every time- 25 -

JAEA-Conf 2020-001

Fig. 5. Results of the experiment of thermal neutron production at Kyushu University. Left (Right) figure showsa 2D detector image with (without) a polyethylene moderator.

5.

Summary

Precise measurements using low-energy neutrons provide complementary information to the collider
experiments using the energy frontier accelerator. A beamline named BL05 for the fundamental physics
using low energy neutron has been constructed in the MLF of the J-PARC and operated since 2008.
Currently, a precise measurement of the neutron lifetime is on-going at the J-PARC/MLF/BL05 as aflagship experiment. The accelerator-driven compact neutron source is useful for such as a detector R&D
before conducting an experiment at the J-PARC. A test experiment of the thermal neutron production
has been successfully carried out by using the tandem accelerator at Kyushu University. We will try to
produce cold neutrons in the future.
References[1] ATLAS Collaboration, Phys. Lett. B 716, 1 (2012).
[2] CMS Collaboration, Phys. Lett. B 716, 30 (2012).
[3] J-PARC web site (available at https://j-parc.jp/c/en/index.html, date last accessed March 31,2020).
[4] M. Tanabashi et al. (Particle Data Group), Phys. Rev. D 98, 03001 (2018).
[5] K. Mishima et al., Nucl. Instrum. Meth. A 600, 342 (2009).
[6] R. Kitahara et al., Prog. Theor. Exp. Phys. 9, 093C01 (2019).
[7] C. C. Haddock et al., Phys. Rev. C 100, 064002, (2019)[8] C. C. Haddock et al., Phys. Rev. D 97, 062002, (2018)[9] N. Nagakura et al., Proceedings of Science, 191 (2016).
[10] R. Kossakowski et. al., Nuclear Physics A 503, 473 (1989).
[11] Y. Arimoto et al., Nucl. Instrum. Meth. A 799, 187 (2015).
[12] H. Otono, Nucl. Instrum. Meth. A 845, 278 (2017).
[13] K. Hirota et al., EPJ Web Conferences 231, 05002 (2020).

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6

Recent R&D of HTGR and Requirement for Nuclear Data
Y. Fukaya*, M. Goto, S. Nakagawa

Sector of Fast Reactor and Advanced Reactor Research and Development
Japan Atomic Energy Agency (JAEA)
4002 Narita-cho, Oarai-machi, Higashiibaraki-gun,Ibaraki 311-1393, Japan*E-mail: fukaya.yuji@jaea.go.jp
Recently, HTGR attracts a particular attention due to the outstanding safety features
especially after the Fukushima Daiichi nuclear disaster, and the R&D is significantly
promoted. In this presentation, we introduce the R&D plan of HTGR and the activities
related to reactor physics and nuclear data including an experiment by using KUCA.
Furthermore, requirement for nuclear data from the HTGR design is discussed.
1. Introduction
Japan Atomic Energy Agency (JAEA) has been performed the Research andDevelopment (R&D) of High Temperature Gas-cooled Reactor (HTGR) to utilize the hightemperature for multi-purpose heat utilization and electricity generation with high thermal
efficiency. Nuclear power generation including HTGR is an attractive energy source of cleanair and carbon-free electricity, that produces no greenhouse gases or air pollutants unlike
power generation with fossil fuel. Moreover, HTGR heat utilization of hydrogen generation
is expected to reduce the greenhouse gases by replacing the fossil fuel utilization.
Moreover, HTGR has outstanding safety features such as an excellent heat removal
function and inherent safety feature so called self-regulation due to the graphite core
structure. Graphite has high thermal conductivity and neutron moderation function which
realize safety feature of thermal reactor. After the Fukushima Daiichi nuclear disaster,
HTGR attracts a particular attention due to the outstanding safety features. The failure of
decay heat removal of the unit 1 of Fukushima Daiichi lead to the hydrogen explosion. On
the contrary, in HTGR, the decay heat can be removed successfully to final heat sink suchas air and soil in any circumstances even in depressurization accident where the coolant
flows out from the raptured pipe. The integrity of Coated Fuel Particle (CFP) remains inany situations. Therefore, the significant Fission Products (FPs) release is never occurredin HTGR.
Under the circumstances, recently, the R&D of HTGR is significantly promoted suchas a description in policy and international corporations. Furthermore, the R&D of reactor
physics technique with criticality experiment is also started. Those are reported in thisarticle.

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JAEA-Conf 2020-001

2. History of HTGR R&D and Future PlanFirst, JAEA had completed the design of High Temperature engineering Test Reactor(HTTR) with the demonstrate the elemental technologies and code validation. For fuelmaterial, in-pile helium loop named OGL-1 was built and operated. The loop was deployed
into the core of Japan Material Testing Reactor (JMTR), and the actual CFPs are loaded in
the similar condition in the HTGR core. For reactor physics, Very High Temperature Reactor
Critical Assembly (VHTRC) was built, and criticality experiment was performed. For
thermal hydraulics, Helium EngiNeering Demonstration Loop (HENDL) was built, andthermal hydraulics experiment was performed. Experimental correlation formulas relating
to Nusselt number and Reynolds number for the particular coolant flow path of HTTR were
determined, and flow distribution was measured. By using the validated code and theexperimental correlation formulas, HTTR was designed, licensed and constructed. The firstcriticality was achieved in 1998. In 2010, Loss Of Forced Cooling (LOFC) test was performed
as OECD/NEA project. Without scrum, the reactor power is safely reduced by Doppler effectand the decay heat is successfully removed without the forced cooling. The result attracts
the particular attention as the evidence of the outstanding safety features of HTGR after
the Fukushima Daiichi nuclear accident.
Next, JAEA had completed the design for the commercial use of high temperature gas
reactor by introducing the HTGR concept, called GTHTR300. It is planned to start theconstruction from 2030. The thermal output is 600MW, and the electricity output isapproximately 300MW. By employing direct gas-turbine system, the cheaper electricity costthan Light Water Reactor (LWR) can be achievable. By using the high temperature,hydrogen generation is also available.
3. Recent Promotion of R&D of HTGR
Due to the outstanding safety and advantage of hydrogen generation, the R&D wasrecommended in “Strategic Energy Plan”, which was formulated by the government ofJapan on April 11, 2014. In this document, the R&D of HTGR with international corporation
was recommended. As a result, the corporation with Poland was established. The Polishgovernment published “Strategy for responsible development” in 2017. In the policy,
preparation of HTGR development for industrial heat production in cogeneration wasplanned. The Polish ministry of energy accept the report “HTGR development strategy”
proposed by HTGR committee and published it in 2018. The development roadmap and thespecification of HTGR were clearly determined. The experimental reactor with output of
10MWt is planned to be built in 2025. The commercial reactor with output of 165MWt isplanned to be built in 2030. JAEA have also investigated the experimental reactor andcommercial reactor, now. In this study, we report the R&D relating to reactor physics.

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JAEA-Conf 2020-001

4. Recent R&D Activity Relating to Reactor Physics
The R&D have been performed by using Kyoto University Criticality Assembly(KUCA) experiment. There are two objectives for the R&D. First one is to introduce noise
analysis technique in HTTR experiment to determine subcriticality and neutron source
strength independently. This is necessary to elucidate the re-criticality of the LOFC
experiment of HTTR. It is very important for the commercial HTGR design because thetemperature increase by the tiny power burst due to the re-criticality exceeds a safety
limitation. The neutron source sustains the power level during the subcriticality, and thepower level determines the timing to occur the power burst and its height. However, theneutron source strength, in other words, number of Bq, is not equal to that of the pointkinetics, which is employed for safety analysis of licensing. To evaluate the source strengthof point-kinetics, the value should be weighted by forward and adjoint flux. The forward
flux cannot be evaluated by a criticality calculation, because the profile is exponential in a
subcriticality state. For the adjoint flux, there are several definitions in a subcriticality
state. Measurement of source strength is preferable from the viewpoint of accuracy.
Furthermore, by inverse kinetics, which is widely employed as reactivity meter, it cannot
be determined independently from subcriticality. Therefore, we are planning to develop
noise analysis method which can determine the subcriticality independently from neutron
source strength. With the relation between the measured power, neutron source, andsubcriticality in a steady state, the neutron source strength can be determined as well.
For first of a kind HTGR, propagation of the error of the nuclear data may have to beconsidered for safety analysis for licensing. Therefore, its reduction method may have to be
introduced. Representative one is bias method, which needs full-scale mock-up experiment.
It needs a huge budget. On the contrary, by using Generalized Bias Method (GBM), the error
propagation can be reduced by synthesizing the sensitivity coefficient to close to that of thetarget design. GBM needs the several criticality experiment data. Fortunately, the scaleeffect is not significant. The database can be established by small scale criticality
experiment.
For the two objectives, KUCA experiments was performed from 2018. The experiment
was planned to be performed for three years.
5. Criticality Experiment in KUCA
The experiment performed in 2018 is reported. The detail will be published in PYHSOR
2020. The core should realize the similar characteristics of HTGR. Here, HTTR is assumed
to be a representative one. In this experiment, the graphite-moderation core is composed insolid moderator system rack of KUCA so called “B-rack” to mimic the HTTR core. The
enrichment is mimicked by the combination of Highly Enrichment Uranium (HEU) plate- 29 -

JAEA-Conf 2020-001

which is composed of the alloy of 93 wt% HEU and aluminum, and Natural Uranium (NU)
metal plate. The average enrichment is 5.41 wt%. It is similar to the HTTR’s of 5.9 wt%.
The HEU plate and NU plate are assembled with moderator plate to realize the spectrumof HTGR. Graphite is also employed as the moderator for the core in B-rack. However, thevolume fraction of fuel material in HTGR is very small because the CFPs distribute in thegraphite structure. The Vm/Vf ratio is 76.8 in HTTR. To realize the soft spectrum of HTGR,
polyethylene moderator is necessary in addition to the graphite moderator because the sizeof the B-rack is small. Moreover, to achieve criticality with the large leakage core, driver
fuel assembly composed of HEU plate and polyethylene plate are deployed around the core.
The configuration of the fuel assemblies is shown in Fig.1, and the core configurationis shown in Fig.2. “F” is the fuel assembly composed of 8 unit cells, which include a 1/16”
thickness HEU plate, a 1 mm thickness NU plate, three 1/2” thickness graphite plates, a1/4” thickness graphite plate, and two 1/8” thickness polyethylene plates, with axial
graphite reflectors. The fuel plates were designed to realize the averaged fuel enrichmentof HTTR as described above. The polyethylene plate was used to mimic the HTTR spectrum.
“D” stands for the driver fuel assemblies composed of 38 unit cells, which include a 1/16”
thickness enriched uranium plate, and, three 1/8” thickness of polyethylene, with axial
graphite reflector. “d” stands for the partial length driver fuel to adjust criticality.
The BF3 detectors were deployed to observe reactor noise. By using the data, noise
analysis method based on Feynman-α method was developed. For the GBM, the criticality
experiment data can be regarded as representative core. The neutron spectrum similar to
HTTR’s can be realized and similar sensitivity coefficient is also expected.

1/16”EU + 1mmt NU + ½”G x 3 + ¼”G + 1/8”p x 2
Al2” Graphite x 10

½”G
EU NU
x1

½”G
x1

F
G

½”G ¼”G 1/8”x 1 x 1 PE
x2

2” Graphite x 10¼”G
x1

Fuel assembly (7/8”G2/8p8EUNU)
1/16”EU + 1/8”p x 3
Al2” Graphite x 10IJ

8 times repetition

1/8” 1/8” 1/8”
EU PE PE PE
x1 x1 x12” Graphite x 10½”G
x1

38 times repetition

Driver fuel assembly (3/8”p38EU)
1/16”EU + 1/8”p x 3
Alx 11

1/8” 1/8” 1/8”
2” Graphite x 10 EU PE PE PE
x1 x1 x12” Graphite x 10½”G
x1

16 times repetition

Partial length driver fuel assembly (3/8”p16EU)

Fig.1

Al
x 11

G

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74"G14P8EUNU Fuel3/8"p38 Fueld3

3/8"p18EU Fueld4

3/8"p14EU FuelG

Graphite Reflector●

1"φBF3 detector

G

K

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L

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FC# 2

FC#1

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G d4 d3 d4

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AA G

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G

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2

3

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9

10 11 12

Configuration of fuel assemblies

Fig. 2- 30 -

13 14 15 16 17

UIC# 418 19 20 21 22 23

G

G

G

G
G

G

G

24 25 26 27 28

Configuration of coreC

Control rod

S

Safety rod

N

Neutron source

G

UIC#6

G

Polyethylene reflector

FC

Fission chamber

UIC

UIC detector

G

JAEA-Conf 2020-001

6. Nuclear Data for HTGR and Requirement
For HTGR design, the nuclear data of graphite, namely cross-section of carbon andthe neutron scattering law in graphite, are important. Table 1 shows the history of themodification of the capture cross-section of carbon. The first JENDL was published in 1977
and its capture cross-section of carbon is 3.40 mb. In 1989, JENDL-3 was published, andthe capture cross-section of carbon was modified to 3.53 mb. Before JEDNL-4.0 was
published, we reported our HTTR criticality analyses results to the nuclear data center of
JAEA. In 2010, JENDL-4.0 was published, and the capture cross-section of carbon was
modified to 3.85 mb based on

Table 1 Capture cross-section of carbon

the latest measurement data.

Japan

ENDF/B-VI was published in
1974, and its capture crosssection of carbon was 3.36 mb.
The latest ENDF is ENDF/BVII.0 and its value is same as
the ENDF/B-VI, so this value
was not modified for 36 years

U.S.A.

Capt. X-secof C-nat.
1974
77
78
82
89
90
92
94
2001
2
5

at least.

JENDL-1

3.40mbJENDL-2
JENDL-3
JENDL-3.1

3.53mbCapt. X-secof C-nat.
ENDF/B-IV

Capt. X-secof C-nat.

3.36mb

ENDF/B-V

JEF-1

ENDF/B-VI

3.36mb

JEF-2.2

JENDL-3.2
JENDL-3.3

JEFF-3.1

68
10Europe

ENDF/B-VII
JENDL-4.0

3.85mb

3.36mb

JEFF-3.1.1

3.36mbRef. http://wwwndc.jaea.go.jp/ndd/symposium/2010/program.html

Table 2 Multiplication factors evaluated by criticality calculation for KUCA experiment
JENDL-4.0

JEFF-3.2

ENDF/B-VII.0

Critical state 1

0.99835 ± 0.00018

1.00043 ± 0.00018

1.00100 ± 0.00018

Critical state 20.99808 ± 0.00018

1.00059 ± 0.00018

1.00103 ± 0.00018

Critical state 3

0.99828 ± 0.00018

1.00069 ± 0.00018

1.00097 ± 0.00018

Critical state 40.99804 ± 0.00018

1.00048 ± 0.00018

1.00094 ± 0.00018

Critical state 50.99826 ± 0.00018

1.00046 ± 0.00018

1.00088 ± 0.00018

The criticality calculations for the KUCA experiment is performed by MVP code, which
is neutron transport calculation code based on Monte Carlo method with recent nuclear data
JENDL-4.0, JEFF-3.2, and ENDF/B-VII.0. The multiplication factors are listed in table 2.
The calculations are performed for each criticality condition to measure each control rod
worth. The result should coincide with unity. The difference is supposed due to the capture
cross-section of carbon, which is 3.85, 3.81, and 3.36 mbarn, respectively. Therefore,
calculations based on JENDL-4.0 in which only the carbon cross-sections are changed tothat of each library is performed, and similar tendency was confirmed. It comes from thedifference of the cross-section. The result of JEFF-3.2 shows a good agreement with thecriticality experiment. The cross-section of C-12 is 3.86 mbarn, but that of C-nat. is reduced- 31 -

JAEA-Conf 2020-001

to 3.81 mbarn with considering the abundance of C-13. In addition, the cross-section of
ENDF/B-7.1 is also increased to 3.86 mbarn. The result of JENDL-4.0 shows slightly lower
criticality.
Recently, neutron scattering law data for graphite is revised in ENDF/B-8.0. Thechange is to consider the porosity of the graphite. By this change, the criticality may beslightly changed. With this change, the cross-section may have to be changed. In addition,neutron scattering law data for SiC, which is also employed in recent HTGR design due tothe high oxidation durability, was also evaluated. These data are expected to be added to
JENDL-5.0. We are planning to validate the data for the HTGR design.
7. Summary
In this report, the history, recent, and future plan of R&D of HTGR are described, andthe requirement for nuclear data is introduced. The major points are as follows:

The LOFC experiment demonstrated the outstanding safety feature of HTGR, and it
attracted a particular attention especially after the Fukushima Daiichi Nuclear
Accident.



The R&D of HTGR with international cooperation have been recommended by the“Strategic Energy Plan” formulated by the government of Japan due to the attention tothe outstanding safety feature and hydrogen generation.



The R&D corporation with Poland was established. The plan to build commercial HTGR
is clearly described in the report.

To elucidate re-criticality in LOFC experiment is important for the commercial HTGR
design. And, GBM may be necessary to avoid full scale mock-up experiment, which may
be requested from the viewpoint of licensing of first of a kind commercial HTGR.

The R&D of noise analysis needed by the LOFC experiment and that of GBM started
by using KUCA in 2018.



The neutron scattering law of graphite should be improved by using ENDF/B-8.0 withconsidering porosity. With this revision, the capture cross-section may have to be
reduced. And, we are planning to validate the data for HTGR design.
With high expectations, recently R&D of HTGR are rapidly promoted. We should

promote the R&D related to reactor physics and nuclear data as well.
Acknowledgement
Authors appropriate to Dr. T. Sano, Dr. A. Sakon, and the Prof. K. Hashimoto of Kindai
Univ., who corporate the KUCA experiment, and also to operators of HTTR who corporates
the R&D of noise analysis system. This project is founded by the president of JAEA.

- 32 -

JAEA-Conf 2020-001

7

Performance of a gas-cooled reactor as a tritium production device
for fusion reactors
Hideaki MATSUURA
Applied Quantum Physics and Nuclear Engineering, Kyushu University
744 Motooka, Nishi-ku, Fukuoka 819-0395, Japane-mail: matsuura@kyudai.jp

To start up an initial fusion reactor and for technical tests for tritium circulation and blanket system, it isnecessary to provide sufficient amount of tritium from an outside device. At this point, how to supply asufficient amount of tritium for a future fusion DEMO and commercial reactors has not been clarified. We
have proposed tritium production using a high temperature gas-cooled reactor (HTGR). In this paper, we
introduce a HTGR system with a tritium production function for fusion reactors. Advantages of HTGRs fortritium production, performance of the typical HTGRs for tritium production, and current issues fordevelopment are discussed.

1. Introduction
To start up an initial fusion reactor and for technical test for tritium circulation and blanket system, it isnecessary to provide a sufficient amount of tritium from an outside device. A required condition necessary
for achieving self-sufficiency of deuterium-tritium (DT) fuel in a fusion reactor was investigated [1], and theuncertainties caused by nuclear data, calculation method, and system definition were revealed. Asaoka et al.
[2] evaluated the tritium breeding ratio and the initial inventory required for the initial fusion power reactor
with 2.28 GW thermal (~1GW electric) power as 1.10 and 27.6 kg respectively. In recent evaluations,amounts of tritium to be prepared for fusion reactor startup differ from 100 g to more than 10 kg [3-7]. The
uncertainties of the amount of tritium-supply scenario still seem to be increasing. So far, the tritium for fusion
reactors has been produced by using the D(n,γ)T reaction in the Canadian Deuterium Uranium (CANDU)
reactors [8]. Since the cross section of the D(n,γ)T reaction has almost 6 orders smaller around the thermalneutron energy range compared with the 6Li(n,α)T reaction (see Fig. 1), the tritium production efficiency per
year and per 1 GW electricity of CANDU reactor is smaller compared with other thermal fission reactors fortritium production using 6Li(n,α)T reaction, i.e., 0.17-0.23 T kg/year/GWe. In Fig. 2 an annual change of theelectricity produced and estimated tritium production by CANDU reactors in Canada, Korea, and Romania
are plotted [9]. It is shown that these reactors have an ability of 2.2-3.0 kg/year tritium production in 2018.
However, it is not clear whether theses reactors can provide sufficient amount of tritium to fusion DEMO
reactors after 20-30 years, because most of the reactors have already been operated over 30 years. To continue
the development of fusion reactors, it is required a sure scenario of tritium supply at this point. We haveproposed a tritium production method using high-temperature gas-cooled reactor [10,11].

- 33 -

JAEA-Conf 2020-001

Fig. 1 Cross sections of Li(n,α)T andD(n,γ)T reactions (from JENDL-4.0).
6

2. Advantage of HTGR for tritium production

Fig. 2 Tritium and electricity production by
CANDU reactors. Yellow (blue) bars
represent the maximum (minimum)
estimation of tritium producedin each year.

We consider the tritium production using the HTGR by inserting a Li compound as a burnable poison(BP) instead of a boron compound, i.e., an isotope as a neutron absorber is changed from 10B to 6Li. We
intend to produce the tritium utilizing the nuclear reaction of 6Li(n,α)T. The main core structure in the HTGR
is graphite, which is chemically stable and does not react with the Li compound. Although the large core sizeof the HTGR is not attractive from the viewpoint of economy, it provides enough space for loading the Licompound without 6Li enrichment, along with structural materials to prevent the leakage of tritium. In the
HTGR, the BP is usually used in a solid state (i.e., as B4C), and thus the Li compound can be loaded into thereactor’s core without significantly changing the original structural design. The nuclear characteristics andfuel temperature were analyzed to confirm the nuclear and thermal feasibility of a lithium-loaded HTGR [12],and it was shown that the analysis results satisfied the design requirements and hence the nuclear and thermal
feasibility was confirmed for a lithium-loaded HTGR that produces thermal energy and tritium [12].
Figure 3 presents the horizontal cross section of a gas turbine high-temperature reactor of 300 MWe
nominal capacity (GTHTR300) core [13]. The core consists of 90 fuel columns arranged in an annular ring;
55 and 36 inner and outer removable reflector columns, respectively; 18 and 12 inner and outer control-rod
guiding columns, respectively; and 8 outer fixed reflector sectors. Each fuel column is composed of 8hexagonal graphite blocks, which are piled up in the vertical direction. The hexagonal fuel block is 407 mm
wide across the flats (including 1-mm gaps at both sides) and 1000 mm high, with arrays of 57 fuel channels
and 3 burnable poison (BP) insertion holes (Fig. 4). The fuel rod, which is placed in the fuel channel enclosed
in a graphite block, is composed of fuel compacts and a graphite sleeve 26 mm in diameter. The diameter of
each fuel channel in the fuel block is 39 mm, and helium gas flows in the space between walls of the fuelchannel and the fuel rod as a coolant.
The horizontal cross section of the HTTR core is also shown in Fig. 5 (a) [14]. The core consists of 2
regions, i.e., actual core and reflector regions. The actual core consists of 30 fuel and 7 control-rod guidecolumns, each of which is composed of a stack of 5 fuel and 4 reflector blocks. The actual core is surrounded- 34 -

JAEA-Conf 2020-001

by the reflector region, i.e., replaceable and permanent reflector blocks and 9 control-rod guide blocks. The
core is 2.9 m in height and 2.3 m in diameter. Each hexagonal block is 360 mm wide across the flats and 580
mm high [see Fig. 5 (b)]. The fuel block contains 31 or 33 fuel channels and 3 BP insertion holes. 480 mm
high Li rods with 15 mm Al2O3 caps on both upper and bottom sides are loaded into the BP insertion holes.

Fig. 3 Horizontal cross-section of the GTHTR300
core.

Fig. 4 Fuel block with BP insertion holes
and BP rod for GTHTR300.

Fig. 5 (a) Horizontal cross-section of the HTTR core,(b) Fuel block with Li-rod insertion holes.

3. Performance of HTGR for tritium production
The estimation of the amount of tritium produced and the effective multiplication factor during the
GTHTR300 and HTTR 360-day operations was carried out using the continuous-energy Monte Carlotransport code MVP-BURN [15] using the whole core model. Throughout the calculations, nuclear data weretaken from JENDL-3.3 [16] and JENDL-4.0 [17], and the results for each nuclear cross-section dataset arepresented and compared in Figs. 6 and 7. The fuel region temperature was 1,100 °C and the Li-rod
temperature was the same as the moderator (900 °C). In this study we assumed 360-day reactor operation
period. All control rods were assumed to be pulled out. The time steps in the burn-up simulations were taken
as 0, 1, 5, 30, 60, 120, 180 and 360 days. For each of the time steps, 6,000,000 neutrons were generated. The
statistical errors of the effective multiplication factor and the reaction rate were less than 0.1% in all
calculations, which is sufficient accuracy in our discussion.

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JAEA-Conf 2020-001

In Figures 6 and 7, the effective multiplication factor keff keeps decreasing over the operation and reachedthe minimum value after the 360-day operation. On the contrary, the cumulative weight of the produced
tritium is increasing over the time and reached the maximum value after the 360-day operation. In thesesimulations, as a typical value, a rod with a 44 mm diameter, a 6.4 (6.3) mm thick Al2O3 layer, a 1 mm thick Zr
layer, and a 2.4 (2.3) mm thick LiAlO2 layer are assumed when JENDL-4.0 (3.3) is used for GTHTR300. For
HTTR, a rod with a 14 mm diameter, a 1.9 mm thick Al2O3 layer, a 0.1 mm thick Zr layer, and a 2.4 (2.7) mm
thick LiAlO2 layer are assumed when JENDL-4.0 (3.3) is assumed. In order to operate the reactor during 360 days,
at least 1.02 of keff is required at the end of the operation. From Fig. 6 we can estimate that the possible weight oftritium produced in the GTHTR300 is roughly 800 g during 360-day operation. From Fig. 7 we can estimate thatthe possible weight of tritium produced in the HTTR is roughly 30 g during 360-day operation.
In JENDL-4.0 the neutron capture cross section of carbon is evaluated almost 10% larger compared withprevious other cross section sets. It is well known that by using the JENDL-4.0 the effective multiplication factor
in criticality is evaluated in smaller value, and the simulations become consistent with experimental observationin HTGRs [18]. When we use the JENDL-4.0 instead of 3.3, the effective multiplication factor is evaluated smaller,
thus the amount of tritium produced in unit time is also decreased. As was explained before, the weight of Li
loading into the core is adjusted so that the effective multiplication factor is the same value at the end of theoperation (shown in Figs. 6 and 7). From the viewgraphs, we can find that the amount of tritium produced decreases
by about several percentages when we adopted the JENDL-4.0.

Fig. 6 The effective multiplication factor andthe amount of tritium produced for
GTHTH300.

Fig. 7 The effective multiplication factor andthe amount of tritium produced for
HTTR.

4. Status of the development and current issues
During the operation, the amount of tritium flowing out from the Li rod into helium gas should be
suppressed to a lower level as much as possible. Numerical simulations have predicted that if we could
operate the HTGR in a low temperature range, keeping the rod temperature below 500 °C, the tritium leaking
from the Li rod can be suppressed to almost 1% of the amount produced [19]. However, if we intended tooperate the HTGR in a much higher temperature range (i.e., the rod temperature reaching 800-900 °C) so as
to increase the electricity generation efficiency, the leakage of the tritium would rapidly increase. This is- 36 -

JAEA-Conf 2020-001

because the hydrogen permeability of the Al2O3 layer in the Li rod increases with the increase of the rod’s
temperature [20]. In order to reduce the leakage of tritium from the rod, we attempted to adopt Zr in the rod
as a tritium absorption material to avoid increasing the inner tritium partial pressure [21-23]. To prevent thetritium absorption capability from decreasing due to oxidization via interacting with coexisting oxides (i.e.,
LiAlO2 and Al2O3), some kind of coating with an antioxidant material is necessary [21]; it is desirable tohave a surface area of Zr is as large as possible. We propose the Li rod including Zr pebbles with Ni coating
as shown in Fig. 8 [24].
We have experimentally measured the hydrogen absorption performance of Zr (i.e., hydrogen absorption
speed) [24]. The performance is governed by concentration of hydrogen in Zr, surface area, H/Zr ratio,pressure and temperature of the circumstance [25, 26], in addition to the surface status, e.g., oxidization via
interacting with coexisting oxides. On the basis of the measured data, we numerically simulated the tritium
leakage from the Li rod during the reactor’s operation and the correlation with the rod structure, assuming
several patterns of the tritium absorption capability. The optimal size of each layer and the diameter of thepebble (e.g., x, y, and z in Fig. 8) should be clarified. At this point, we consider that a rod with a 14 mm
diameter, a 1.9 mm thick Al2O3 layer, a 0.1 mm thick Zr layer, and a 2.7 mm thick LiAlO2 layer, involving
Zr pebbles with Ni coating and a 0.5 mm radius, is one of the most desirable structures of a Li-rod for an
HTTR [24]. By using the presented rod, we estimate approximately 30 g of tritium can be produced in 360day operation in HTTR, and the tritium flowout into He is less than 1% (when tritium absorption time issmaller than 1 hour [24]).

Fig. 8 A schematic view of the Li-loading rod
including Zr with Ni coating for HTTR.

5. Summary
To confirm the optimal Li-loading structure and to produce the tritium for fusion reactor stably, it isnecessary to ascertain the performance of the Li-loading rod presented. We are now planning an irradiation
test of the tritium production and containment. In order to complete the final design, the amount of tritium
produced and leaked (i.e., the analysis model and the effective hydrogen absorption time of Zr) should bevalidated in a demonstration test in an HTTR.

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JAEA-Conf 2020-001

Acknowledgement
This work is supported by JSPS KAKENHI Grant-in-Aid for Scientific Research (B) JP18H1200 andcooperative program with BA (The Joint Implementation of the Broader Approach Activities in the Field ofFusion Energy Research) reactor design team in Japan. The author acknowledges the collaborators, Dr. K.
Katayama, Dr. T. Otsuka, Dr. M. Goto, S. Nakagawa, Dr. E. Ishitsuka, Dr. S. Hamamoto, Dr. K. Tobita, Dr.
R. Hiwatari, Dr. Y. Someya, Dr. Y. Sakamoto, and students Y. Koga, T. Suganuma, and K. Nakagwa for many
useful discussions.

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[22] S. Nagasumi, H. Matsuura, K. Katayama, et al., Fusion Sci. Tech., 72 (2017) 753.
[23] Y. Koga, H. Matsuura, Y. Ida, et al., Fusion Eng. Des., 136 (2018) 587.
[24] H. Matsuura, et al., Fusion Eng. Des., 146 (2019) 1077.
[25] E. Zuzek, et al., Bull. Alloy Phase Diag., 11 (1990) 385.
[26] S. Yamanaka, et al., J. Nucl. Mater., 167 (1989) 231.

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8Research and development of an innovative transmutation system
of LLFP by fast reactors
Satoshi CHIBATokyo Institute of Technology
e-mail: chiba.s.ae@m.titech.ac.jp

A project on transmutation of LLFPs, supported by MEXT, is described. In this project,
6 LLFPs, namely, 79Se, 93Ze, 99Tc, 107Pd, 129I and 135Cs, were selected as target materials.
We combine materials including these LLFPs with novel moderators YH2 and YD2 tosoften neutron spectra of fast reactors to make transmutation rate larger.

The ultimate

goal is to make the support factor, which is defined as a ratio of amount of a specific
LLFP transmuted and amount of the same LLFP populated in the core of fast reactors,
to be larger than 1 so that we can actually reduce amount of LLFPs.

We have shownthat such a transmutation scheme is possible without performing isotopic separation nor
even-odd separation of LLFPs.
1. Introduction
Disposal and proper treatment of radioactive wastes generated by nuclear power
plants are considered to be important national problems, which is directly connected toacceptance of nuclear technology from the society.

Partitioning and transmutation

scenario is an efficient way to achieve reduction of radioactive wastes.
is the major target of such a scheme.

Minor actinide

However, if we perform transmutation of MAs,

they are destroyed by the fission reaction and hence enhancement of long-live fission
products (LLFPs).

We recognize that this problem reduces to the problem how to treat

the LLFPs. On the other hand, LLFPs are the major source of dose level which generalpublic may be exposed from deep geological repository site after several hundred
thousands years, which makes construction of such site extremely difficult.

Therefore,

we aim to design a transmutation scheme for the LLFPs. Among 7 LLFPs, we try to
propose a scheme to transmute 6 LLFPs, namely, 79Se, 93Ze, 99Tc, 107Pd, 129I and 135Cs by
fast reactors.

126Sn was excluded since its neutron capture cross section is estimated tobe very small, even though its uncertainty is quite large (there is no data).
At the beginning, we aimed to utilize MONJU experimental reactor for thistransmutation.

However, after 1 year of research, it was decided to shut down MONJU

by the government, so we switched to utilize a small fast reactor equivalent to MONJU.

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JAEA-Conf 2020-001

2. Outline of the research
We performed research in conceptual design of LLFP transmutation system which
comprised reactor physics analysis, selection of target materials, thermal hydraulics,
safety, retrieval and reprocessing of LLFPs, irradiation scheme, nuclear data, planningof irradiation test, feasibility of transmutation of LLFP by large FBR, and social
scientific aspects of transmutation technology. These results were already published inRefs. [1-6] as well as press releases [7-10] and patent application [11], so we do notdescribe them in detail here.

The basic concept is to irradiate materials including thetarget LLFPs in the blanket region of MONJU-like small fast reactors.

The basic

concept is shown in Fig. 1.

Fig. 1 Basic concept of LLFP transmutation considered in the present project.
We made several trial and error in analyses, then reached to the following 4 concepts:
1. A reactor containing 6 LLFPs simultaneously in the radial and axial blanket andeven in some structural part of the reactor (Fig. 2)[3]2. A system of 3 reactors, each transmuting 2 LLFPs (Fig. 3)
3. A system of 1 reactor containing 4 LLFPs 9Se, 99Tc, 107Pd and 129I
4. A system of 2 reactors, 1 containing 3 LLFPs 79Se, 99Tc and 129I, and 1 used forbreeding of 239Pu
These systems can be selected according to the needs of the society.

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JAEA-Conf 2020-001

Gas Plenum

Inner Core

Outer
Core

Fig. 2 A system for simultaneous transmutation of 6 LLFPs [3]Fig. 3 A system of combination of 3 fast reactors, each transmuting different LLFPs- 41 -

JAEA-Conf
JAEA-Conf 2020-001

At the same time, researches on nuclear data have been performed, and the results
were published as Ref. [2] and [5-6].

Especially, new fission yield data was generated,

and it will be included in the next version of JENDL FPY data library.
We have found that the transmutation ratios of LLFPs in these systems haverather large uncertainty reflecting the fact that all of them are radioactive nuclei (sonuclear data is not accurate enough).

It is high desirable to quantify uncertainty of

nuclear data of these isotopes.
3. Summary
A project on transmutation of LLFPs by MONJU-like small fast reactors is
described.

The details can be found in the given references, and more papers are goingto appear.
Acknowledgement
This study comprises the results of “ Research and development of an innovativetransmutation system of LLFP by fast reactors”entrusted to the Tokyo Institute ofTechnology by the Ministry of Education, Culture, Sports, Science and Technology ofJapan (MEXT).
References[1] S.Chiba, T.Wawkabayashi, Y.Tachi, N.Takaki, A.Terashima, S.Okumura andT.Yoshida, “Method to Reduce Long-lived Fission Products by Nuclear Transmutations
with Fast Spectrum Reactors”, Scientific Reports, 7, 13961(2017).
[2] M.D.Usang, F.A.Ivanyuk, C.Ishizuka and S.Chiba, “Correlated transitions in TKE
and mass distributions of fission fragments described by 4-D Langevin equation”,
Scientific Reports, 9, 1525(2019).
[3] T.Wakabayashi, M.Takahashi, S.Chiba, N.Takaki, Y.Tachi and Y.Tahara, “Coreconcept of simultaneous transmutation of six LLFP nuclides using a fast reactor”,
Nucl. Eng. Design, 352, 110208(2019).
[4] T.Wakabayashi, Y.Tachi, M.Takahashi, S.Chiba and N.Takaki, “Study on method toachieve high transmutation of LLFP using fast reactor”, Scientific Reports, 9,19156(2019).
[5] C.Ishizuka, X.Zhang, M.D.Usang, F.A. Ivanyuk and S.Chiba, “The effect of the doubly
magic shell closures in 132Sn and 208Pb on the fission fragments mass distributionsof superheavy nuclei”, Phys. Rev. C, 101, 011601(R)(2020).
[6] K.Tsubakihara, S.Okumura, C.Ishizuka, T.Yoshida, F.Minato and S.Chiba,- 42 -

JAEA-Conf
JAEA-Conf 2020-001“Systematic Evaluation of Independent Yields Based on Nuclear Shell Correction
and Boltzmann Weight” (to be submitted to J.N.S.T.).
[7] Tokyo Tech. News, https://www.titech.ac.jp/news/2017/039840.html (accessed 2020/03/24).
[8] Tokyo Tech. News, https://www.titech.ac.jp/news/2019/043509.html (accessed 2020/03/24).
[9] Tokyo Tech. News, https://www.titech.ac.jp/news/2020/046068.html (accessed 2020/03/24).
[10] Tokyo Tech. News, https://www.titech.ac.jp/news/2020/046173.html (accessed 2020/03/24).
[11] Patent application 2020-046979, March 17, 2020 (Tokyo City University).

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9 Nuclear Data Activities and Related Database Services
at the IAEA Nuclear Data Section
Shin OKUMURA1
1International Atomic Energy Agency, 1400 Vienna, Austria

e-mail:s.okumura@iaea.org

The Nuclear Data Section (NDS) of IAEA conducted the completeness assessment of fission product yield data
in EXFOR database against two lists of articles that were used in ENDF/B-VI and UKFY3.0 library evaluations.
We found that 194 articles were found to be relevant for new entry. The IAEA-NDS also launched a new online
application, the Medical Isotope Browser, which allows professional users to calculate and predict the medical
isotope production yield of any diagnostic, therapeutic or theranostic isotope with accelerators. We here review
above two main activities.

1

Introduction

The Nuclear Data Section (NDS) of IAEA assembles, develops and disseminates nuclear data and related
databases for basic science and nuclear applications. The IAEA-NDS develops nuclear data, nuclear structureand decay data libraries, disseminates them, and maintains databases, as well as organize the various meetings to
produce outputs by encouraging collaboration worldwide.
The fission product yield (FPY) is important for many application fields because FPY determines the compositions of radioactive nuclear wastes, affect the safety-operation of nuclear power plants (because of delayed neutron
emissions), design accelerator-driven systems (ADS) systems for transmutations of minor actinides, and so on [1].
The IAEA-NDS has extensively reviewed the user needs in these fields in the previous IAEA Coordinated Research
Projects (CRPs) and meetings [1, 2, 3, 4].
In recent years, several theories and models have been developed to study nuclear fission phenomenon, such as
microscopic theories [5, 6] and dynamical models [7]. Despite many theories and models have been developed,
it is still a challenging problem to understand nuclear fission phenomenon, because nuclear fission is a very
complex process that incorporates large-amplitude collective motion of nucleons. Besides theoretical and modelling
approaches, the experimental observation of FPY is also still a challenging subject while various experimental fission
techniques have been developed [8]. The current evaluated libraries, i.e. later than ENDF/B-VI [9], JENDL/FPY2011 [10], and JEFF3.3 FPY library [11], comprise from the evaluation of the experimental data as well as some
empirical models such as Z P model from Wahl systematics [12]. Such evaluation works have been done about
two decades ago and only a few additions have been applied [13]. Recent increase in interests on the FPY frommany fields stresses the importance to develop the new evaluated FPY data libraries [3]. Under this circumstance,
the IAEA-NDS will launch the new CRP in 2020 on “Updating Fission Product Yield Data for Applications” forindependent and cumulative FPY important actinides. Prior to start the CRP, the IAEA-NDS investigated thecompleteness of FPY data in the experimental nuclear reaction (EXFOR) database [14, 15] against FPY datasetsthat were used in the evaluation of in ENDF-B/VI [16] and UKFY3.0 [17].

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The IAEA-NDS also launched a new online application, the Medical Isotope Browser [18], which brings data
directly to professionals in radiopharmaceutical research and industry. In this report, we review the result ofcompleteness investigation of FPY data in EXFOR and the brief instruction of the Medical Isotope Browser.

2

FPY data assessment

The EXFOR database was established in 1960s originally exclusively neutron-induced reaction data and contains
data for more than 22,000 experiments, resulting in more than 13,000,000 data points, and continue the effort tocompile new and overlooked data, and to remove errors. Nevertheless, some of the FPY data have not beenaccessible due to the classification or dissertation. Prior to start new CRP on FPY evaluation, the IAEA-NDS
conducted the completeness investigation of FPY data in EXFOR database to compare with the articles cited in the
ENDF-B/VI [16] and UKFY3.0 [17] reports. The detail of the assessment procedure is described in Refs. [19, 20].
The complete list of articles reviewed in this assessment are listed in the Appendix of Ref. [21].
The ENDF/B-VI report cites 1,602 references, of which 924 references provide FPY data adopted, while the rest
678 references are not used for the evaluation because of superseded later publication. The UKFY3.0 evaluation
summary cites 646 references. About 600 references are overlapped in both ENDF/B-VI and UKFY3.0 reports
and about 1,200 references are unique. Table 1 shows the summary of this investigation. We reviewed all articles
that possibly have experimental FPY data that is relevant to compile into EXFOR database. The articles were thencategorized into (a) Already compiled, (b) not found in EXFOR and need to be compiled, (c) need to check further
details, or (d) not relevant. The world wide collaboration, NRDC network, has started to compile (a) cases.
Table 1: The number of FPY references categorised during our assessment for those cited in evaluation summaries
of ENDF/B-VI [16] and UKFY3.0 [17]. See Refs. [19, 20, 21] for details.
SourceExista Create new entryb Checkc Not relevantd Total
ENDF/B-VI
UKFY3.0

618
409

154
139

36
32

116
64

924
645

Totale

720

194

54

167

1203

a Already compiled in EXFOR.
b Not found in EXFOR and need to create new entry.
c Need to check if all relevant data are compiled in EXFOR entry.
d Not FPY or not relevant for EXFOR or the original article cannot identify.
e Total after elimination of overlap.

Figure 1 shows the number of FPY entries that have been stored in EXFOR database with the publication year
on the horizontal axis. The number of FPY entries in EXFOR is supposed to reflect the number of publications in ayear. Based on this data, the first FPY measurement data was published in 1943. Until 1960s, FPY measurements
are mostly dominated by the neutron induced fission of important actinides as targets, namely Th, U, Pu, and Cf
isotopes. The experimental FPY data drastically increase in 1970s. After 1980s, variations of target isotopes, e.g.
Np, Am or Cm, have increased while the majority of trend remains. Up to now, the majority of reactions are theneutron induced fission followed by the spontaneous fission, photon, and proton induced fission.
By comparing the number of FPY data in Fig. 1 extracted from EXFOR database on July 2019 and on December
2019, the number of entries has increased in wide range of published year as a result of compilation efforts of
NRDC network.

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JAEA-Conf 2020-001

Figure 1: Number of FPY related entries extracted from EXFOR database on July 2019 (before starting compilations) and December 2019 (some FPY data from this assessment have been compiled.)

3

Medical Isotope Browser

The IAEA-NDS launch a new web application product for the prediction of medical isotope production yield of
any diagnostic, therapeutic or theranostic isotope with accelerators. The isotope production yield can be calculated
for any isotope production route for many targets, natural or enriched, reacting with incident protons, deuterons,
tritons, helions or alpha particles. IAEA has evaluated cross sections in a series of IAEA research projects thatstarted in 2005. The cross sections from IAEA-evaluated medical isotope data library [22, 23] for about 150
reaction channels are used for the calculation. For reactions or extended energy ranges (up to 200 MeV) not coveredby IAEA-evaluated cross sections, TENDL-2017 library [24] are used. User can also use own cross section data as
input. The radioactive decay from isotope to any other isotope is calculated based on the JEFF-3.3-RD decay datalibrary [11].
The Medical Isotope Browser can be reached via IAEA-NDS web site (URL https://nds.iaea.org/mib).
The essential fields in the form are Product, Projectile, Target, Thickness, Incident energy, and Irradiation time.
On the left panel of Figure 2 shows the sample inputs for medical isotope production yield of 44 Sc from 44 Ca+preaction with 20 MeV incident energy and 1 eµA current for 1 day irradiation followed by 12 hour elapsed timeafter irradiation. The cross section used in the calculation can be plotted compared with available experimental
data as shown in the right panel of Fig. 2. As above example shows, the Medical Isotope Browser gives a first
guess to users who aim to find a new production route, an optimal incident energy, and so on. The IAEA-NDS
acknowledges for any feedback to make the product better.

4

Conclusion

The IAEA-NDS investigated the completeness of the FPY data in the EXFOR database by crosschecking
against the articles listed in the ENDF-B/VI and UKFY3.0 evaluation reports. It was found that approximately 194
references are relevant to create new entry in EXFOR database, and about 54 references must be checked with theexisting EXFOR entries to ensure that the all experimental FPY data are properly compiled. The compilation has
been started by data centers of NRDC network.

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JAEA-Conf 2020-001

Figure 2: Screenshots of Medical Isotope Browser. (Left) Sample isotope yield calculation of 44 Sc from 44 Ca+ p at 20 MeV incident energy. (Right) Cross section view used in the calculation showing together with someexperimental data.
The new web application, the Medical Isotope Browser, has been released. The isotope production yield canbe calculated for any isotope production route for many targets, natural or enriched, reacting with incident protons,
deuterons, tritons, helions or alpha particles.

References[1] International Atomic Energy Agency. CRP on Compilation and evaluation of fission yield nuclear
data (1991-1996). International Atomic Energy Agency; 2000. IAEA-TECDOC-1168.
[2] CRP on Fission Product Yield Data for the Transmutation of Minor Actinide Nuclear Waste(1997-2002). International Atomic Energy Agency; 2008. STI/PUB/1286.
[3] Dimitriou P, Hambsch FJ, Pomp S (eds.). Fission Yields: current status and perspective in measurements,
theory and evaluations. International Atomic Energy Agency; 2016. INDC(NDS)-0713.
[4] Minato F, Okumura S, Koning A, Kawano T. Implementation of the Hauser-Feshbach theory for
Fission Product Yield Evaluation and Fission Modelling. International Atomic Energy Agency; 2019.
IAEA-NDS-230. https://www-nds.iaea.org/publications/nds/iaea-nds-0230/ (accessed 2020/3/24).
[5] Bertsch GF, Loveland W, Nazarewicz W, Talou P.
may;42(7):077001.

Benchmarking nuclear fission theory. 2015[6] Schunck N, Robledo LM. Microscopic theory of nuclear fission: a review. Reports on Progress in Physics.
2016 oct;79(11):116301.
[7] Ishizuka C, Usang MD, Ivanyuk FA, Maruhn JA, Nishio K, Chiba S. Four-dimensional Langevin approach
to low-energy nuclear fission of 236U. Phys Rev C. 2017;96:064616.
[8] Andreyev AN, Nishio K, Schmidt KH. Nuclear fission: a review of experimental advances andphenomenol-ogy. Reports on Progress in Physics. 2017 nov;81(1):016301.

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JAEA-Conf 2020-001[9] Brown DA, Chadwick MB, Capote R, Kahler AC, Trkov A, Herman MW, et al. ENDF/B-VIII.0: The 8th
Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards andThermal Scattering Data. Nuclear Data Sheets. 2018;148:1 – 142. Special Issue on Nuclear Reaction Data.
[10] Katakura J. JENDL FP Decay Data File 2011 and Fission Yields Data File 2011. Japan Atomic Energy
Agency; 2012. JAEA-Data/Code 2011-025.
[11] Kellett MA, Bersillon O, Mills RW. The JEFF-3.1/-3.1.1 radioactive decay data and fission yields sub-libraries.
OECD NEA; 2009. JEFF Report 20 OECD NEA.
[12] Wahl AC. Systematics of Fission-Product Yields. Los Alamos National Laboratory; 2002. LA-13928.
[13] Kawano T, Chadwick MB. Estimation of 239Pu independent and cumulative fissio product yields fromthe chain yield data using a Bayesian technique. Journal of Nuclear Science and Technology.
2013;50(10):1034-1042.
[14] Otuka N, Dupont E, Semkova V, Pritychenko B, Blokhin AI, Aikawa M, et al. Towards a More Completeand Accurate Experimental Nuclear Reaction Data Library (EXFOR): International Collaboration Between
Nuclear Reaction Data Centres (NRDC). Nuclear Data Sheets. 2014;120:272 – 276.
[15] Zerkin VV, Pritychenko B. The experimental nuclear reaction data (EXFOR): Extended computer
database and Web retrieval system. Nuclear Instruments and Methods in Physics Research Section A:
Accelerators, Spectrometers, Detectors and Associated Equipment. 2018;888:31-43.
[16] England TR, Rider BF. Evaluation and compilation of fission product yields. Los Alamos National Laboratory;
1994. ENDF-349, LA-UR-94-3106.
[17] Mills RW. Fission product yield evaluation: Appendix 5. University of Birmingham; 1995.
[18] Medical Isotope Browser;. https://nds.iaea.org/mib (accessed 2020/3/24).
[19] Okumura S. Experimental fission product yields adopted in ENDF and UKFY evaluation but missingin EXFOR (A29). 2019;NRDC Meeting 2019.
[20] Okumura S. Experimental fission product yields missing in EXFOR. NRDC Memo CP-D/979. 2019.
[21] Fleming M, Kawano T, Otuka (eds ) N. Fission Product Yield Experimental Database. International
Atomic Energy Agency; 2019. INDC(NDS)-793.
[22] Tárkányi FT, Ignatyuk AV, Hermanne A, Capote R, Carlson BV, Engle JW, et al. Recommended nuclear data
for medical radioisotope production: diagnostic positron emitters. Journal of Radioanalytical andNuclear Chemistry. 2019;319(2):533–666.
[23] Engle JW, Ignatyuk AV, Capote R, Carlson BV, Hermanne A, Kellett MA, et al. Recommended Nuclear Datafor the Production of Selected Therapeutic Radionuclides. Nuclear Data Sheets. 2019;155:56 – 74. Special
Issue on Nuclear Reaction Data.
[24] Koning AJ, Rochman D, Sublet JC, Dzysiuk N, Fleming M, van der Marck S. TENDL: Complete Nuclear
Data Library for Innovative Nuclear Science and Technology. Nuclear Data Sheets. 2019;155:1 – 55. Special
Issue on Nuclear Reaction Data.

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10Status of the Nuclear Energy Agency Data Bank

– Services for Nuclear Data and Computer Programs
Kenya SUYAMA, Franco MICHEL-SENDIS, Alice DUFRESNE and Maria-Eleni RAGOUSSI
Nuclear Energy Agency (NEA) Data Bank
Organisation for Economic Cooperation and Development (OECD)2, rue André Pascal, 75775 Paris Cedex 16, France
e-mail: kenya.suyama@oecd-nea.org

The Nuclear Energy Agency (NEA) Data Bank is responsible for the coordination of the Joint Evaluated Fission
and Fusion (JEFF) Nuclear Data Library project, which is a collaboration between NEA Data Bank participating
countries. The JEFF library combines the efforts of its participating institutions to produce sets of evaluated nuclear
data, for fission and fusion applications. This evaluated nuclear data library is a reference for many European
countries, and in particular for the French nuclear industry. In addition to the Nuclear Data Services (NDS), theData Bank Computer Program Services (CPS) is in charge of providing direct services to end users by collecting
and disseminating computer programs in the Data Bank participating countries and organising training courses on
the most widely used codes. The Data Bank also supports other NEA standing technical committees (STCs). For
example, the Data Bank has been developing the Thermochemical Database (TDB) Project since 1984, following
recommendations by the NEA Radioactive Waste Management Committee (RWMC) concerning the need for a
database of high-quality thermochemical data for species relevant to performance assessments of deep geologicalrepositories.

1. Introduction
With the consolidation of the Computer Program Library (CPL) and the Neutron Data Compilation Centre(CCDN) – established in Ispla (Italy) and Saclay (France), respectively in 1964 as common service providers for
the European Nuclear Energy Agency (ENEA) – the Data Bank of the Nuclear Energy Agency was founded in
1978. Since then, the NEA Data Bank has been providing computer codes and nuclear data to end users of its
participating countries and a platform to exchange information in these technical fields.
It is widely known that the main activity of the Data bank is its: i) Computer Program Services (CPS);
and ii) Nuclear Data Service (NDS) to disseminate computer codes and nuclear data. Since the end of 1980s,
however, the Data Bank has been supporting the activities of NEA standing technical committees (STCs), which
has in fact become one of the most important activities of the Data Bank today. Indeed, the Data Bank is also
overseeing the controlled distribution of integral experimental data and databases, and of joint-project outcomes,
in addition to preserving all scientific content. Therefore, the activity and the role of the Data Bank in terms ofknowledge preservation and management has been increasing in importance and is now referred to as “Knowledge
Preservation and Management” (KPM).
This article provides end users of the Data Bank not only with an outline of current Data Bank activities,
i.e. NDS, CPS and KPM, but also with some strategic directions for the future.
2. Nuclear Data Services [1]
The Nuclear Data Services (NDS) of the Data Bank overlooks all nuclear data related activities carriedout within the framework of the Data Bank. These activities include: the co-ordination of the JEFF Nuclear Data
Library project; the compilation of experimental data (EXFOR) by agreement with the network of international
Nuclear Reaction Data Centres (NRDC); and, in recent years, the definition and implementation of verification- 51 -

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and benchmarking processes for nuclear data, under the scope of the nuclear data validation mission of the Data
Bank. A brief description of these activities is given below:
As co-ordinator and secretariat to the JEFF Nuclear Data Library Project, the NDS organises the “Nuclear
Data Week” twice a year with the aim of gathering experimentalists, evaluators and end users belonging to the
JEFF community, along with other representatives of Nuclear Data projects carried out under the NEA auspices(WPEC) or in other NEA member countries (in particular, the European Commission). This project is a mandated
activity of the Data Bank. The current mandate of the JEFF project is aimed at preparing and releasing the next
update of the library, JEFF-4, by 2024. This release would follow the latest JEFF-3.3 library release, dating fromthe end of 2017.
Another main activity of the NDS is the coordination of the Data Bank contribution to the Nuclear
Reaction Data Centres (NRDC) Network of the International Atomic Energy Agency (IAEA), which oversees thecompilation of experimental nuclear reaction data (EXFOR) database. The Data Bank is responsible for thecompilation of reaction data in the EXFOR database that has been obtained from Data Bank participating countries.
The NDS also carries out technical developments in the area of nuclear data visualisation, verificationand validation for end users of the nuclear data library, with the development of tools such as JANIS for thevisualisation of nuclear data and the implementation of verification and validation (V&V) sequences such as the
Nuclear Data Evaluation Cycle (NDEC) for the benchmarking and validation of nuclear data. This work also
provides an important infrastructure that is routinely used by the JEFF project.
3. Computer Program Services [2]
The Computer Program Services (CPS) collects computer codes, integral experimental data and otherrelated documents and disseminates them to end users. The CPS also verifies deposited computer programs toensure that they can easily be installed and used by those who will receive them. With more than 55 years ofexperience, the CPS distributes approximately 3 500 packages of computer code and integral experimental data
every year, with a careful screening of the requests to comply with the corresponding restrictions. This distribution
also includes the outcomes of joint projects conducted by NEA STCs.
The CPS acts as an important hub in the field of exchanging nuclear related computer codes around theworld. The Data Bank has an agreement, in particular, with the Radiation Safety Information Computational Centre(RSICC) for the exchange of computer programs between Data Bank participating countries and the RSICC
service area (i.e. in the United States and Canada). The Data Bank also exchanges computer codes with non-OECD
countries according to an arrangement with the International Atomic Energy Agency (IAEA). This service to nonOECD countries is carried out in compliance with the procedures set out by NEA Data Bank countries, whichdeveloped the requested code.
The CPS also organises approximately 10+ training courses every year on the most popular computercodes, where code developers are leading the course. In recent years, this activity has been recognised as key tosustaining the technical knowledge of nuclear technology with its vast contribution to the education of nuclear
professionals.
For the past two years, the CPS has been attempting to change the licensing framework for its computercodes catalogue so as to comply with requests from Data Bank participating countries and trends in the licensingof codes with applications in the nuclear science and technology domain. Currently, two main licence types coexist, the “single site” versus “single user” licences. The “single site” licence, which allows the designatedestablishment to share codes among all staff members, will disappear upon the introduction of the generic “single
user licence”, expected to enter into force during 2020.
4. Knowledge Preservation and Management and NEA standing technical committees
In addition to the CPS and NDS, the Data Bank has another important function, which is to collaborate
with and offer support to NEA STCs in the area of knowledge preservation and management (KPM) by supporting
the development and dissemination of databases.
For example, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) andinternational Reactor Physics Experiment Evaluation (IRPhE) Project are evaluated and revised under the working
parties of the NEA Nuclear Science Committee (NSC). It is the CPS, however, that receives and handles the- 52 -

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requests for these two projects. The Data Bank also supports the development and maintenance of databases suchas the Code Validation Matrix (CCVM), Information System on Occupational Exposure (ISOE), International
Fuel Performance Experiments (IFPE) and SINBAD, all of which are under the auspices of the NEA Committee
on the Safety of Nuclear Installations (CSNI).
The Data Bank directly manages the Thermochemical Database (TDB) project [3], which started in 1984
following a recommendation by the NEA Radioactive Waste Management Committee (RWMC). The TDB is adatabase of evaluated and qualified thermochemical data, which is required for the safety evaluation of radioactive
waste disposal facilities. The TDB project is carried out through separate funds from 12 countries, and the data isused as an international reference.
Because of the increased importance of peer-reviewed and internationally recognised experimental data,
for example the validation of computer simulation systems, the share of integral experimental data in the total
dispatch number of the CPS has been increasing, particularly in recent years. Assembling, preserving andmanaging data obtained through past and current research and development (R&D) activities is the first step incontributing to the effective use of resources invested in R&D programmes.
NEA Data Bank experience in the area of KPM has thus become a strength to build upon, and the Data
Bank now considers databases as a key pillar to this strategy. The idea is therefore to transition from Computer
Program Services to Computer Program and Database Services, which will centralise NEA wide efforts, including
those involving STCs, so as to preserve, evaluate and distribute data in support of V&V activities for NEA member
countries. This idea was presented to the Data Bank management committee, the Management Board for theDevelopment, Application and Validation of Nuclear Data and Codes (MBDAV), at its 2019 meeting, and will befurther discussed at the MBDAV meeting in 2020.
5. Joint CPS-NDS initiative for a nuclear data and code performance comparison
Aligned with the objective of better data preservation and evaluation, the NDS and CPS are currently
joining efforts to provide better services with regard to the benchmarking of nuclear data in conjunction withcomputer programs, from nuclear data processing to transport codes. The aim is to leverage the unique position of
the NEA Data Bank – as a central point for both nuclear data and computer programs – to share quality-assured
input data for integral experimental benchmark evaluations. This initiative will be carried out using a downselection of both internationally recognised test cases and of computer codes, while promoting the inclusion of thecode suites used in the Data Bank participating countries. This new initiative is now being actively discussed
within the Data Bank and will be debated in more depth by MBDAV members in 2020.
6. Conclusion
The NEA Data Bank has been a centre of excellence for computer codes and nuclear data for over 50
years. Based on this unique experience, the Data Bank is striving to improve its activities through live and directcommunication with end users from all its participating countries.
Each activity at the NEA Data Bank, including those involving Computer Program Services (CPS),
Nuclear Data Services (NDS) and Knowledge Preservation and Management (KPM), can entail a substantial
technical contribution from end users and for end users, and this activity must keep evolving to ensure better
services for end users. This evolution includes not only a change of the framework of our service, such as thetransformation of user licensing, but also new technical proposals related to database development for nuclear dataand computer code validation. Activities related to database development are key to the future of the Data Bankas these offer real benefits to Data Bank participating countries.
The NEA Data Bank sees the needs of its end users as of utmost importance, and it highly appreciates,
and welcomes, any comments and suggestions from end users.
The opinions expressed and arguments employed herein are those of the authors and do not necessarily reflect theofficial views of the OECD, the NEA or of the governments of their member countries.

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JAEA-Conf 2020-001

References[1] F. Michel-Sendis, L. Fiorito and M. Fleming (2018), “Perspectives on nuclear data activities at the Data
Bank: Enhancing the validation function,” NEA News, Vol. 36.2, pp.4-8.
[2] A. Dufresne, E. Poplavskaia, and K. Suyama (2019), “Overview of the Activities of the OECD Nuclear
Energy Agency Data Bank Computer Program Services,” Proceedings of the International Conference onMathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2019), 2529 August 2019, Portland, Oregon, United States.
[3] M.E. Ragoussi and D. Costa, (2019) “Fundamentals of the NEA Thermochemical Database and its
influence over national nuclear programs on the performance assessment of deep geological repositories”,
J. Environ. Radioact. 2019, Vol. 196, p. 225.

- 54 -

JAEA-Conf 2020-001

11 From WKHthe UHVRQDQFH
resonance WKHRU\theoryWR
to WKHthe VWDWLVWLFDO
statisticalPRGHO
model)URP
7RVKLKLNR.$:$12/RV $ODPRV 1DWLRQDO /DERUDWRU\ /RV $ODPRV 10  86$(PDLO NDZDQR#ODQOJRY

WepresentabriefintroductiontotheR-matrixtheoryforthecompoundnuclearreaction,andexplainitsrelationtotheenergy-averagecrosssectionintheunresolvedresonanceregion.
To bridge these two representations of the compound reaction cross section, we utilize theGaussian Orthogonal Ensemble (GOE) embedded in the scattering matrix to calculate theaverage cross sections, and demonstrate calculation of the decay width from a transmission
coeﬃcientthatisamodelinput.

1.

Introduction

157

Gd Capture Cross Section [b]Whenaslowneutroninteractswithanucleus,thereac5
10tion cross section shows a distinct resonating structure,in which each of the peak location corresponds to an
4
10
eigenstate of the compound system just above the neu3
10tron separation energy. In the case of a fast-energy
neutron (in the keV to MeV region), the reaction cross102
section no longer shows sharp resonances because the101
compound states strongly overlap each other. Under
this circumstance, each of the resonances cannot be
100
resolved, and an energy-average cross section is only
-1
10
meaningful. The average cross section can be related
-1
0
1
2
3
10
10
10
10
10
Neutron
Incident
Energy[eV]
to statistical properties of resolved resonances, namely
the average resonance spacing D and decay widths Γ.
157 Gd capture cross
These statistical properties are often studied by apply- Fig. 1: Evaluateding the Gaussian Orthogonal Ensemble (GOE) imple- section from the thermal to the fast
mented in the S and K-matrices. Although this is energy. The thin dotted curve is foran old problem, our recent development on the GOE the 0 K cross section reconstructed
Monte Carlo technique [1, 2] sheds a new light on some from the resolved resonance parameters. The thick histogram is a groupedlong standing compound nuclear reaction problems.
Figure 1 shows the neutron radiative capture cross cross section in the SAND-II 640 ensection of 157 Gd in JENDL-4.0 at low energies, where ergy group structure.
the ﬂuctuation suddenly disappears beyond 300 eV.
This is because the representation of cross section changes at this boundary energy. Above- 55 -

JAEA-Conf 2020-001

300 eV, the cross sections are energy-averaged, while they are not below that energy. This canbe seen when we calculate a group cross section shown by the thick histogram; the cross section
shape in the 100 - 300 eV region is converging to the smooth straight line at higher energies.

2.

Compound Nuclear Reaction

2.1

R-matrix theory

We still base an interpretation of compound
nuclear reaction upon Bohr’s Hypothesis; (i)an incident particle shares its energy withthe target nucleons, (ii) a compound nucleus(CN) attains statistical equilibrium, and (iii)the decay modes of CN are independent offormation. Several theories were formulated
to characterize the resonances in CN, which
are by Breit and Wigner [3], Kapur andPeierls [4], and the most commonly quoted
one is the R-matrix theory of Wigner andEisenbud [5]. R-matrix is deﬁned by the energy of λ-th resonance Eλ , and the decay amplitude γλc into the channel c,










 





    






Fig. 2: The 12 C total cross section is shown(1) on the right side, and the left side is nuclear
Eλ − Eλstructure of 13 C and some of the excited states
which yields the scattering matrix S [6]. In- that correspond to the resonances.
2 isstead of γλc , the decay width Γλc = 2Pc γλcoften used, where Pc is a penetration factor [6]. These parameters are determined by ﬁtting
calculated quantities to available experimental data, and in this sense the R-matrix formula is
not a predictive theory, nevertheless no approximation was made to derive the equation. The
resonance energy Eλ can be estimated from the nuclear excited state in CN when the nuclear
structure is known. In Fig. 2, resonances for neutron-induced reaction on 12 C are depicted on
the right, and the excited state in 13 C corresponding to each resonance is shown on the left.
The ﬁrst resonance occurs at 2.077 MeV, while the 6.864 MeV (5/2)+ level corresponds to thisresonance.
One of the signiﬁcant features of the R-matrix theory is the S-matrix unitarity, S † S = 1,
which ensures the ﬂux conservation. Because of this one can estimate some experimentally
unknown cross section by inverse reactions, e.g., prediction of (n, α) by (α, n) data.
Rcc =2.2

 γλc γλc,Approximations to R-matrix theory

Although the R-matrix theory is exact, a downside is that it requires enough experimental data
to determine all the channels. This situation becomes more severe when the radiative capture- 56 -

JAEA-Conf 2020-001

channel has a lot of ﬁnal states. This is one of the main reasons why the full R-matrix analysis
is limited to light elements.
In the case of resonance reactions on
medium to heavy nuclei, there are many small
photon channels γλc , and their sign is random. When all the photon emission channels
1500
are lumped into a single capture channel Γγ ,as an approximation often made in the single- 1000
level (SLBW) or multi-level Breit-Wigner
1500(MLBW) formulae [7], interference between
0.5
resonances will have some issues. The Reich0
Moore (RM) approximation [8], which elim0
-1
-0.5
Imag
inates all the photon emission channels and
-0.5
0
0.5
lumps them into one capture channel, has
1 -1
Real
better treatment of resonance interference.
The RM R-matrix reads
Fig. 3: S-matrix elements (elastic channel) onthe complex plane for neutron induced reaction
on 184 W with the Reich-Moore approximation.
λThe vertical axis is the neutron incident energy
When these approximations are applied, thein eV.
S-matrix is no longer unitary, and the ﬂux
deﬁcit depends on the size of capture width. Figure 3 demonstrates the S-matrix elements of
184 W for the elastic channel calculated from the RM resonance parameters in JENDL-4. The
incident neutron energy changes from zero at the bottom plane to 1.5 keV at the top. At eachresonance, the S-matrix draws a circle in the complex plane, and smaller circles correspond to
the large capture cross section case, where the unitary deﬁcit is also large.
Although RM and MLBW S-matrices are not unitary, one can still calculate angular distributions of elastically scattered neutrons by applying the Blatt-Biedenharn formula [9]. Thisis particularly important for nuclides near the magic numbers, where non-statistical behaviorof the resonance distribution sometimes enhances the neutron scattering in the forward andbackward angles. Such scattering angular distributions are very diﬀerent from predictions by
the optical model that often gives a very small p-wave contribution in the resonance range [10].
Neutron leakage is enhanced when more neutrons are scattered in the forward angles, whilelarger backward scattering increases an eﬀective neutron multiplicity keﬀ .
RM
Rcc
 =

2.3

γ c γ c.
Eλ − E − iΓγ /2(2)

Random matrix approach in unresolved resonance region

Since individual resonances cannot be resolved anymore above the resolved resonance region,energy-average cross sections and/or average resonance parameters (�Γc � and D) are given in
evaluated nuclear data ﬁles. The channel degree-of-freedom νc of the χ2 -distribution for Γcalso characterizes the distribution of cross sections. When νc = 1, the width distribution is
reduced to the well-know Porter-Thomas distribution. The resonance spacing is also known to- 57 -

JAEA-Conf 2020-001

form the Wigner distribution. Sometimes it is claimed that these distributions are obtained byapplying GOE to the compound nuclear reaction. However, they are not directly connected but
inspired [7]. A realistic implementation of GOE in the scattering matrix S was ﬁrst proposed
by Verbaarschot et al. [11], and ensemble average of the calculated cross sections was performedby the Grassmann integration technique.
The energy-average compound reaction cross section from channel a to b can be deﬁned
by the average decay width
CN
σab
=

2π
D



Γa Γb
c Γc





=

2π �Γa ��Γb �
T a Tb
Wab = 
Wab ,
D
�Γ
�
c c
c Tc

(3)

where Wab is the width ﬂuctuation correction factor [12], and Tc is the particle transmission
coeﬃcient. In Eq. (3), a weak-coupling limit approximation Tc � 2π�Γc �/D is used. This was
examined in the strong-coupling regime [13] by applying the GOE model [1].
Deﬁning the compound cross section by Γc is somewhat ambiguous, nevertheless the statistical model has been developed in such a way. In fact the partial decay amplitude γc in the
GOE model can be deﬁned by two matrices, S and K. When the random matrix H GOE is ﬁrst
diagonalized, the K-matrix that has a similar form to R includes the partial decay amplitudein the numerator,
K = πW T

1W,
E − H GOE

Kcc =

1  γ̃λc γ̃λc
,2 λ E − Eλ(4)where W is the coupling matrix [1]. The average width �Γc � can be evaluated by an ensemble
2 . On the other hand, a pole-expansion form of the S-matrix gives another decay
average of γ̃λcwidth [6].
∗γνc γνcScc = δcc − i.
(5)
ν E − Eν + iΓν /2

The top panels in Fig. 4 show an example of GOE realizations for a two-channel case (elastic
and inelastic scattering only). The left panel is for the weak-coupling case (Tc = 0.1), and theright panel is for the strong-coupling case (Tc = 0.9). The bottom panels are the normalized
average width π�Γa �/D plotted at each resonance energy Eλ . In the small Ta case, both the K
and S-matrices give very similar decay width, while they diﬀer in the strong absorption case.
This situation becomes more complicated when we calculate �Γa Γb / Γc � or �Γa ��Γb �/ �Γc �by the GOE model, hence deﬁnition of the cross section by the decay width would be open
to argument. We are investigating the equality in Eq. (3) by performing the Monte Carlo
calculation for �Γa Γb / Γc � and �Γa ��Γb �/ �Γc �. However, so far what we can conclude isthat the last term of Ta Tb / Tc Wab is the most accurate expression for the compound nuclear
reaction.

3.

Conclusion

A brief introduction to the R-matrix theory for the compound nuclear reaction is given, and its
relation to the energy-average cross section in the unresolved resonance region was summarized.

- 58 -

JAEA-Conf 2020-001

1

10

0

Ta = Tb = 0.9

Elastic
Inelastic
Elastic Cross Section

Elastic Cross Section

Ta = Tb = 0.1
10

-1

10

-2

10

-3

10

-4

10

0

0.1

10

0

Elastic
Inelastic

10-1
-2

10

-3

10
10

0.2

K-matrix
S-matrix

-1

10

-2

-0.1

0

0.1

0.2

-0.1

0

0.1

0.2

K-matrix
S-matrix

100
-1

10

-2

10

10

-3

-0.2

-0.2

101

πι<Γ>/Dπι<Γ>/D

-0.1

100

10

1

-4

-0.2

101

10

-3

-0.1

0

0.1

0.2

10

-0.2

Fig. 4: An example of GOE model realization. The top panels are the calculated cross sections
for the weak and strong coupling cases, where the same random number sequence was used.
The bottom panels are the calculated resonance decay width by K and S matrices.
We utilized the Gaussian Orthogonal Ensemble (GOE), which is embedded in the scatteringmatrix, to calculate the energy-average cross section by a given transmission coeﬃcient. It was
shown that deﬁning compound reaction cross sections in terms of the decay width might besomewhat ambiguous.

Acknowledgment
This work was carried out under the auspices of the National Nuclear Security Administrationof the U.S. Department of Energy at Los Alamos National Laboratory under Contract No.
89233218CNA000001.
References<> T. Kawano, P. Talou, and H. A. Weidenmüller, Phys. Rev. C, 92, 044617 (2015).
<> T.Kawano,R.Capote,S.Hilaire,andP.ChauHuu-Tai,Phys.Rev.C,94,014612(2016).
<> G. Breit and E. Wigner, Phys. Rev.  49, 519 (1936).
<> P. L. Kapur and R. Peierls, Proc. Royal Society A, 166, 277 (1938).
<> E. P. Wigner and L. Eisenbud, Phys. Rev., 72, 29 (1947).
<> A. M. Lane and R. G. Thomas, Rev. Mod. Phys., 30, 257 (1958).

- 59 -

JAEA-Conf 2020-001[7] F.H. Fröhner, “Evaluation and analysis of nuclear resonance data,” JEFF Report 18,
Nuclear Energy Agency Data Bank (2000).
[8] C. W. Reich and M. S. Moore, Phys. Rev., 111 929 (1958).
[9] J. M. Blatt and L. C. Biedenharn, Rev. Mod. Phys., 24, 258 (1952).
[10] T. Kawano and D. A. Brown, J. Nucl. Sci. Technol., 52 274 (2015).
[11] J. J. M. Verbaarschot, H. A. Weidenmüller, and M. R. Zirnbauer, Phy. Rep., 129 367(1985).
[12] P. A. Moldauer, Nucl. Phys. A, 344 185 (1980).
[13] Y. Alhassid, G. F. Bertsch, P. Fanto, and T. Kawano, Phys. Rev. C, 99 024621 (2019).

- 60 -

JAEA-Conf 2020-001
2020-001

12

Current status of search for element 119

Satoshi SAKAGUCHI1,2,3 for nSHE collaboration and for Research Center for Superheavy
Elements
1 Department of Physics, Kyushu University, Motooka 744, Nishi, Fukuoka, 819-0395 Japan
2 Research Center for Superheavy Elements, Kyushu University
3 RIKEN Nishina Center for Accelerator-Based Sciencee-mail: sakaguchi@phys.kyushu-u.ac.jp

Search for the new element 119 is ongoing at RIKEN by the nSHE collaboration. Current status of theexperiment is briefly described. Research Center for Superheavy Elements was recently founded at Kyushu
University. Experimental and theoretical research activities at the center are also explained.

1. Introduction
Search for new elements is a part of an attempt to find the limits of existence of the elements and atomic
nuclei. In addition, new elements may have unique chemical properties which could open up new possibilities
of application. Up to present, 118 elements have been found and named. Those heaviest elements wereartificially synthesized by using the fusion reaction, and the 7th row of the periodic table has been completed.
To extend our knowledge on the physical and chemical properties of heavy elements and nuclei, further
search for the element 119, a new element in the 8th period, was started at RIKEN in 2018 by an internationalnSHE collaboration composed of more than ten institutes. Around the same time, Kyushu University founded
a new research center, Research Center for Superheavy Elements, to reinforce the research activities on thescience of superheavy elements in Japan.

2. Current status of search for element 119
For the efficient synthesis of element 119, a hot fusion method was adopted since it is expected to havea higher production cross section than the cold fusion which was used to synthesize the element 113,
nihonium [1]. It is well known that the cross section increases as the system becomes more “asymmetric”,
i.e. when the atomic number (Z) of the beam nucleus becomes lower and that of the target nucleus higher.
On the other hand, the amount of high-Z target material is severely limited since it has to be artificially
produced in a nuclear reactor making use of its high-flux neutron environment. For the nSHE experiment,248

Cm material is supplied from Oak Ridge National Laboratory (ORNL). The material is processed andprepared as a rotating target at RIKEN. In addition to the 248Cm target, one needs a 51V beam, a separator,
and a detection system. A new separator, GARIS-II was newly constructed [2] aimed at enlarging the angular
acceptance for a higher collection efficiency for hot-fusion products and also at decreasing the backgroundevent rate. The GARIS-II was installed at the RIKEN Ring Cyclotron (RRC) facility in 2017. The search for- 61 -

JAEA-Conf 2020-001
2020-001

element 119 was initiated in January 2018 with a 51V beam accelerated by the RRC [3]. From 2018 to 2019,the nSHE experiment has been performed intermittently. The detection system, consisting of MCP-ToF
detectors and DSSDs, has been continuously upgraded.
In parallel with the nSHE experiment at the RRC facility, the RILAC facility has been upgraded toincrease the beam intensity by 5‒10 times. A 28 GHz superconducting ECR ion source and a superconducting
quarter-wavelength resonator are being installed. To perform the nSHE experiment at the upgraded RILAC
facility, a new separator GARIS-III is under construction. The nSHE experiment will soon be started withthese new devices at the RILAC facility.

3. Research Center for Superheavy Elements at Kyushu University
Research Center for Superheavy Elements (RCSHE) was founded in 2018 at Kyushu University.
Activities related to the nSHE experiment are ongoing as below. (1) A beamline dedicated for SHE research
was constructed at the Center for Accelerator and Beam Applied Science (CABAS) at Kyushu University.
(2) Reaction studies such as measurement of quasi-elastic barrier distribution has been carried out using atandem accelerator. (3) An analysis group of nSHE data has been launched independently of the RIKEN
group to secure complementarity and objectivity of the interpretation of the data. (4) Basic development of
the detectors (MCP-ToF detector for nSHE experiment and new type of an implantation detector utilizing the
inorganic scintillator) has been started.
Related to the superheavy element physics, theoretical studies on fusion and fission processes are being
carried out. Collective inertia of 240Pu was calculated along the fission path by the density functional theory[4]. Development of computer code for estimating evaporation-residue cross sections is planned.
Experimental studies of structures of heavy nuclei are also ongoing at JAEA. Decay spectroscopy of ndeficient Am isotopes was done in FY2019. Through EC decay of 234Am, excited states in 234Pu areinvestigated by the method of gamma-ray spectroscopy at JAEA-ISOL [5]. Isomer spectroscopy using multinucleon transfer reaction on actinides targets will also be performed in near future.

References[1] K. Morita et al., New result in the production and decay of an isotope, 278113, of the 113th element, J.
Phys. Soc. Jpn., 81, 103201 (2012).
[2] K. Morita et al., Discovery of new element, nihonium, and perspectives, Prog. Nucl. Sci. Technol., 5, 8
(2018).
[3] H. Haba, A new period in superheavy-element hunting, Nat. Chem., 11, 10 (2019).
[4] Private Communication[5] Private Communication- 62 -

JAEA-Conf 2020-001

13 Application of CDCC to many-body breakup reaction
Takuma MATSUMOTO
Department of Physics, Kyushu University, Fukuoka 819-0395, Japanemail: matsumoto@phys.kyushu-u.ac.jp
The continuum-discretized coupled-channels method (CDCC) has been successful in analyzing
breakup reactions involving many-body projectiles. The many-body breakup reactions are important for not only nuclear physics but also nuclear engineering studies. In this article, we give someresults of CDCC analyses for studies of unstable nuclei and nuclear data evaluations.

1

Introduction

Breakup reactions have played a key role in investigating the exotic properties of unstable nuclei.
The observables such as breakup and neutron removal cross sections include properties of the groundand resonant states. In order to exclude such exotic properties from the observables, an accurate
analysis of breakup reactions is highly required.
As one of the most reliable methods for treating breakup processes, the continuum-discretizedcoupled-channels method (CDCC) [1, 2, 3] has been proposed. At ﬁrst, CDCC has been applied toanalyses of three-body scattering problems, in which a projectile breaks up into two constituents.
Recently, we have developed CDCC as a method of treating four-body scattering with a three-bodyprojectile. This new version of CDCC is called four-body CDCC [4, 5, 6, 7]. In the four-body CDCC,
breakup continuum states of a projectile are calculated by the pseudo-state method, in which aninternal model Hamiltonian of the projectile is diagonalized in a space spanned by a ﬁnite numberof L2 -type basis functions. The advantage of the pseudo-state method is applicable to analyses
of more than four-body breakup system. As another development of CDCC, the eikonal reaction
theory (ERT) [8] has been proposed. In ERT, we can calculate inclusive breakup cross sections suchas neutron removal cross sections, in which Coulomb breakup processes are consistently describedby CDCC without making the adiabatic approximation used in the Glauber model.
In this workshop, we reviewed recent studies of CDCC for analyses of various reactions involving
unstable nuclei and application to nuclear engineering. This presentation is based on the review
paper of CDCC [3] and Ref. [9].

2Borromean Feshbach resonance of 11 Li

In this section, we introduce the study of a resonance of 11 Li via CDCC analysis of the 11 Li(p,p ) reaction. The existence of a resonance of 11 Li is a long-standing open question. Very recently,
a measurement of the 11 Li(p, p ) reaction at 6 MeV/nucleon [10] with high statistics and high- 63 -

JAEA-Conf 2020-001

resolution has been performed, and a low-lying excited state of 11 Li has clearly been identiﬁed. To
clarify the low-lying state, we analyze the 11 Li(p, p� ) reaction with CDCC.
9Li+n+n nonresonant continuum

10Li+n bound andcontinuum states

εR [MeV]Re[αθνc] > 0.9Re[αθνc] ~ 0
0.46

10Li+n threshold

0.42(Closed Channel)Re[αθνc] = 0.92

0.0

9Li+n+n threshold

Figure 1: A schematic representation of complex-scaled states of 11 Li for I π = 1− is shown.
In CDCC analysis, we adopt a 9 Li + n + n three-body model for 11 Li, assuming for simplicity
that 9 Li is a spinless and inert particle. This simpliﬁed model has been applied to analyses of somereactions of 11 Li, and the details of the calculation are shown in Ref. [9]. In this model of 11 Li, we
found the resonance state for I π = 1− with the resonance energy (decay width) 0.42 MeV (0.14
MeV). Furthermore, we concluded that the resonance is interpreted as a bound state of a valence
neutron with respect to 10 Li, which is a resonance state for the n + 9 Li system with a resonanceenergy (decay width) of 0.46 MeV (0.36 MeV). This type of resonance, which is regarded as abound state embedded in continuum, is referred to as a Feshbach resonance. We thus refer to thisresonance of 11 Li as a Borromean Feshbach resonance. In Fig. 1, properties of the low-lying statesin 11 Li are summarized.
Using the bound, resonant and continuum states of 11 Li calculated above, we analyzed the
11 Li(p, p� ) reaction with CDCC. In the CDCC calculation, we adopt the folding potential between
11 Li and p with the JLM eﬀective interaction [11], and introduce the adjustable parameter λ that isI
a normalization factor of the imaginary part of the folding potential. The details of the calculation
are shown in Ref. [9]. Figure 2 shows the breakup cross section with respect to the three-body energy
ε after breakup. Here the energy resolution of the experimental data are taken into account. One
sees that the contribution from the dipole resonance of 11 Li shown by the thin solid line dominates
the low-lying peak. Up to ε ∼1.0 MeV, it can be concluded that the low-lying continuum structureof 11 Li including the resonance is consistent with the measured cross sections. On the other hand,the calculated cross section undershoots the data for ε  1.0 MeV, which will be due to some other
degrees of freedom that are not taken into account in the present calculation, e.g., a transition
to higher spin states and a core excitation in 9 Li. In order to clarify properties of the continuum
structure of 11 Li including the higher excited states, we require more reliable calculations of 11 Li
with the core excitation in 9 Li.

- 64 -

d σ dε

JAEA-Conf 2020-001

εFigure 2: The breakup cross section as a function of the three-body breakup energy ε of 11 Li in the
11 Li(p, p� ) reaction [10]. The cross section is integrated over θ◦◦c.m. from 115 to 124 . The dotted,
dashed, and dot-dashed lines represent calculated cross sections to the 0+ , 1− , and 2+ breakup
states, respectively, and the sum of them is shown by the thick solid line. The thin solid line shows
the contribution of the three-body resonance of 11 Li.

3Application of CDCC to nuclear engineering

Next we show results of an application of CDCC to evaluating of nuclear data. Since CDCC is afully quantum-mechanical method, it is applicable to reactions at low incident energies, which areimportant for nuclear engineering. Recently, accurate nuclear data of nucleon induced reactions on
6,7 Li that is important material for DT fusion reactor are highly required. In Ref. [12], we analyzed
n + 6 Li scattering by the three-body CDCC method, in which 6 Li is described as a d + α system.
In this analysis, we calculate diagonal and coupling potentials between n and 6 Li by using thefolding model with the JLM eﬀective interaction [11] with λw that is optimized to reproduce theelastic cross sections. Details of the calculation are shown in Ref. [12].
Figure 3 shows the diﬀerential elastic cross sections of n + 6 Li for incident energies between 7.47
and 24.0 MeV. One sees that the results calculated by CDCC (the solid lines) are in good agreementwith the experimental data. The dashed lines represent the results of a single-channel calculation
without couplings to the breakup states of 6 Li. For all incident energies, we take λw = 0.1 toreproduce the data. It should be noted that the single-channel calculation cannot reproduce theexperimental data with any values of λw . Thus breakup eﬀects are signiﬁcant to reproduce theangular distributions of the elastic scattering. Very recently we also applied CDCC to scattering of
7 Li and the results are summarized in Ref. [16].

- 65 -

σ Ωσ Ω

JAEA-Conf 2020-001

θθ

Figure 3: Angular distribution of the elastic diﬀerential cross section of n + 6 Li scattering for incident energies between 7.47 and 24.0 MeV. This ﬁgure is taken from Ref. [12]. The solid and dashedlines correspond to the result with and without couplings to breakup states of 6 Li, respectively.
Experimental data are taken from Refs. [13, 14, 15].

4

Summary

Many-body breakup reactions are important for not only nuclear physics but also nuclear engineering studies. CDCC is more reliable method to describe many-body breakup processes accurately.
Recently, we analysed the the 11 Li(p, p ) reaction with CDCC. In the analysis, we found that 11 Li
might have a resonance in the low-lying structure, and the resonance is regarded as a Feshbach typeresonance. The resonance of 11 Li is referred to as Borromean Feshbach resonance. Furthermore,
we also applied CDCC to analyses of 6 Li(n, n’) and 7 Li(n, n ) reactions, which are important for
nuclear data evaluations. In the analyses, CDCC well reproduces both the elastic and inelastic crosssections, and we then expect that CDCC can predict the cross sections if the experiment data do
not exist. Thus CDCC is indispensable for not only study of unstable nuclei but also nuclear data
evaluations.

Acknowledgements
The author is grateful to K. Ogata and J. Tanaka for helpful discussion. Furthermore I wish to thankto members of JCPRG in Hokkaido University and Y. Watanabe and H. Guo for the discussion of- 66 -

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nuclear data evaluation. This work was supported in part by a Grant-in-Aid for Scientiﬁc Research(No. JP18K03650) from the Japan Society for the Promotion of Science (JSPS).

References[1] M. Kamimura et al., “Chapter I. Projectile Breakup Processes in Nuclear Reactions”, Prog.
Theor. Phys. Suppl. 89 (1986), 1.
[2] N. Austern et al., “Continuum-discretized coupled-channels calculations for three-body models
of deuteron-nucleus reactions”, Phys. Rep. 154 (1987), 125.
[3] M. Yahiro et al., “The continuum discretized coupled-channels method and its applications”,
Prog. Theor. Exp. Phys. 01A206 (2012).
[4] T. Matsumoto et al., “Continuum-discretized coupled-channels method for four-body nuclear
breakup in 6 He+12 C scattering”, Phys. Rev. C 70 (2004), 061601(R).
[5] M. Rodrı́guez-Gallardo et al., “Three-body continuum discretization in a basis of transformed
harmonic oscillator states”, Phys. Rev. C 72 (2005), 024007.
[6] T. Matsumoto et al., “Coulomb breakup eﬀects on the elastic cross section of 6 He+209 Bi
scattering near Coulomb barrier energies”, Phys. Rev. C 73 (2006), 051602(R).
[7] M. Rodrı́guez-Gallardo et al., “Four-body continuum-discretized coupled-channels calculations”, Phys. Rev. C 80 (2009), 051601(R).
[8] M. Yahiro, K. Ogata, and K. Minomo, “Eikonal Reaction Theory for Neutron Removal Reaction”, Prog. Theor. Phys. 126 (2011), 167.
[9] T. Matsumoto, J. Tanaka, and K. Ogata, “Borromean Feshbach resonance in 11 Li studied via
11 Li(p, p� )”, Prog. Theor. Exp. Phys. 123D02 (2019).
[10] J. Tanaka et al., “Halo-induced large enhancement of soft dipole excitation of 11 Li observed
via proton inelastic scattering”, Phys. Lett. B 774 (2017), 268.
[11] J.-P. Jeukenne, A. Lejeune, and C. Mahaux, “Optical-model potential in ﬁnite nuclei from
Reid’s hard core interaction”, Phys. Rev. C16 (1977), 80.
[12] T. Matsumoto et al., “Systematic description of the 6 Li(n, n� )6 Li∗ → d + α reactions with themicroscopic coupled-channels method”, Phys. Rev. C 83 (2011), 064611.
[13] S. Chiba et al. , “Measurement and theoretical analysis of neutron elastic scattering andinelastic reactions leading to a three-body ﬁnal state for 6 Li at 10 to 20 MeV”, Phys. Rev.
C58 (1998), 2205.
[14] H. Hogue et al., “Elastic and Inelastic Scattering of 7- to 14-MeV Neutrons from Lithium-6
and Lithium-7”, Nucl. Sci. Eng. 69 (1979), 22.
[15] L. Hansen et al., “Proton and neutron transition densities in 6,7 Li from low energy neutron
and proton scattering”, Phys. Rev. C38 (1988), 525.
[16] H. Guo et al., “Analysis of nucleon and triton emissions from nucleon-7 Li collisions below 20
MeV”, Phys. Rev. C99 (2019), 034602.

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14

Topics from Radiation Safety Design of J-PARC
Hiroshi NAKASHIMA
Japan Atomic Energy Agency
Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 Japane-mail: nakashima.hiroshi@jaea.go.jp

J-PARC (Japan Proton Accelerator Research Complex) is a high-energy proton accelerator complex of theworld’s highest beam power.

Because of its very high beam power and its high energy as well as thelarge-scale accelerator complex, we encountered some very difficult problems on radiation safety design.
Various examinations and countermeasures were considered in order to overcome the difficulty. This paper
introduces some of them. In addition, some new knowledges obtained during 10 years after completion are
described.
1. Introduction
J-PARC is a complex of GeV class high energy accelerator facilities with the world's largest beam power,
jointly organized by JAEA (Japan Atomic Energy Agency) and KEK (High-energy accelerator laboratory).[1]The accelerator consists of 400MeV LINAC, 3GeV synchrotron and a main ring (currently 30GeV). These
accelerator facilities include the Materials and Life Science Experimental Facility, the Hadron Experimental
Facility, and the Neutrino Experimental Facility. The experimental facilities are conducting research in various
fields. In the future, a transmutation experiment facility will be also built.
This paper summarizes J-PARC's topics related to radiation safety design, including measures at the timeof design and new knowledges obtained 10 years after the completion of the facility.
2．Basic concept in radiation safety design[2]J-PARC has many problems to be overcome from the viewpoint of radiation safety as the followings.
1) It is necessary to conduct a large-scale radiation safety estimation integrally.
2) Radiation shielding design is needed to be performed effectively and economically.
3) It is necessary to take measures for activation on structures, equipment, air, cooling water, etc.
Therefore, radiation safety design was performed based on the following concept.
1) Use a design method system that combines the simple methods and the detailed calculation methods.
2) Accelerator facilities are basically installed underground.
3) Air and cooling water are circulated in a closed loop during operation, and after shutdown, monitoring is
performed to control the activities. Low activation materials are used for buildings and equipment.
4) Monitor and limit the beam loss to keep radiation source condition on safety.
3. Topics from radiation safety design
3.1 Radiation source condition
It is the most important in a radiation safety design to determine the radiation source conditions. In
accelerator facilities, the source condition is proportional to the beam loss at each loss points. Therefore, we
determined the beam loss amount at the initial design stage for the entire facility.
Beam losses of targets and beam dumps in each facility were determined by the facility plan, and those of
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assume those of the accelerators itself. Beam loss of 1W/m was adopted based on reports of the ICFA
(International Committee on Future Accelerators) and KEK.[3] The effective dose caused by the beam loss inworking environment makes possible hands-on-maintenance after shutdown. The assumption has been verified
by estimation on the effective doses caused by the beam loss.[4] During operation, the beam losses are measured
by a real-time monitor along all accelerators.
3.2 Design method
Semi-empirical formulas and/or simplified methods were applied for the overall design. On the other hand,detailed design methods such as Monte Carlo methods were used for high beam loss points and complicatedstructures such as targets and injectors.[2]As for the simplified methods on deep penetration, the Moyer’s Model [5] was used in the proton energy
region above 1 GeV, while the Tesch’s equation [6] was used below 1 GeV. In the calculation the KEK parameter
was adopted as the attenuation coefficient. After the design, it has been confirmed that the attenuation coefficient
has been evaluated more safely by experiments at BNL (the Brookhaven National Laboratory) and FNAL (theFermi National Accelerator Laboratory).[7, 8] For streaming in mazes and ducts, DUCT-III [9] was applied andthe results were validated by experimental data at TIARA (Takasaki Ion Accelerators for Advanced Radiation
Application)[10]. As for skyshine, calculations using Stapleton's equation [11] were performed to estimate thedoses at the site boundary. The accuracy was verified by comparing it with calculation using the Monte Carlo
method and SHINE3 [12], because no experiment existed on the high energy accelerator facility. It is expectedthat benchmark experiments will be conducted.
An integrated calculation system: SSCAT (Simplified Shielding Calculation Table System) [13] wasdeveloped to comprehensively calculate the dose of all facilities on bulk shielding and skyshine. As for the doseat the site boundary, the dose calculated by the Monte Carlo methods in some facilities is also added to the dose
calculated by this system.
In the detailed calculation system, Monte Carlo codes: PHITS [14], MCNPX [15] and MARS [16] wereused for high-energy particle transport, as shown in figure 1. They were used differently according to facilities
and their conditions, because each code has its own characteristics. DCHAIN-SP [17] was used as a radioactivity
calculation code, and QAD-CGGP2 [18] was used for radiation generated by radioactivity.
The results by each code show a difference among them. Therefore, the difference was validated based onshielding experiments[19]. The benchmarking was carried out for the following shielding items:1) Thick Target
Neutron Yield, 2) Beam dump, 3) Deep penetration, 4) Streaming and 5) Skyshine. All are based on experiments
except for skyshine. A safety factor of 2 was obtained from the benchmarking. By using the safety factor,
estimation values were obtained for licensing. A report summarizing the results has been submitted as a reference
for licensing. Even after the benchmarking was conducted, benchmark experiments have been conducted using
RCNP [20], FNAL [8], and so on. Some of the experimental results have been compiled as an experiment
database: SINBAD [21] in OECD / NEA (the Organization for Economic Co-operation and Development / theNuclear Energy Agency).
At J-PARC, the accuracy was systematically verified. However, an accuracy verification system forshielding calculation codes has not been established in Japan as a whole. As a result, old-fashioned shielding
codes are still used.[22] In order to apply the newest knowledges on shielding to licensing, it is desirable to build
an authorized protocol that apply newly developed data and codes in licensing.

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Radiation

Activity

High-energy particle flux

Residual
nucleiPHITS, MCNPX, MARS
(En>20MeV, Ei>1MeV
i=p, pion, meson)
Neutronsource forlow energy
calculation

Nuclear decay
data set
Neutron flux
Particle flux
2nd-ray

Nuclear decay
DCHAIN-SP 2001
Nuclear
Production
X’sectLow-energy particle flux
MCNP

(En<20MeV, 2nd-ray)

Neutrontransport
X’sect for
low energy

Air,Water
Dose
Site boundary
In siteControlled area

Devices

Residual Activity

ray fromresidual
nucleiray transport
QAD-CGGP2, MCNP
Dose due toresidual nuclei

Figure 1 Calculation flow of radiation and activity for J-PARC.[2]3.3 Radiation confinement system
The high-energy radiation produced by the beam loss causes a nuclear reaction with surrounding air andcooling water, and generates various radioactivity. It is very difficult to evaluate the produced radioactivity
accurately, because the energy range is very wide, up to GeV, and many kinds of reaction channels open.
However, the reactions at high energy generate unstable nuclides such as 13N (T1/2 9.97 minutes) and 15O (T1/2
2.04 minutes) having a half-life of several minutes or less. Therefore, those that affect the environment are
confined and released to the outside after confirming the activity concentration by monitoring them.
A conceptual diagram is shown in figure 2.[2] A buffer region, that is a controlled area, is placed between
the accelerator room: controlled area, and the outside uncontrolled area. During operation, air is circulated in theaccelerator room for air conditioning. The accelerator room is maintained at a negative pressure compared to thebuffer region, and the air in the buffer region is monitored. If this monitor is abnormal, the accelerator is stopped
immediately. After the operation and the cooling time, it is possible to access to the accelerator room withconfirming safety by monitoring the concentration in air. During exhaust, 7Be is removed by using a HEPA filter.
Cooling water is also stored once in disposal tanks. After removing 7Be using an ion exchange resin andmonitoring the concentration, the water is drained to the outside.

Negative pressure and monitoring
Way, Duct(Door, Caulking）

Accelerator room
(Contaminated area)

Buffer region（Controlled area）

Outside（Uncontrolled area）

Figure 2 Conceptual view of air confinement system- 71 -

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3.4 Low-activation concrete
The high-energy radiation generated by the beam loss also activates concrete on the accelerator room wall.
The radioactivity makes increase the irradiation dose to workers in the accelerator room during maintenance.
Therefore, low-activation concrete was adopted to reduce the dose.
There are two types of low-activation concrete depending on the purpose. One is to reduce theenvironmental burden at decommissioning, and the other is to improve the working environment as this case. In
the former case, the long-lived 60Co and 152, 154Eu are the nuclides of interest. On the other hand, in the latter case,
Na is the nuclide of interest, which makes increase the dose in working environment several days after the24

shutdown. Therefore, J-PARC has adopted criteria for 24Na.[23]In concrete, 24Na is mainly produced by the reaction of 23Na(n, γ)24Na, 24Mg(n, p)24Na, 27Al(n, α)24Na, andSi(n, X)24Na. Although the production rate is precisely calculated by the neutron flux and each reaction cross28

section, it is very difficult to do so. Thus, the 24Na production ratio has been set as follows by their cross sections
approximately: 23Na = 1, 24Mg = 0.02, 27Al = 0.01 and 28Si = 0.002. The 24Na-equivalent is defined by thefollowing formula with weight percent of the concrete composition.
Na-equivalent (g/m3) = ∑24Na production ratio (%) × weight of origin nuclei on 24Na (g/m3)

24(Natural abundance %)Resultantly, limestone was selected as a concrete aggregate for low activation concrete, and the residual
dose in the concrete was reduced by an order of magnitude. It should be noted here that the quality of limestone
must be well managed during construction to keep the criteria.
3.5 Activation by muon
A series of experimental study on shielding and radiation effects at FNAL have been carried out undercollaboration between FNAL and Japan.[8] The purposes of the experiments were mainly measurement ofparticle flux and residual activity inside and around shields at Pbar station (anti-proton target station) and NuMI
beamline (beamline for neutrino oscillation experiment) with 120GeV protons. In the study a new result relatedwith J-PARC was obtained by Prof. Matsumura, et. al, that was activation caused by muon interaction withmatters.[24]At NuMI, as neutrino experimental facility in J-PARC, proton beam strikes a carbon target and produce
pions in the forward direction. The pions decay into neutrinos and muons at the decay volume, and the neutrinos
are used for neutrino experiments. Most of the muons are absorbed by the iron shield located at downstream ofthe decay volume.
In this experiment, the distributions of neutron fluxes and of generated radioactivity were measured by theactivation foil method in the tunnel downstream of the iron shield. The mass distributions produced in the copper
foil were measured, and the dependency on the incident particles was identified. As a result, it was founded thatthe dependency was similar to that of the photonuclear reaction. Only a few photons are measured in the tunnel
and it is known that a pseudo-photonuclear reaction occurs in the muon nuclear reaction. Therefore, it isconfirmed that the reaction is caused by muon.
It is necessary to confirm the muon effect on activation, because the muon reaction is not considered in
the radiation safety design of J-PARC.
3.6 Effects of Λ and Σ on neutron transmission
International comparisons among codes developed by research institutes around the world has been

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conducted at SATIF (the task- force on Shielding aspects of Accelerators, Targets and Irradiation Facilities) of
OECD/NEA for more than 20 years, in order to compare the accuracy of high-energy particle transport
calculation codes. In the SATIF-10, the neutron dose distribution, neutron energy spectrum and attenuation
coefficient in iron and concrete bulk shields under the conditions of 0.1 to 100 G eV proton injection werecompared.[25]Figure 3 Comparison of dose equivalent rate in iron for incident of 10 GeV protons.
The neutron dose distribution in a 10 GeV incident iron shield are compared in figure 3. The comparison
shows that the calculation by PHITS became larger as compared with calculations by other codes. This seemed
to come from the fact that PHITS considers transport of baryons such as Λ and Σ. Because baryons have a higher
mass than neutrons and protons, more energy is carried forward. When Λ and Σ decays, neutrons and protons are
generated. Further, Λ and Σ have a higher generation ratio in this energy region than those of Ξ and Ω. As a result,
it is considered that Λ and Σ increase the neutron dose. The phenomenon has not been calculated by other codes,
and it is uncertain whether other codes calculate baryon production.
This may affect to neutron dose downstream of the hadron experimental facility at J-PARC. It is necessary
to verify the phenomenon experimentally.
4. Summary
J-PARC is a high-energy proton accelerator complex of the world’s highest beam power, and its
characteristics caused many difficulties on radiological safety. In order to overcome the difficulties and securesafety, various countermeasures are adopted and many shielding design methods were also applied. The designand construction of J-PARC have affected to improve radiation safety of other facilities. After completion, manykinds of newly knowledges have been obtained. Based on the knowledges, the radiological safety of J-PARC andother high energy accelerator facilities may be the time to be reconfirmed.
References[1] S. Nagamiya, J. of Nucl. Materials, 343, 1 (2005).
[2] H. Nakashima, et al., “Radiation Shielding Design for the J-PARC Project”, Proc. on shielding aspects ofaccelerators, targets and irradiation facilities - SATIF 8 (2006), NEA/NSC/DOC(2010)6, 79-87 (2010).
[3] I. Yamane, “Radiation Protection Measures for the Maintenance Service of JHF 3-GeV Ring”, KEK
Internal Report, 98-7 (1998).

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2020-001[4] H. Nakashima, et al., “Estimation of activity and dose distributions around a proton linac induced by beam
spill”, J. Nucl. Sci. Technol., Suppl.1, 870-874 (2000).
[5] B. J. Moyer, “Method of Calculation of the Shielding Enclosure for the Berkley Bevatron”, Proc. on 1st Int.
Conf. Shielding around High Energy Accelerators, Presses Universitaires de France, Paris, 65 (1962).
[6] K. Tesch, “A Simple Estimation of the Lateral Shielding for Proton Accelerators in the Energy Range 50 to1000 MeV”, Radiat. Prot. Dosim., 11, 165 (1985).
[7] H. Nakashima, et al., “Current Status of the AGS Spallation Target Experiment”, Proc. on shielding aspectsof accelerators, targets and irradiation facilities - SATIF 6 (2002), OECD/NEA No.3828, 27-36 (2004).
[8] H. Nakashima, et al., “Research activities on JASMIN: Japanese and American Study of Muon Interaction
and Neutron detection”, J. Nucl. Sci. Technol., Suppl. 4, 191 (2014).
[9] R. Tayama, et al., “DUCT-III: A Simple Design Code for Duct-Streaming Radiations”, KEK Internal,
2001-8, (2001).
[10] F. Masukawa, et al., “Analyses of Streamed Neutron Spectra at TIARA Using DUCT-III”, J. Nucl. Sci.
Technol., Suppl. 4, 46 (2004).
[11] G. B. Stapleton, “Accelerator Skyshine: Tyger, Tyger, Burning Bright”, Particle Accelerators, 44, 1 (1994).
[12] T. Tsukiyama, et al., “Simple Code for Skyshine Dose Calculation up to 3 GeV Neutrons”, J. Nucl. Sci.
Technol., Suppl. 1, 640 (2000).
[13] T. Oguri, et al., “Revised Version of SSCAT: Simplified Shielding Calculation System for High Energy
Proton Accelerator Facilities”, JAEA-Data/Code 2011-026 (2012).
[14] T. Sato, et al, “Particle and Heavy Ion Transport code System, PHITS, version 2.52”, J. Nucl. Sci. Technol.,50[9], 913 (2013).
[15] N. V. Mokhov, “The Mars Code System User's Guide,” Fermilab-FN-628 (1995), N. V. Mokhov and O. E.
Krivosheev, “MARS Code Status”, Fermilab-Conf-00/181 (2000).
[16] L. S. Waters (Ed.), “MCNPX Users Manual – Version 2.1.5”, TPO-E83-G-UG-X-00001, LANL (1999).
[17] T. Kai, et al., “DCHAIN-SP 2001: High Energy Particle Induced Radioactivity Calculation Code”,
JAERI-Data/Code 2001-016, JAERI (2001) (in Japanese).
[18] K.A. Litwin, et al, “Improvements to the Point Kernel Code QAD- CGGP: A Code Validation and User's
Manual”, RC-1214, COG-94-65, AECL Research (US) (1994).
[19] N. Matsuda, et al., “Analyses of Shielding Benchmark Problems for the Shielding Design of the HighIntensity Proton Accelerator Facilities”, JAEA-Technology 2008-030 (2008).
[20] T. Matsumoto, et al., “Shielding experiments of concrete and iron for the 244 MeV and 387 MeV
quasi-mono energetic neutrons using a Bonner sphere spectrometer”, Les Ulis 153 EDP Sciences, (2017).
[21] I. Kodeli, et al., “20 Years of SINBAD (Shielding Integral Benchmark Archive and Database)”, J. Nucl. Sci.
Technol., Suppl. 4, 308 (2014).
[22] H. Nakashima, et al., “Old-fashioned system on development of radiation transport calculation codes inJapan”, Isotope News, 745, 11 (2016) (in Japanese).
[23] N. Matsuda, et al., “A study on induced radioactivity in a low-activationized concrete for J-PARC”, J. Nucl.
Sci. Technol., Suppl. 4, 74 (2004).
[24] H. Matsumura, et al., “Shielding Experiments under JASMIN Collaboration at FERMILAB (III)Measurement of High-Energy Neutrons Penetrating a Thick Iron Shield from the Antiproton Production Target
by Au Activation Method”, Journal of the Korean Physical Society, 59[2], 2059 (2011).
[25] H. Hirayama, “Inter-comparison of Medium-Energy Neutron Attenuation in Iron and Concrete (7)”, Proc.
on shielding aspects of accelerators, targets and irradiation facilities – SATIF9 (2010), NEA/NSC/DOC(2010)7,
261-274 (2010).

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Production cross sections of 45Ti via deuteron-induced reaction on

15

45

Sc

Tsoodol ZOLBADRAL1,2*, Masayuki AIKAWA1,3, Dagvadorj ICHINKHORLOO2,3, Khishigjargal
TEGSHJARGAL4, Yukiko KOMORI5, Hiromitsu HABA5, Sándor TAKÁCS6, Ferenc DITRÓI6, ZoltánSZŰCS6
1

Graduate School of Biomedical Science and Engineering, Hokkaido University, Sapporo 060-8638,
Japan
2

Nuclear Research Center, National University of Mongolia, Ulaanbaatar 13330, Mongolia
3

4

Faculty of Science, Hokkaido University, Sapporo 060-0810, Japan

School of Engineering and Applied Sciences, National University of Mongolia, Ulaanbaatar 14201,
Mongolia
5

Nishina Center for Accelerator-Based Science, RIKEN, Wako 351-0198, Japan
6

Institute for Nuclear Research (ATOMKI), 4026 Debrecen, Hungary*

E-mail: zolbadral@nds.sci.hokudai.ac.jp

Production cross sections of the medical radionuclide 45Ti with the deuteron-induced reactions on 45Sc
were investigated. The stacked foil activation method and γ-ray spectrometry were used. The physical yieldof 45Ti was deduced from the measured cross sections. Our results are consistent with previous experimental
data.
1. Introduction
The radinuclide 45Ti has a half-life of 184.8 min and is a positron emitter (𝐸𝐸��� � 1040 keV, 𝐸𝐸��� � 439

keV, 𝐼𝐼�� � 84.8%) [1]. The emitted positrons followed its decay are suitable for positron emission

tomography (PET) [2]. The deuteron-induced reaction on a 45Sc target is a possible route to produce thisradionuclide at cyclotrons. However, only one experimental study [3] on the cross sections of the45

Sc(d,2n)45Ti reaction is available in the literature. Therefore, we measured the cross sections of the45

Sc(d,2n)45Ti reaction.

2. Experimental methods
The experiment was performed at the AVF cyclotron of the RIKEN RI Beam Factory. The stacked-foil
activation technique and γ-ray spectrometry were adopted to measure the cross sections.
The stacked target consisted of metallic foils of 45Sc (25 m thick, 99.0% purity, Nilaco Corp., Japan and250 m thick, 99.9% purity, Johnson Matthey Alfa Products company, USA), 27Al (99.6% purity, Nilaco
Corp., Japan) and natTi (99.6% purity, Nilaco Corp., Japan). The size and weight of the foils were measured
to determine the average thicknesses of the foils. The derived thicknesses were 7.71 and 76.0 mg/cm2 for--7575--

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Sc, 4.99 mg/cm2 for 27Al and 9.13 mg/cm2 for natTi, respectively. The foils were cut for the size of 8×8 mm

to fit a target holder served also as a Faraday cup.
The target was irradiated for 30 min with a 24-MeV deuteron beam. The incident beam energy wasmeasured by the time-of-flight method [4]. The energy degradation in the stacked target was calculated by
the SRIM code [5]. The beam intensity was determined by collecting the charge in the Faraday cup. γ-rays
were measured for each irradiated foil by a high-resolution HPGe detector. The detector was calibrated by amixed standard γ-ray point source. The dead time was kept less than 7% in the measurements.
The cross sections of the natTi(d,x)48V monitor reaction were used to assess the beam parameters. The crosssections were derived from measurements of the 983.5-keV γ-ray (Iγ = 99.98%) emitted with the 48V decay(T1/2 = 15.9735 d), and the result was compared with the IAEA recommended values [6]. The beam intensity
measured by the Faraday-cup was decreased by 3.0 % to have good agreement among the derived and therecommended excitation function of the natTi(d,x)48V reaction (Fig. 1). The corrected beam intensity (175.2
nA) was adopted in the data assessment for the cross sections.

Fig. 1. The excitation function of the natTi(d,x)48V monitor reaction compared with the recommended
values [6].
3. Results and Discussion
The cross sections of the 45Sc(d,2n)45Ti reaction were derived from the measurement of the 719.6-keV γray (𝐼𝐼� = 0.154 %) emitted in the decay of 45Ti, and are shown in Fig. 2 in comparison with the previous data
[3] and the theoretical estimation of TENDL-2017 [7]. Our result is consistent with the previous data. The

peak position of the TENDL-2017 data [7] is slightly shifted to the lower energy than those of the two
experimental datasets.

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Fig. 2. Excitation function of the 45Sc(d,2n)45Ti reaction.
The physical yield of 45Ti was deduced from a spline fitted curve of the measured excitation function andstopping power calculated using the SRIM code [5]. The derived yield is shown in Fig. 3 in comparison withthe only experimental data [8]. The present yield curve of 45Ti is slightly larger than the experimental data at
22 MeV [8].

Fig. 3. Physical yield of 45Ti via the deuteron-induced reactions on 45Sc.
There are no isotopic impurities produced below 15.3 MeV, which is the threshold energy of the45

Sc(d,3n)44Ti reaction. Production of 45Ti without radio-contamination is possible by using the 45Sc(d,2n)45Ti

reaction in the energy range from 15 to 5 MeV. Only the stable 46Ti is co-produced in the (d,n) reaction incomparable quantity by the prediction based on the cross section of the (d,n) reaction taken from the TENDL2017 data library.
4. Conclusion
The excitation function of the 45Sc(d,2n)45Ti reaction was measured up to 24 MeV. The stacked-foil
activation technique and the high-resolution γ-ray spectrometry were used for the cross section measurements.

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The obtained data were compared with the previous experimental data and the TENDL-2017 data. The
derived excitation function of the 45Sc(d,2n)45Ti reaction is consistent with the data of Hermanne et al. (2012).
The physical yield deduced from measured cross sections is slightly larger than the experimental data of
Dmitriev et al. (1983). The production of radioactive-contamination-free 45Ti can be obtained via the45

Sc(d,2n)45Ti reaction using cyclotrons in the energy region below 15 MeV.

Acknowledgements
The experiment was carried out at RI Beam Factory operated by RIKEN Nishina Center and CNS,
University of Tokyo, Japan. This work is supported by JSPS KAKENHI Grant Number 17K07004. Ts.Z was
granted a scholarship by the M-JEED project (Mongolian-Japan Engineering Education Development
Program, J11B16).
References[1]

National Nuclear Data Center, Nuclear structure and decay data on-line library, Nudat 2.7 [Internet]. 2017.
Available from: http://www.nndc.bnl.gov/nudat2/[2]

Vavere AL, Laforest R, Welch MJ. Production, processing and small animal PET imaging of titanium-45.
Nucl. Med. Biol., 2005, 32, 117–122.

[3]

Hermanne A, Adam Rebeles R, Tarkanyi F, et al., Cross sections of deuteron induced reactions on 45Sc up
to 50 MeV: Experiments and comparison with theoretical codes, Nucl. Instruments Methods Phys. Res.
Sect. B Beam Interact. with Mater. Atoms, 2012, 270, 106–115.

[4]

Watanabe T, Fujimaki M, Fukunishi N, et al., Beam energy and longitudinal beam profile measurement
system at the RIBF, Proc. 5th Int. Part. Accel. Conf. (IPAC 2014), 2014, 3566–3568.

[5]

Ziegler JF, Ziegler MD, Biersack JP, SRIM - The stopping and range of ions in matter (2010), Nucl.
Instruments Methods Phys. Res. Sect. B Beam Interact. with Mater. Atoms, 2010, 268, 1818–1823.

[6]

Hermanne A, Ignatyuk A V., Capote R, et al., Reference Cross Sections for Charged-particle Monitor
Reactions, Nucl. Data Sheets, 2018, 148, 338–382.

[7]

Koning AJ, Rochman D, Sublet JC, et al., TENDL: Complete Nuclear Data Library for Innovative Nuclear
Science and Technology, Nucl. Data Sheets, 2019, 155, 1–55.

[8]

Dmitriev PP, Krasnov NN, Molin GA, Yields of radioactive nuclides formed by bombardment of a thicktarget with 22-MeV deuterons, INDC(CCP)-210, 1983, 1–8.

--7878--

JAEA-Conf 2020-001

16

Production of 169Yb by the proton-induced reaction on 169Tm

Moemi SAITO1, Masayuki AIKAWA1,2*, Tomohiro MURATA1, Yukiko KOMORI3, Hiromitsu HABA3,
Sándor TAKÁCS4, Ferenc DITORÓI4 and Zoltán SZÜCS4
1

Graduate School of Biomedical Science and Engineering, Hokkaido University, Sapporo 060-8638, Japan
2
3

Faculty of Science, Hokkaido University, Sapporo 060-0810, Japan

Nishina Center for Accelerator-Based Science, RIKEN, Wako 351-0198, Japan
4

Institute for Nuclear Research (ATOMKI), 4026 Debrecen, Hungary*e-mail: aikawa@sci.hokudai.ac.jp

Activation cross sections of the 169Tm(p,n)169Yb reaction were measured using the MGC-20E cyclotron
at Institute for Nuclear Research (ATOMKI), Hungary. The stacked-foil activation technique and the highresolution gamma-ray spectrometry were adopted to derive the cross sections. The preliminary cross sections
were compared with experimental data studied earlier and with theoretical model calculation.

1. Introduction
The radionuclide 169Yb (T1/2 = 32.018 d) decays with emissions of Auger electrons and X-rays. These
properties of 169Yb are appropriate for brachytherapy [1]. For such practical use, the best reaction for its
production should be selected among all possible reactions. One of them is the neutron capture reaction on
168

Yb, of which the isotopic ratio is only 0.123%. Another route is a reaction using charged particles, such as

proton, deuteron and alpha particles. We have systematically investigated the production reactions of 169Yb
with a focus on the charged-particle induced reactions. The experiments on deuteron- [2] and alpha-induced
reactions on 169Tm, and alpha-induced reactions on natEr [3] have already been performed. Therefore, we
carried out an experiment of the proton-induced reaction on 169Tm. In this paper, we report the preliminary
result of this experiment.

2. Experimental
The experiment was performed using the MGC-20E cyclotron at Institute for Nuclear Research(ATOMKI), Hungary. The stacked-foil activation technique and the high-resolution gamma-ray spectrometry
were adopted to derive cross sections.
The target consisted of metallic foils of 169Tm (99.0% purity) and natTi (99.6% purity). The sizes andweights of the foils were measured for determining the average thicknesses of the Tm and Ti foils. The
thicknesses were found to be 21.0 and 9.1 mg/cm2, respectively. The foils were cut into small pieces of 6×
6 mm2 and stacked into a target holder, which was also served as a Faraday cup. The target was irradiated for
30 min with an 18 MeV proton beam. Energy degradation in the target was calculated using the SRIM code

- 79 -

JAEA-Conf 2020-001[4]. The average beam intensity was 210 nA, which was measured by the Faraday cup. Gamma rays emitted
from each irradiated foil were measured by an HPGe detector without chemical separation. Nuclear datarequired to derive cross sections were obtained from NuDat 2.7 [5] and QCalc [6] and summarized in Table 1.
Table 1. Reaction and decay data for 169Yb production
NuclideHalf-life

Decay mode (%)

E (keV)I (%)

Contributing reaction

Q-value (MeV)

169Yb

32.018 d

EC (100)

177.2122.28(11)169Tm(p,n)-1.7

3. Results
The 177.21-keV gamma rays (I = 22.28%) emitted with the 169Yb decay (T1/2 = 32.018 d) were
measured after a cooling time of about 4 days. The cross sections of the 169Tm(p,n)169Yb reaction were
derived from the measurements and the nuclear data in Table 1. The preliminary result is shown in Fig. 1 incomparison with the previous experimental data [7–9] and the TENDL-2017 data [10]. The peak amplitudeof the data of Birattari et al. (1973) [7] is almost consistent with our result. However, the data of Spahn et al.
(2005) [8] shows two times larger amplitude than ours. The peak of the TENDL-2017 data shows the similar
amplitude with ours, however smaller width and lower energy position than ours.

4. Summary
We measured activation cross sections of the 169Tm(p,n)169Yb reaction up to 18 MeV. The experiment
was performed using the cyclotron at ATOMKI. The well-established methods, the stacked-foil activation
method and the gamma-ray spectrometry, were adopted. The preliminary result of the cross sections wascompared with previous experimental data and theoretical model calculation. The further analysis is being
performed.

Fig. 1.

Excitation function of the 169Tm(p,n)169Yb reaction with the previous data [7–9] andthe TENDL-2017 data [10].

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JAEA-Conf
JAEA-Conf 2020-001

Acknowledgements
This work was carried out at Institute for Nuclear Research (ATOMKI), Hungary. This work was partly
supported by “Optimization of accelerator production routes of the new theranostic radioisotopes Sc-47 andCu-67 (FY2019-2020)”, under the Japan - Hungary Research Cooperative Program, JSPS and HAS and JSPS
KAKENHI Grant Number 17K07004.

References[1]

Leonard KL, DiPetrillo TA, Munro JJ, et al., A novel ytterbium-169 brachytherapy source anddelivery system for use in conjunction with minimally invasive wedge resection of early-stage lungcancer, Brachytherapy, 2011, 10, 163–169.

[2]

Saito M, Aikawa M, Komori Y, et al., Production cross sections of 169Yb and Tm isotopes in deuteroninduced reactions on 169Tm, Appl. Radiat. Isot. [Internet], 2017, 125, 23–26, Available from:http://dx.doi.org/10.1016/j.apradiso.2017.04.010 (accessed 2020/03/24).

[3]

Saito M, Aikawa M, Sakaguchi M, et al., Production cross sections of ytterbium and thulium
radioisotopes in alpha-induced nuclear reactions on natural erbium, Appl. Radiat. Isot. [Internet],
2019, 154, 108874, Available from: https://doi.org/10.1016/j.apradiso.2019.108874 (accessed
2020/03/24).

[4]

Ziegler JF, Biersack JP, Ziegler MD, SRIM: the Stopping and Range of Ions in Matter [Internet],
2008, Available from: http://www.srim.org (accessed 2020/03/24).

[5]

National Nuclear Data Center, Nuclear structure and decay data on-line library, Nudat 2.7 [Internet],2017, Available from: http://www.nndc.bnl.gov/nudat2/ (accessed 2020/03/24).

[6]

Pritychenko B, Sonzogni A, Q-value Calculator (QCalc) [Internet], 2003, Available from:http://www.nndc.bnl.gov/qcalc/ (accessed 2020/03/24).

[7]

Birattari C, Gadioli E, Gadioli Erba E, et al., PRE-EQUILIBRIUM PROCESSES IN (p,n)
REACTIONS, Nucl. Phys. A, 1973, 201, 579–592.

[8]

Spahn I, Takács S, Shubin YN, et al., Cross-section measurement of the 169Tm(p,n) reaction for the
production of the therapeutic radionuclide 169Yb and comparison with its reactor-based generation,
Appl. Radiat. Isot., 2005, 63, 235–239.

[9]

Tárkányi F, Hermanne A, Takács S, et al., Activation cross-sections of proton induced nuclear
reactions on thulium in the 20-45MeV energy range, Appl. Radiat. Isot., 2012, 70, 309–314.

[10]

Koning AJ, Rochman D, Sublet J-C, et al., TENDL: Complete Nuclear Data Library for Innovative
Nuclear Science and Technology, Nucl. Data Sheets, 2019, 155, 1–55.

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JAEA-Conf 2020-001

17

Measurements of nuclear cross section data for proton-induced
reactions on Bi and Pb targets at intermediate energy

M. BAKHTIARI1, L. Mokhtari ORANJ2, N. S. JUNG2, A. LEE2, H. S LEE1,2*
1

Division of Advanced Nuclear Engineering (DANE), POSTECH, Korea
2

Pohang Accelerator Laboratory (PAL), POSTECH, Korea* Email: lee@postech.ac.kr
Nuclear cross sections data of 209Bi(p, xn)207,206,205,204,203Po, 209Bi(p, pxn)207,206,205,204,203,202Bi,
and natPb(p, xn)206,205,204,203,202,201Bi reactions were measured using the stack-foil activation
technique and gamma-ray spectrometry over the 40 to 100 MeV proton energy range. The
targets were arranged in different stacks consisting of Bi, Pb, Al, Au foils and Pb plates. The
proton beam intensity was determined by activation analysis method using 27Al and 197Au foils.
Furthermore, excitation functions of the mentioned reactions were calculated by using thetheoretical models based on the TALYS code and compared to the measured data in this work
as well as with other data in the literature. Combinations of the nuclear input parameters ofdifferent level density models, optical model potentials, and γ-ray strength functions wereconsidered. It was concluded that the level density models influenced the reaction cross sections
significantly.

I. Introduction
Production cross sections of radionuclide via proton-induced reactions play an importantrole in different applied physics fields [1]. A literature review showed that there is a lack of the
production cross sections on bismuth and lead in the energy range of 40 to 100 MeV. In a seriesof the experiments, the production cross sections of 209Bi(p, xn)207,206,205,204,203Po, 209Bi(p,pxn)207,206,205,204,203,202Bi, and natPb(p, xn) 206,205,204,203,202,201Bi were measured [2–4]. Here a partof the measured data are reported. In addition, theoretical calculations were performed using
the TALYS code [5] to understand the capability of the nuclear model calculations in thereaction cross sections estimation.
II. Methods
A. Experiment
The cross sections of the proton-induced reactions on Bi and Pb were measured in four
rounds of experiment. The targets were arranged in stacks and were irradiated with 100-MeV
protons (two rounds) and 69-MeV protons (two rounds). Five 209Bi foils (purity: 99.97%,
thickness: 50 µm, density: 9.8 gcm-3) were stacked and irradiated by 100-MeV (irradiation time:
120 s) and 69-MeV (irradiation time: 240 s and 25 s) protons. In addition, five natPb foils (purity:
99.99%, thickness: 50 µm, density: 11.35 gcm-3) were used as target and irradiated by 100-MeV
protons for 204 s. Au (purity: 99.95%, thickness: 30 µm, density: 19.3 gcm-3) and Al (purity:- 83 -

JAEA-Conf 2020-001

99.999%, thickness: 100 µm, density: 2.69 gcm-3) foils were stacked in the target to monitor
the proton beam intensities via 197Au(p, x)196,194Au and 27Al(p, 3pn)24Na reactions, respectively.
Figure 1 shows the schematic view of the target stack. Proton energy distribution on each foil
was determined using the Monte Carlo code FLUKA [6]. The experiments were conducted at
Korea Multi-purpose Accelerator Complex (KOMAC). The detail information of theexperimental set-up is explained in Refs. [2–4].

Fig 1. Schematic view of the target stack.
B. Data analysis
Measurement of γ-ray spectra on each activation foil was performed using a high-resolution
HPGe detector. The intense characteristic γ rays of each radionuclide were selected to identify
the interested radionuclide [3]. The distance between the detector and activation foils was
decided to keep the dead time low enough. The decay data of all radionuclei were taken from
Ref. [7].
The production cross sections of the investigated radionuclei were determined by thefollowing equation in case, the radionuclide is generated via only nuclear reaction of (p, xn):
𝜎𝜎 �

���
���
� � � ���� ���� ���� ��� �� �� �����

(1),

where Np is net area of a full energy peak related to γ-ray transition of a nuclide, λ is the decay
constant of the nuclide (s-1), εγ and Iγ are the efficiency of detector and absolute γ-ray intensity,
respectively. The tm and tc denote the cooling time and measuring time (s), respectively. N0 is
the number of target nuclei (atoms) and Φ is the proton flux (protons/cm2/s). A radionuclide
could be generated by a nuclear reaction as well as by the decay of a mother radionuclide. In
such a case, the contribution of the precursor needs to be taken into account and the production
cross section of the daughter radionuclide is determined using Eq. (2) for (p, pxn) reactions:𝜎𝜎� ��� ��� ����� � ��� � �� � ����� �𝜎𝜎�

���� ���

��

�1 � ����� � ��

�,where the factor Fk (k = m, d) is related to the time condition as follows:- 84 -(2)

JAEA-Conf 2020-001

𝐹𝐹� � �1 � � ����� �� ����� �1 � � ����� �,(3)

where σd and σm are the cross sections for formation of daughter and mother radionuclei (cm2),
λd and λm are the decay constant of daughter and mother radionuclei, respectively. P is thebranching fraction of mother nuclei decay to daughter, ti is the irradiation time (s). The resultsof 209Bi(p, xn)206,205Po, 209Bi(p, pxn)204Bi and natPb(p, xn)204Bi are presented.
C. Nuclear model calculations
Reaction cross sections of the assessed radionuclei were calculated using the nuclear model
calculations based on the TALYS [5] code and were compared to the measured experimental
data. TALYS is a nuclear model code that uses protons, neutrons, photons, 3He and 4He as
projectiles in the energy range of 1 keV to 200 MeV. The calculations were performed using
different nuclear ingredients to consider their effects on the reaction cross sections. Nuclear
level densities (NLDs), optical model potentials (OMPs) and γ-ray strength functions (γSFs)
with all 768 possible combinations were used in the calculations. The detail information of themodels and their definitions can be found in Refs. [2,5]. In the figures, the shaded areas indicate
the variation of the cross sections by using different combinations.

III. Results206

Po (T1/2 = 8.8 d; EC, 94.55%; and α, 5.45%) can be produced through 209Bi(p, xn)206Po
reaction. The activities of this radionuclide were measured using 286.4-, 522.5-, 807.4- and980.2-keV γ rays. The measured cross sections are shown in Fig. 2(a) and are compared withother experimental data together with theoretical calculations. Generally, our measured data arein agreement with other experimental data in the literature. As it is seen from the figure,experimental and theoretical results show two different peaks at 40 MeV. The TALYS
calculations using different model combinations showed that by changing the type of the NLD,
the cross section values are changed significantly. TALYS calculations predict the reaction
threshold well.
The 205Po (T1/2 = 1.74 h; EC, 99.96%; and α, 0.04%) production cross sections were measured
using 836.8-, 849.8- and 872.4-keV γ rays in the energy range of 40 to 100 MeV and are shownin Fig. 2(b). Measured cross sections are on a similar trend to experimental data in the literature.
The TALYS calculation predicts the maximum cross section of 890 mb at 45 MeV, while themeasured cross section is 875.8 mb at 51 MeV. It seems that TALYS is indicating a lower energy
threshold than the experimental data.
204
Bi has a ground state with a half-life of 11.22 h and two short-lived metastable states withhalf-lives of 13 and 1.07 ms which decay to the ground state with branching ratios of 100%.
The 209Bi(p, p5n)204Bi reaction and 204Po (T1/2 = 3.519 h) decay could lead to the production of204
Bi. The production yield of 204Bi was measured using 374.76-keV γ ray in the spectrum. The
measured independent cross sections are illustrated in Fig. 2(c). Our measured data in two are
newly measured cross sections. Nuclear model calculations predict the reaction thresholdenergy well. Theoretical calculations vary significantly with level density models, indicating- 85 -

JAEA-Conf 2020-001

the effectiveness of the NLDs on the reaction cross sections. The reaction cross sections did notchange significantly by changing the type of OMP and γSF [2].
204
Bi production cross sections on natPb target were also measured and compared to themeasured data in the literature as well as the TALYS calculations as shown in Fig. 2(d).
Measured cross sections in this work are in agreement with other experimental data. TALYS
default calculations seem to predict the maximum measured cross sections quite well. On theother hand, microscopic NLDs show lower cross sections values. The reaction cross sections
did not change significantly by changing the type of OMP and γSF. All nuclear model
calculations underestimate the experimental data after 60 MeV.

209

Bi(p,4n)1400

This work (2019)
This work (2017)
Chung et al. (2011)
Michel et al. (2002)
Kuhnhenn (2001)
Titarenko et al. (1998)
Zongyu et al. (1990)
Ward et al. (1981)
Miyano et al. (1974)
Birattari et al. (1971)
Belyaev et al. (1963)
Bell and Skarsgard (1956)
TALYS-1.8-Default
KD-Avrigeanu OMP-GSM-RMF
JLM-OMPIII-HFB-GHFB
JLM-Avrigeanu OMP-HFB-GHFB

206

Po

Cross section [mb]

1200
1000
800
600
400
200
0

1200

Bi(p,5n)205Po

1000

800

600

400

200

0
10

30

50

70

90

110

130

150

30

50

70

110

130

150

600

This work (2019)
This work (2017)

Bi(p,p5n)204Bi

250

nat

Pb(p,xn)204Bi

Belyaev et al. (1963)
TALYS-1.8-Default
KD-Avrigeanu OMP-CTM-GTHFB
JLM-Avrigeanu OMP-MHF-RMF
KD-Avrigeanu OMP-GSM-GHFB
KD-OMPIII-HCM-RMF
JLM-Nolte OMP-HFB-GHFB

200

150

100

This work (2017)
Ditroi et al. (2014)
Titarenko (2009)
Kuhnhenn et al. (2001)
Gloris et al. (2001)
TALYS-1.8-Default
JLM-Avrigeanu OMP-MHF-KU
KD-OMPIII-HCM-RMF
JLM-Nolte OMP-HFB-GHFB

500

Cross section [mb]209

Cross section [mb]90

Proton Energy [MeV]Proton Energy [MeV]
300

400

300

200

100

50

0

This work (2019)
This work (2017)
Chung et al. (2011)
Titarenko et al. (1998)
Belyaev et al. (1963)
Bell and Skarsgard (1956)
TALYS-1.8-Default
KD-Avrigeanu OMPII-GSM-RMF
JLM-OMPII-HCM-HFBQ
JLM-Nolte OMP-HFB-GHFB

209

Cross section [mb]

1600

0

30

50

70

90

110

130

150

Proton Energy [MeV]

0

20

40

60

80

100

120

140

160

Proton Energy [MeV]Fig 2. Measured independent cross sections for the (a) 209Bi(p,4n)206Po, and (b) 209Bi(p,5n)205Po,(c) 209Bi(p,p5n)204Bi, and (d) natPb(p,xn)204Bi reactions compared with the previously published data
together with theoretical calculations using the TALYS code. Our measured data are published inRefs. [2,3]. Other experimental data are taken from EXFOR [8].

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JAEA-Conf 2020-001

IV. Conclusions
Production cross sections of proton-induced reactions on 209Bi and natPb were measured using
over the energy range of 40 to 100 MeV. The results of 209Bi(p, xn)206,205Po, 209Bi(p, pxn)204Bi
and natPb(p, xn)204Bi are presented. Our measured data were in agreement with the otherexperimental data reported in the literature. Theoretical calculations were performed using the
TALYS code by applying different nuclear ingredients of NLDs, OMPs and γSFs and werecompared to the theoretical calculations. It was concluded that the most effective models were
NLDs rather than OMPs and γSFs for the investigated nuclear reactions between 40 and
100 MeV.
Acknowledgement
The authors thank the KOMAC team for their support during the experiment. This work is
supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear
Safety (KOFONS), granted financial resources from the Nuclear Safety and Security
Commission (NSSC) of Republic of Korea (Grants No.1303026 and No. 1603005).
References[1]

Gloris M, Michel R, Herpers U, Sudbrock F, Filges D. Production of residual nuclei from
irradiation of thin Pb-targets with protons up to 1.6 GeV. Nucl. Instruments Methods
Phys. Res. Sect. B Beam Interact. with Mater. Atoms. 1996;113:429–433.

[2]

Mokhtari Oranj L, Jung NS, Bakhtiari M, Lee A, Lee HS. Cross sections of protoninduced nuclear reactions on bismuth and lead up to 100 MeV. Phys. Rev. C. 2017;95:1–
15.

[3]

Mokhtari Oranj L, Bakhtiari M, Jung NS, Lee A, Lee HS. Extension of excitation
functions of proton-induced reactions on bismuth. Phys. Rev. C. 2020;101:14602.

[4]

Mokhtari Oranj L, Bakhtiari M, Kye YU, Jung NS, Lee A, Lee HS. Benchmarking
FLUKA, PHITS, MCNPX, and MARS15 codes with product yields of 209Bi(p, x)
reactions. Nucl. Instruments Methods Phys. Res. Sect. B Beam Interact. with Mater.
Atoms. 2020;462:154–162.

[5]

Koning AJ, Hilaire S, Duijvestijn MC. TALYS-1.0. ND2007. Les Ulis, France: EDP
Sciences; 2007. Available from: http://www.talys.eu/ (accessed 2020/03/24).

[6]

Fasso A, Ferrari A, Ranft J, Sala PR. FLUKA: a multi-particle transport code [Internet].
CERN-2005-10, INFN/TC_05/11, SLAC-R-773; 2005. Available from:http://www.fluka.org/ (accessed 2020/03/24).

[7]

ENSDF database [Internet]. Available from: http://www.nndc.bnl.gov/ensarchivals/(accessed 2020/03/24).

[8]

Zerkin VV, Pritychenko B. The experimental nuclear reaction data (EXFOR): Extended
computer database and Web retrieval system. Nucl. Instruments Methods Phys. Res. Sect.
A Accel. Spectrometers, Detect. Assoc. Equip. 2018;888:31–43.

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JAEA-Conf 2020-001

18 Measurement of nuclide production cross sections for Z = 26-30 elements
irradiated with 0.4 - 3.0 GeV proton in J-PARC

Hiroki MATSUDA1, Hayato TAKESHITA2, Hiroki IWAMOTO1, and Shin-ichiro MEIGO1
1
2

J-PARC Center, Japan Atomic Energy Agency, 2-4 Tokai-mura, Ibaraki 319-1195, Japan

Department of Advanced Energy Engineering Science, Kyushu University, Kasuga, Fukuoka 816-8580, Japanemail: matsuda.hiroki@jaea.go.jp
For the design of Accelerator-Driven nuclear transmutation System (ADS), accurate cross section
data are required to estimate the number of radioactive nuclides for treating radioactive wastes in
the ADS plant. Although much effort has been devoted to obtaining the nuclide production crosssection irradiated by protons in the energy region utilized for the ADS at several facilities so far,data with systematic uncertainties of ~10% exist. Furthermore, no experimental data still exist
around for protons in a few GeV region. In order to validate the calculation code and the evaluateddata utilized for the ADS, we started the experiment to obtain the cross section at 3-GeV
synchrotron in J-PARC. In this study, we obtained nuclide production cross section for targets with
Z numbers from 26 to 30 (i.e., natFe, natNi, and natZn), which are important regions for the estimationof the radioactivity of the ADS structural materials. Furthermore, we investigated the incident
energy dependency on the cross section from several hundred MeV to 3 GeV using the present andthe previous experimental data, comparing with different intra-nuclear cascade and evaporation
models of INCL4.6/GEM, INCL++/ABLA07, and Bertini/GEM, and evaluated nuclear data of
JENDL-HE2007.

1. Introduction
Accelerator-Driven nuclear transmutation System (ADS) transmutes minor actinides (MA) by
supplying neutrons continuously. Neutrons are supplied by a spallation reaction of lead-bismuth
eutectic (LBE) irradiated by 1.5 GeV energy protons, which is also utilized as coolant. For theestimation of the radioactive nuclides to treat wastes and evaluate the chemical effect of spallation
products in the LBE, accurate cross section data are required. Although much effort has been
devoted to conducting nuclide production cross section measurements at several facilities since the1950s, the uncertainties of data, being typically about 10%, are not good enough to validate thecalculation model. Furthermore, data around in a few GeV region, which are candidates for
projectile energies of the ADS, have larger uncertainty than other regions. Thus, the experiment
was performed in J-PARC. In this study, we obtained nuclide production cross section for targets
with Z numbers from 26 to 30 (i.e., natFe, natNi, and natZn), which are important regions for theestimation of the radioactivity of the ADS structural materials. They were irradiated by proton- 89 -

JAEA-Conf 2020-001

beams having different energies of 0.4, 1.3, 2.2, and 3.0 GeV. A comparison between theexperimental data and the calculations were demonstrated. In this paper, the details of theexperiment and analysis procedure for Zn target are described. The results of Fe and Ti have been
already reported [1,2].
2. Experiment
2.1 Setup
The basic experiment setup was the same as Ref. [3]. The experiment was carried out at thebeam transport line from the RCS [4] to the Materials and Life Science Experimental Facility(MLF), which is called as 3NBT. Thin square foils of Fe, Ni, and Zn, 0.1 mm thick and 25mm long,
were sandwiched with thin aluminum foil with 0.1mm thick to evaluate the number of recoil
nuclides. Four sets of this sample were placed at the head of each linear stage guide. They wereplaced in the vacuum chamber that was installed in the beam dump line. A linear stage guide wascontrolled remotely to insert and extract samples.
2.2 Irradiation
Each sample set was irradiated by 0.4, 1.3, 2.2, and 3.0 GeV protons, respectively. The 0.4 GeV
beam was transported from the LINAC without acceleration. The other energies were obtained bychanging the extraction timing of the kicker magnet at RCS. The beam width and position were
measured with the multi-wire profile monitor (MWPM) installed in the beamline. The number ofprotons in the beam was monitored by the current transformer (CT). Beam profile measurement
using an imaging plate (IP) was performed after irradiation to improve the accuracy of the beam
position on the samples. Repetition of shots was set to 0.4 Hz to avoid melting of samples. In total,
40 shots, i.e., 2.32 � 10�� protons, were irradiated for each target. After irradiation, the linear stage
guides were removed from the chamber. The cooling time, which is the time interval between theend of the irradiation to the beginning of the measurement, was approximately six to nine hours.
2.3 Analysis
Decay gamma-rays from irradiated samples were measured by the high pure germanium
detector (HPGe, Mirion Technologies GC2018). The list of activation products analyzed in thispaper was summarized in Table 1.
The samples were mounted on the acrylic spacer apart from the head of HPGe by about 5 cm
or 25 cm to keep detector-to-sample geometry rigidly. The production cross section of specified
nuclei 𝜎𝜎 was written as

𝜎𝜎 �

�

�� ��� ������ ��� ����� ��������� �
�� �

… (1)

where 𝐶𝐶 is the number of count of a peak, 𝑡𝑡� is so-called live-time which includes deadtime- 90 -

JAEA-Conf 2020-001

correction during measurement, 𝜖𝜖 is the efficiency at the peak energy, 𝐼𝐼� is the absolute gamma-

ray intensity, 𝜇𝜇 is the self-absorption correction, 𝑡𝑡� is so-called real-time, 𝜆𝜆 is the decay constantof the nuclide, 𝑁𝑁� is the number of protons irradiated, and 𝑛𝑛 is the number density of the sample.

Table 1 The list of nuclides analyzed.
NuclideHalf-life

Here the correction during irradiation was not

7Be

53.22 dapplied since the irradiation time was considerably

22Na

2.6018 yshorter than the half-life of products. The recoil

48Cr

21.56 hparticles could be trapped by the forward and

54Mn

312.20 d

backward aluminum foils. They were observed in

52Fe

8.275 hthe actual aluminum foil. However, the activity of

55Co

17.53 hthem was evaluated as a negligible amount. Thus,

58Co

70.86 d57Ni

35.60 h61Cu

3.389 h64Cu12.701 h62Zn

9.193 h65Zn

243.93 dthis correction was not applied in this experiment.
The effects of secondary particles affecting the
backward samples were also estimated by thesimulation. Since very thin samples were employedin this experiment, the number of secondary
particles was lower than the irradiated protons in
the order of three or four digits. Hence, this effectwas not considered in this analysis. The number density of the sample 𝑛𝑛 is derived by

𝑛𝑛 �

�

��

𝑁𝑁� … (2)where 𝑚𝑚 is the sample mass that is measured by the electric scale, 𝑆𝑆 is the surface area of thesample, 𝑀𝑀 is the atomic weight of the sample, and 𝑁𝑁� is Avogadro constant. The self-absorption
correction 𝜇𝜇 is calculated by𝜇𝜇 ��������att ����att ��

… (3)where 𝜇𝜇att is the attenuation coefficient, 𝜌𝜌 is the sample density, and 𝑡𝑡 is the thickness of thesample. Detector efficiency of the HPGe was evaluated by using 241Am, 152Eu, 60Co, and 137Cs

standard gamma-ray sources, which were put at the same position as the samples. Sum-peakcorrections for multi gamma-ray sources were applied [5].

The number of protons and beam profile (width and position) were measured precisely by the
CT and the MWPM, respectively. Beam profile measurement using an IP (FUJI-FILM BASSR2040) was also performed to improve the accuracy of the beam profile. The irradiated samples
were placed onto the IP to expose. After exposure, the IP was scanned by the image processor (GEhealthcare Typhoon FLA 7000). Scanned images were fit by a two-dimensional Gaussian toevaluate the width and position of the beam. Here the Gaussian shape was guaranteed by themeasurement of the MWPM. Finally, the fractions of the proton beam on the sample foils were

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JAEA-Conf 2020-001

derived by integrating the function over the sample area. As a result, the fractions ranged from
0.98 to 1.00. This fraction was used for the final correction.
The sample surface area was measured by a typical image scanner (EPSON GT-S650).
2.4 Uncertainty estimation

Table 2 Table of uncertainties.
Description

Estimated uncertainty %Statistics of gamma-ray count<0.1 – 10

Detector efficiency

1.0

Sample weight<0.1

Sample surface area

1.4

Time

<0.1

Self-absorption

0.1

The number of protons

1

Imaging plate<0.1

Branching ratio<0.1 – 30

In total

2.0 – 31.7

The

list

of

uncertainties

concerned is shown in Table 2.
The uncertainty of the detector
efficiency

was

the

standard

deviation of differences between
measured points and calculated
ones. Imaging plate uncertainty,
namelytheuncertaintyofirradiated proton fraction, wasevaluatedasthestandard

deviation of proton fraction of all
samples at the same energy. The

reading uncertainty of scanning of 1mm scale with 800 dots/inch, i.e., 314.96 dots/mm, was 1 dot.
In addition, an ambiguity for an edge detection could be estimated as 2 dots. Thus, the uncertaintyof the sample surface was evaluated as 1.4% by a square root of the sum of the square of 3/314.96.
3. Results and observations
The PHITS code [6] was employed to calculate the production cross sections with two
intranuclear cascade models (Bertini and INCL4.6 [7]) and generalized evaporation model(GEM [8]). The calculation by using the INCL++ code [9] with INCL++v6.0.1 and ABLA07 [10]models was also performed. The number of histories was sufficiently large, so that calculation
uncertainties were negligible. In Fig. 1, the measured, the calculated, and the evaluated crosssections of JENDL-HE/2007 [11] for each production cross sections of zinc were shown. The otherexperimental data were taken from EXFOR [12].

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JAEA-Conf 2020-001

Figure 1 Production cross sections of 7Be, 22Na, 48Cr, 54Mn, 52Fe, 55Co, 58Co, 57Ni, 61Cu, 64Cu, 62Zn,
and 65Zn from natZn. This works (red filled circle), other experiments (colored other symbols) takenfrom EXFOR, calculations (colored solid lines), and the evaluated data (magenta solid line) areplotted.
For all production cross sections shown in this paper, PHITS calculations agreed with theexperimental data. In the light nuclide productions such as 7Be and 22Na, Bertini model gave better
agreement than INCL4.6. INCL4.6 calculations agreed with heavier nuclide productions, i.e., lessthan 10 nucleons emission reactions. In contrast, INCL++ code overestimated or underestimatedfor almost all productions. This tendency was also observed in the cross section with Fe and Ni
targets [1,2]. JENDL-HE/2007 agreed with the experimental data for all cross sections though it
showed strange behavior below 150 – 250 MeV, in which the evaluation method was changed fromthe GNASH code [13] to a microscopic simulation code (either JAM [14] or JQMD [15]).
4. Summary
The production cross sections from Zn with 0.4, 1.3, 2.2, and 3.0 GeV protons were measured
at J-PARC. In total, nearly 50 cross sections were obtained with higher accuracy than otherexperimental data in the past thanks to quite stable J-PARC beam and monitoring systems. The
present results were compared with the calculations and the evaluated nuclear data. The present- 93 -

JAEA-Conf 2020-001

experimental data suggested that further improvement of models is mandatory in the GeV energy
region. The evaluated data was in good agreement with the experimental data.
Acknowledgment
The authors are grateful to Mrs. Nishikawa and Hirano, and the members of Nippon Advanced
Technology (NAT) for technical assistance with the experiments and to Drs. Saha, Hotchi, andHarada for tuning of the RCS.
References[1] H. Matsuda, et al., Atomic Energy Society of Japan, 2019 Spring Meeting, Mar. 2019 (inJapanese).
[2] H. Takeshita, et al., Atomic Energy Society of Japan, 2019 Fall Meeting, Sep. 2019 (inJapanese).
[3] H. Matsuda, S. Meigo, H. Iwamoto, J. Nucl. Sci. Technol. 55, 8 (2018).
[4] S. Meigo, F. Noda, S. Ishikura, et al., Nucl. Instrum. Meth. A 562, 569 (2006).
[5] E. Tomarchio, S. Rizzo, Radiat. Phys. Chem. 80, 318 (2011).
[6] T. Sato, Y. Iwamoto, S. Hashimoto, et al., J. Nucl. Sci. Technol. 55, 6 (2018).
[7] A. Boudard, J. Cugnon, J.C. David, et al., Phys. Rev. C 87, 014606 (2013).
[8] S. Furihata, Nucl. Instrum. Meth. B 171, 3 (2000).
[9] D. Mancusi, A. Boudard, J. Cugnon, et al., Phys. Rev. C 90, 054602 (2014).
[10] A. Kelic, M.V. Ricciardi, K.H. Schmidt (2009), arxiv:0906.4193.
[11] Y. Watanabe, K. Kosako, S. Kunieda, et al., J. Korean Phys. Soc. 59, 1040 (2011).
[12] V. Zerkin, B. Pritychenko, Nucl. Instrum. Meth. A 888, 31 (2018).
[13] P. G. Young, E. D. Arthur, M. B. Chadwick, LA-12343-MS, LANL (1992).
[14] Y. Nara, N. Otuka, A. Ohnishi, et al., Phys. Rev. C 61, 024901 (1999).
[15] K. Niita, S. Chiba, T. Maruyama, et al., Phys. Rev. C 52, 2620 (1995).

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JAEA-Conf 2020-001

19 Study of an angular correlation of γ-rays emitted by
117
Sn(n,γ) reactions for T-violation search
Jun Koga1 , Shunsuke Endo2 , Hiroyuki Fujioka3 , Katsuya Hirota4 , Kohei Ishizaki4 ,Atsushi Kimura2 , Masaaki Kitaguchi4 ,So Makise1 , Yudai Niinomi4 , Takuya Okudaira2 , Kenji Sakai2 , Tatsushi Shima5 , Hirohiko
M. Shimizu4 , Sshusuke Takada1 , Yuika Tani3 , Tomoki Yamamoto4 , Hiromoto Yoshikawa5 ,and Tamaki Yoshioka1
1

Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
Japan Atomic Energy Agency, 2-4 Shirakata, Tokai 319-1195, Japan
3
Tokyo Institute of Techonology, 2-12-1 Ookayama Meguro-ku, Tokyo 152-8550, Japan
4
Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8601, Japan
5
Research Center for Nuclear Physics, Osaka University, 10-1 Mihogaoka, Ibaraki 567-0047, Japan2

E-mail: j.koga@epp.phys.kyushu-u.ac.jp
A discovery of violation of time-reversal symmetry (T-violation) can lead to a solution for an asymmetry between matter and antimatter in the current universe, which is one of the important problems
in particle physics and astrophysics. An enhancement of T-violation is theoretically proposed in several compound nuclear reactions. The experimental sensitivity to ﬁnd a T-violating eﬀect depends ona spin factor κ(J) which is diﬀerent from each nuclide. It can be determined from an angular dependence of a diﬀerential cross section of neutron-nucleus reaction, and the experiments to measure thisangular correlation have conducted at J-PARC. In this paper, the measurement result and the analysis
status of experiments using the target nucleus 117 Sn are reported.

1. Introduction
An asymmetry between the number of matter and antimatter in the current universe is one of theimportant problems which should be solved in particle physics and astrophysics. A. Sakharov proposedthat violation of charge conjugation and parity symmetry (CP-violation) stronger than expected within
the Standard Model of particle physics is necessary to explain this asymmetry [1].
Compound nuclear reactions are expected as one of CP-violation search beyond the Standard Model
under the assumption that the CPT theorem, which means that CP-violation is equal to the violation oftime-reversal symmetry (T-violation), is correct. In several compound nuclear reactions, the violation ofparity symmetry (P-violation) is observed with an enhancement factor of 106 compared to the protonproton scattering [2]. This enhancement is theoretically explained as a sp-mixing model describing aninterference between amplitudes of s-wave resonances and p-wave resonances. It is theoretically suggested that T-violation is enhanced by a similar mechanism in several nuclear reactions [3]. In addition,
this theory proposes that an experimental sensitivity depends on each nuclides. One of the key parameters allowing to become good candidate nuclei is a spin factor κ(J). Thus, there is a possibility that themagnitude of T-violating eﬀect is enhanced if the value of κ(J) is not small. The κ(J) is related to neutron
resonance partial widths via a mixing angle φ describing a superposition of diﬀerent spin components. So
far only 139 La has been determined the value of κ(J) by Okudaira et al. [4]. To identify further candidates
for T-violation search, other nuclei muse be measured.
The mixing angle φ can be determined by measuring an angular correlation of prompt γ-rays emitted
from compound nucleus with respect to a direction of incident neutrons. According to Flambaum [5],- 95 -

JAEA-Conf 2020-001

the diﬀerential cross section for unpolarized neutrons is described as follows:



dσ 1
1
2
=
a0 + a1 cos θγ + a3 cos θγ −
,
dΩ 2
3

(1)

where θγ is an angle between the ﬂight directions of the emitted γ-rays and the incident neutrons. The a0
term corresponds to the Briet-Wigner formula, while the a1 and a3 terms include the mixing angle φ [5].
Equation (1) indicates the shape of p-wave resonance depends on the direction of emitted γ-rays withrespect to the incident neutrons due to their terms. We can verify the sp-mixing model by measuring thisangular dependence of p-wave resonance.
The isotope 117 Sn is one of the candidates for T-violation search. In this paper, the results of measurements of angular dependence of γ-ray emission around the 1.33 eV p-wave resonance, resonanceparameters, and branching ratio of each resonance, which are essential values to calculate precisely theEq. (1), is reported.

2. Experiment
2.1 Experimental setup
Our experiments were carried out with Accurate Neutron-Nucleus Measurement Instrument (ANNRI) in the Material and Life science experimental Facility (MLF) of the Japan Proton Accelerator
Research Complex (J-PARC). ANNRI is installed at the neutron beam line BL04 to accept the pulsed
neutron beam from spallation source of J-PARC MLF. Produced neutron beam was moderated by a liquid hydrogen moderator and supplied to the beam line at a repetition rate 25 Hz. The position of nuclear
target is placed at 21.5 m from the moderator surface. Lead plates (in total 37.5 mm thickness) were
installed at upstream to suppress background events stemmed from fast neutrons and γ-rays produced byspallation reactions. The disk choppers were operated synchronously with the proton-beam injection to
avoid frame overlap due to low-energy neutrons.
ANNRI has a germanium detector assembly shown in Fig. 1. It consists of two types of detector
units: type-A and type-B. The type-A is an assembly of 7 germanium crystals, while the type-B has
a germanium crystal, and it has 22 crystals in total. Each crystal is installed at 36.0, 70.9, 72.0, 90.0,
108.0, 109.1, and 144.0 degrees with respect to the neutron beam direction, respectively. This detector
assembly enables us to measure the deposit energy and the detection time of γ-rays in each germanium
crystal. When we focus on the prompt γ-rays from the nucleus target by (n,γ) reaction, the detection
time can be regarded as time-of-ﬂight (TOF) of neutrons from moderator surface to the target position
because the time diﬀerence between neutron capture and the emission of a prompt γ-ray is negligible.
The incident neutron energy is calculated from the TOF of the neutrons.
2.2 Measured data set
We have conducted three experiments to measure (i) the angular dependence of the shape of p-wave
resonance, (ii) resonance parameters, and (iii) branching ratio of each resonance. Here, the branchingratio means the transition ratio that the compound state decays to a ﬁnal state of a speciﬁc energy level,
which corresponds to the partial gamma width. Table I shows that each experimental condition. Target
and measurement time are diﬀerent in order to achieve each purpose.
Table I. Each experimental condition. Target and measurement time are diﬀerent for each purpose. Proton beam
power also depends on the date conducted in each experiment.
PurposeAngular dependence
Resonance paramter
Branching ratio

TargetnatSn117
Sn (86% enrich)
nat
Sn

Size
40×40×4 mm3
φ5 mm×6 mm
40×40×1 mm3- 96 -

Beam power
150 kW
525 kW
525 kW

Measurement time
65 hours
6 hours
100 hours

JAEA-Conf 2020-001

Fig. 1.

Conﬁguration of the germanium detector assembly in ANNRI.

Counts [A.U.]

Figure 2 shows the spectrum of the energy deposit of γ-rays in all germanium detectors in the experiment (i). The γ-transition in the nucleus 118 Sn with an energy of 9327 keV and its single- and doubleescape peaks can be observed clearly. It is known that the compound state in the p-wave resonance decays
to the ground state of 118 Sn directly. Therefore, we focused the peak with 9327 keV and its single- anddouble-escape peaks for our analysis to examine the angular dependence of the shape of p-wave resonance and to determine the branching ratio that the compound state of 117 Sn + n system decays to theground state of 118 Sn.

105 



 

104

10

3

8000

8500

9000

9500
10000
Gamma-ray Energy [keV]

Fig. 2. Gamma-ray spectrum by neutron capture reactions in the experiment (i) at the range of 8 - 10 MeV. The
peak with an energy deposit of 9327 keV and its single- and double-escape peaks due to 117 Sn(n,γ) reactions in thetarget can be observed clearly.

3. Analysis
3.1 Background subtraction and beam intensity normalization
The neutron energy spectrum gated with events from signal regions (9327 keV, its single- and doubleescape peaks) includes background events from other sources. There are two kinds of background events.

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JAEA-Conf 2020-001

One is caused by the Compoton scattering of the 9563 keV γ-rays emitted from the compound state of
116 Sn. The other stems from pileup events due to simultaneous detection of multi γ-rays. These background events must be subtracted. First, the number of background events in the signal regions wasestimated using a GEANT4 simulation [6], which enables us to obtain a spectrum convoluted a response
function of the germanium detector for monoenergetic γ-rays [7]. Spectra gated with the backgroundregions (9563 keV peak and energies higher than 9600 keV) were scaled that number of events matchedthat of GEANT4 calculated ones. After that, they were subtracted from spectra gated with the signal
regions. The procedure of this subtraction was conducted for the analysis of the experiment (i) and (iii).
On the other hand, this was not conducted for the analysis of the experiment (ii) because the spectrum
for determining resonance parameters was not gated with the energies of γ-rays.
In the energy region of epithermal neutrons, the intensity of neutron beam increases for lower neutronenergies as a result of moderation. The energy spectrum of neutrons captured by the target must be
normalized in order to compare the measured spectrum and the calculated cross section. We used thespectrum gated with the 477.6 keV γ-rays from 10 B(n,αγ)7 Li reactions because the cross section of thisreaction has no resonance at the epithermal energy region [8]. This normalization was conducted for all
analysis, and the beam intensity was measured in each experiment.
3.2 Angular dependence
Figure 3 shows spectra of the neutron energy dependent on the angle of the γ-rays emitted from thecompound states of 118 Sn. An angular dependence of the shape of the p-wave resonance has clearly been
observed. To evaluate quantitatively, we deﬁned an asymmetry value ALH as
ALH =NL − NH,
NL + NH(2)

where NL and NH are the integrated values in a lower (L) and a higher (H) energy region of the p-wave
resonance, respectively. The integral regions were deﬁned using the p-wave resonance energy Ep and theresonance width Γp as follows: Ep − 2Γp < En < Ep for NL and Ep < En < Ep + 2Γp for NH . The valueof ALH in each angle is plotted in Fig. 4. Based on the Flambaum’s formalism, the angular dependence
can be written as
ALH = A cos θγ + B,
(3)
The expression was ﬁtted to the experimental data, with the result:
A = 0.494 ± 0.043, B = 0.040 ± 0.018.

(4)

3.3 Resonance parameters and branching ratios
Resonance parameters are determined by ﬁtting to the neutron energy spectrum gated with γ-raysof more than 2 MeV to suppress background events in low γ-rays’ energy region. Fitting function is
the Briet-Wigner formula convoluted the doppler broadening eﬀect [9] and a time structure of pulsed
beam [10]. Figure 5 shows the neutron energy spectrum around the p-wave resonance and ﬁtting result.
The resonance parameters in this p-wave resonance were determined asγ

Ep = 1.331 ± 0.002 [eV], Γp = 133 ± 5 [meV].

(5)

In s-wave resonances, an inﬂuence of self-shielding can be considered. In general, the cross sectionsof s-wave resonances is very larger than those of p-wave resonances, so that neutron beam cannot reachdeeply inside the target. This causes a shortage of the number of neutrons which interact with nuclei. As
this result, the shapes of resonances can be changed, and this eﬀect must be considered. At the moment,
the estimation of this eﬀect by a Monte Carlo simulation is ongoing.
Branching ratio of each resonance can be determined by ﬁtting like the determination of resonanceparameters. In this analysis, the spectrum which should be ﬁtted must be gated with the signal regions.
In this case, a ﬁtting function has a branching ratio as an only free parameter, while other parameters- 98 -

Asymmetry

JAEA-Conf 2020-001

0.6

χ2 / ndf
9.254 / 5
p0
0.03957 ± 0.01751
p1
0.4935 ± 0.04253

0.4

0.2

0
−0.2
−0.4
−0.6

Counts [A.U.]

Fig. 3. Capture yields as a function of neutron energy around the 1.33 eV p-wave resonance, for variable angles accessible at ANNRI. The central ﬁgure shows the placement
and the shape of each crystal. Two detectors
marked by crosses were not available for theexperiment (i).

−0.8

−0.6

−0.4

−0.2

0

0.2

0.4

0.6

0.8

cosθ

Fig. 4. The angular dependence of ALH .
The solid line shows the ﬁtting result using
the function of Eq. (3).

160
140
120
100
80
60
40
20
0
0.5

1

1.5

2

2.5
Neutron Energy [eV]Fig. 5. Capture yields as a function of neutron energy gated with an γ-rays’ energy regions of more than 2 MeV
around p-wave resonance in the experiment (ii). The solid line shows the best ﬁt line.

are ﬁxed as follows: f (En ) = Cσ(En ), where σ(En ) is the Briet-Wigner formula with ﬁxed resonanceparameters and C is the free parameter corresponding to the branching ratio of its resonance. Thus,
branching ratios are able to be determined after the determination of their resonance parameters. In thenear future, branching ratio of each resonance will be determined.

4. Summary
CP-violation (T-violation) is one of the essential contents to explain the dominance of matter overantimatter in the current universe. For the preparation of T-violation search using compound nuclei, one
has to determine the spin factor κ(J) which directly relates to the experimental sensitivity. The angularcorrelation of γ-rays emitted from the compound states of 118 Sn in the p-wave resonance can be clearlyobserved by the measurement at ANNRI in J-PARC MLF. In addition, the experiments to measure theresonance parameters and the branching ratios have been conducted and the analysis is ongoing. In thenear future, the mixing angle φ, and hence the value of κ(J) will be determined.

- 99 -

JAEA-Conf 2020-001

References[1] A. Sakharov, JETP 5, 24 (1967).
[2] G. Mitchell et al., Phys. Rep. 354, 157 (2001).
[3] V. Gudkov, Phys. Rep. 212, 77 (1992).
[4] T. Okudaira et al., Phys. Rev. C, 97, 034622 (2018).
[5] V. V. Flambaum, O. P. Sushkov, Nucl.. Phys. A 435, 352 (1985).
[6] Website of GEANT4, Available from: https://geant4.web.cern.ch/, accessed 2020/01/06.
[7] S. Takada et al., JINST 13 P02018 (2018).
[8] Website of JENDL-4.0, Available from: https://wwwndc.jaea.go.jp/jendl/j40/j40.html, accessed 2020/01/06.
[9] H. A. Bethe, Nucl. Phys. b 9, 69 (1937).
[10] K. Kino et al., Nucl. Instr. and Meth. A 736, 66 (2014).

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JAEA-Conf 2020-001

20 Measurement of the photon strength function in 115In at the ELBE facility
A. MAKINAGA1, 2,3, R. SCHWENGNER4, R. BEYER4, M. GRIEGER4, S. HAMMER4,
T. HENSEL4, A. R. JUNGHANS4, F. LUDWIG4, T. T. TRINH4, S. TURKAT4
1Department of Radiological Technology, Teikyo, University 6-22, Misaki-machi,

836-8505 Omuta, Japan

2Jein Institute for Fundamental Science, 606-8317, Kyoto, Japan

3Graduate School of Medicine, Hokkaido University, Kita 15 Nishi 7, Kita-ku,

060-8638, Sapporo, Japan

4Helmholtz-Zentrum Dresden-Rossendorf, 01328 Dresden, Germany*E-mail: makinaga@fmt.teikyo-u.ac.jp

The photon strength function (PSF) in 115In is an important parameter for the estimate of theneutron capture cross section on 114In in the field of astrophysics and nuclear engineering. Until
now, the so-called PSF method for 115In was applied only above the neutron-separation energy (Sn),and the evaluated 114In (n,) cross section has uncertainties caused by the lack of the PSF below
Sn. We studied the dipole strength distribution of 115In with a photon-scattering experiment using
bremsstrahlung produced by an electron beam of an energy of 10.3 MeV at the linear accelerator
ELBE at HZDR.

1Introduction

The nuclei heavier than iron are mainly produced via s-, r- or p- processes. The origin of p-nuclei
is assumed from photodisintegration in the O/Ne layers of core-collapse of massive stars, of type I
or II supernovae explosions, or/and s-, r- processes [1]. However, the production abundance of oneof the p-nuclei, 115Sn, still cannot be explained [2-3]. Recently, an s-process contribution is tried to
explain the neutron capture reaction and -decay at 113Cdm as competing reactions, i.e. 112Cd (n, )
113Cdm (-) 113In (n, ) 114In (-) 114Sn (n, ) 115Sn (see Figure 1). -rays were measured following neutron
captures on 112Cd at the MLF in J-PARC and it was found that the s-process contribution from
113Cdm is not big enough to explain the production problem of 115Sn [4]. So, origin of 115Sn is still open
question.
In this study, we will shift the viewpoint to 115In, which is produced via the main s-process. In
the p- or/and r- process environment (T > 2 - 3 x109 K), 115In has 3 possible reactions such as

Figure 1. Nuclear reaction path to produce 115Sn.

- 101 -

JAEA-Conf 2020-001

photodisintegration, neutron capture and -decay. While the effective -decay rate for 115In is theorder of 10-5 s-1 at the temperature higher than 3 x109 K [5], the stellar reaction rate ofphotodisintegration of 115In calculated using the TALYS code [6] is around 107 to 108 s-1 and theneutron-capture rate for 115In is around 108 -107 s-1 at a temperature of 3.5 - 4.0 x 109 K, respectively.
In addition, the neutron capture rate of 114In is around 108 s-1. It means that the production
process of 115In (, n) ⇆ (n, ) 114In (-) Sn (, n) ⇆ (n, ) 115Sn may contribute to produce 115Sn. This
story is also mentioned shortly in Ref. [3]. However, both the photodisintegration rate for 115In andthe neutron capture rate for 114In have not well been known experimentally, so they could not be
discussed quantitatively.
From the view point of nuclear physics, photo-neutron cross sections have been measured above
Sn [7-10]. However, precise measurements have not been performed (see Figure 2). The PSF below
Sn was measured using the Nuclear Resonance Fluorescence (NRF) in the energy range between
2-5 MeV at Darmstadt [11]. A photoactivation yield measurement on 115In performed at the ELSA
facility reported the possibility of an extra enhancement of the PSF in 115In below Sn [12]. On theother hand, neutron capture cross sections for 114In have not been measured because of the unstable
isotope. A possible way to estimate cross sections for 114In (n,) is to use the inverse-reaction method,which means the measurement of the PSF in 115In below Sn. Therefore, we have performed themeasurement of the PSF in 115In below Sn using the NRF method at the ELBE facility.

Figure 2. Current status of photo absorption cross sections of 115In.
Dashed line shows image of typical behavior below Sn.
No experimental data exists in this region.

2

Experimental methods and results

The photon-scattering cross section f(ER) can be measured via the ray transition from a given
excited level ER and deexcitation to a level Ef in the target. In case of non-overlapping resonances,photon scattering is described to process via a compound nucleus reaction with uncorrelatedchannels f characterized by the partial width f, so the photon-scattering cross section f(ER) canbe described as:- 102 -

JAEA-Conf 2020-001

Γ(1)

σγf �ER ��σγ �ER � fΓHere, all partial widths contribute to the total level width =f.
�I� � �� σ�� �E�dE ���� �� ��� ���� ��

�

��

�

(2)

ΔΩ

(3)

� Γ� �
�I� is the integral over the scattering cross section for the level R and Γ� is the partial width for atransition from R to a level f. The measured intensity of -rays emitted to the ground state at E=ER
at an angle can be expressed as,

Iγ �Eγ ,���Is �ER �Φ�ER ���Eγ �Nat W���4π

Here, Nat is number of the target nuclei per area unit, (E) is the absolute full-energy peak efficiency
at E , (ER) is the absolute photon flux at ER , W() is the angular correlation of this transition, andis solid angle of the detector.
If the electron energy is high enough above a particular level, the experiments withbremsstrahlung lead to the possibility of the population of a level by a feeding transition from ahigher-lying level. Such feeding increases the intensity of the transition to the ground state fromthe considered resonance R. The intensity of the transition to the ground state becomes asuperposition of the rate of elastic scattering and the intensity of the transitions feeding level R.
The cross-section integral Is+f can be expressed as:���� �� ��� � ������ ��

���I��� � �� σ�� �E�dE � ∑��� σ�� � dE

��

�

�

� ∑���

���� � ��� ������ � ��� ��

�

��� �
��

�

�� �

� Γ�� �� �
� �

(4)

Here, the summation over i>R means that the energy Ei of a level which feeds the considered
resonance R is higher than the energy ER of this resonance. i, i0, and iR are the total widths ofthe level Ei, the partial width of the transition to the ground state and the partial width of thetransition to the level R, respectively. Details of the experimental method are given in Refs. [13-20].
The photon-scattering cross section measurement on 115In was performed at thesuperconducting electron accelerator ELBE of the Research Center Dresden - Rossendorf.
Bremsstrahlung was produced by hitting 7μm niobium radiator with an electron beam of 10.3
MeV electron kinetic energy and an average current of 490μA. The produced bremsstrahlung wascollimated by an Al collimator with a length of 2.6 m and an opening angle of 5 mrad. A cylindrical
Al absorber of 10 cm length was placed between the radiator and the collimator to reduce the lowenergy part of the bremsstrahlung spectrum. The scattered photons were measured with four 100%
HPGe detectors surrounded by BGO escape-suppression shields. Two Ge detectors were placed at
90 degrees relative to the photon-beam direction. The other two HPGe detectors placed at 127
degrees were used to deduce angular distributions of the  rays. To reduce the intensity of the lowenergy part of background photons, absorbers of 3 mm Pb plus 3 mm Cu were placed in front of thedetectors at 127 degrees and 8 mm Pb plus 3 mm Cu were used for the detectors at 90 degrees. The
target consisted of 2376.4 mg of natural indium, formed into a disk of 2 cm in diameter. The naturalabundance of 115In is 95.7 %. The indium target was combined with a disk of 300.0 mg boron,
enriched to 99.5 % in 11B which was used to determine the photon flux. Spectra of scattered photons
were measured for 118 hours.
The photoabsorption cross section data of 115In obtained from the (,’) experiment are shown in
Figure 3, together with cross sections deduced from (,n) experiments [8] and with the TLO model in
RIPL [21]. The 115In (,’) cross sections smoothly connect to the photoneutron data, and show extra
enhanced resonances around 6 MeV and 9 MeV.

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JAEA-Conf 2020-001

Figure 3. (Color online) Photoabsorption cross sections deduced fromthe present measurement (red circles) in comparison with (, n) datafrom Ref. [8] (green squares), and TLO [21] (black dashed curve).

3

Summary

The dipole-strength distribution in 115In up to the neutron-separation energy has been studied
in an NRF experiment at the ELBE accelerator using a kinetic electron energy of 10.3 MeV.
The 115In (,’) cross section smoothly connects with the experimental photoneutron data. Extraenhancement of the PSF in 115In is also observed. A detailed analysis within nuclear models will be
performed.

Acknowledgments
We thank the crew of the ELBE accelerator for the cooperation during the experiment. This work
was supported by Helmholtz -Zentrum Dresden-Rossendorf and Hokkaido University.

References[1] M. Arnould, Astron. Astrophys. 46, 117 (1976).
[2] S. E. Woosley et al., Astrophys. J., Suppl. 36, 285 (1978).
[3] R. A. Ward et al., Astron. Astrophys. 103, 189 (1981).
[4] T. Hayakawa et al., Phys rev. C 94, 055803 (2016).
[5] K. Takahashi et al., Atom. Data and Nucl. Data Tables 36, 3, 375 (1987).
[6] A. J. Koning et al., AIP Conf. Proc. 769, 1154 (2005).
[7] S. C. Fultz et al., Jour, Phys. Rev. 186, 1255, 6910 (1969).
[8] A. Lepretre et al., Jour. Nucl. Phys. A, 219, 39, 7401 (1974).
[9] S. N. Belyaev et al., Jour. IZV, 62, 2223 (1998).
[10] O. V. Bogdankevich et al., Jour. ZET, 42, 1502 (1962).
[11] P. von Neumann-Cosel et al., Z. Phys. A 350, 303 (1995).

- 104 -

JAEA-Conf 2020-001[12] M. Versteegen et al., Phys. Rev. C 94, 044325 (2016).
[13] R. Schwengner et al., Phys. Rev. C 76, 034321(2007).
[14] G. Rusev et al., Phys. Rev. C 77, 064321(2008).
[15] R. Schwengner et al., Phys. Rev. C 78, 064314(2008).
[16] N. Benouaret et al., Phys. Rev. C 79, 014303(2009).
[17] G. Rusev et al., Phys. Rev. C 79, 061302(R)(2009).
[18] R. Schwengner et al., Phys. Rev. C 81, (2009).
[19] A. Makinaga et al., Phys. Rev. C 82, 024314(2010).
[20] A. Makinaga et al., Phys. Rev. C 90, 044301(2014).
[21] R. Capote et al., Nucl. Data Sheets, 110, 3107(2009).

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JAEA-Conf 2020-001

Comparison between experimental and calculation neutron spectra of the197
Au(γ,sn) reaction for 17 MeV polarized photon*T.K.T8<(71, T.S$1$0,1,2, H.Y$0$=$.,1,2, T.I72*$3, A.T$.(8&+,1,
S.M,<$02724 and Y.A6$124
1

SOKENDAI, Shonan Village, Hayama, Kanagawa 240-0193 Japan
KEK, 1-1 Oho, Tsukuba-shi, Ibaraki-ken 305-0801 Japan

2
3
4

JASRI, 1-1-1, Kouto, Sayo-cho, Sayo-gun, Hyogo 679-5198, Japan

University of Hyogo, 1-1-2, Kouto, Kamigori-cho, Ako-gun, Hyogo 678-1205, Japan
Email: tuyet@post.kek.jp

The double differential cross sections (DDX) were measured for the 197Au(γ,sn) with 17 MeV
polarized photons on a thin target. The DDX were compared with the result of PHITS calculation. To
reproduce the energy spectra of neutrons, especially at the relatively high energy region, a physics modelfor the direct process should be implemented in addition to the evaporation model.

1. Introduction
In an electron accelerator, neutrons are produced by collision between a high energy photon andcomponents of the accelerator through photonuclear reaction. The neutrons induce issues in radiation
protection as well as dosimetry and accelerator designs. Additionally, the study of photo-neutrons fromphotonuclear reaction is useful tool for understanding the elementary processes of nuclear physics. Until
now, there are researches about photo-neutrons production of photonuclear reaction [1,2,3]. In theprevious study of our research group [1], the energy spectra of photo-neutrons, which were produced by
the reaction of 17 MeV polarized photons on the thick Au target, were measured at various angles. The
evaporation and direct components were identified in the energy spectra [1]. The angular distribution of
evaporation component showed isotropic, while the direct component showed a dependence on theinteraction angle between photon polarization and neutron emission [1].
These spectra are useful in the comparison the experimental and calculation spectra to evaluate
parameters of theoretical calculation. However, there is no comparison between experimental andcalculation spectra. For comparison, the double differential cross section (DDX) is preferable than theneutron spectra. It is because the self-attenuation effect in thick target can be avoided. Thus, in this study,
we measured the DDX of the reaction of 17 MeV polarized photons on a thin Au target and compared
the results with that of PHITS calculation.

2. Experiment
The experiment was performed at NewSUBARU-BL01, Hyogo, Japan [4][5]. Figure 1 indicatesthe schematic drawing of NewSUBARU-BL01. The 17 MeV linearly polarized photons were producedby the collision of the polarized laser and electron beam with the energy of 982.4 MeV. The beam
current was 30 mA. A NdYVO4 laser system provided laser photons with a wavelength of 1.063 μm.

- 107 -

JAEA-Conf 2020-001

The power was 20 W. The laser polarization direction was set to be parallel to the floor using aЗ/4
polarizer plate.

Figure 1. Schematic of NewSUBARU-BL01.

Figure 2. Picture of the experimental setup.
Figure 2 shows a picture of the experimental setup. Since the photo-neutron production per
incident photon was considered, a plastic scintillator (thickness of 0.5 cm and surface area of 10 cm2)
was placed at 179.7 cm upstream from the target. The incident photon beam was collimated into thecenter of the target by using two collimators, C1 and C2 as shown in figure 1. The Au target was prepared
as a cylindrical pill whose diameter and thickness were 1 cm. In order to detect the neutron emitted fromthe target, a cylindrical NE213 organic liquid scintillator (12.7 cmIu12.7 cmL) was set at 90 degrees(respected to the photon beam axis) and 59 cm distance from the target. The NE213 detector detected
not only photo-neutrons from target but also gamma radiations. Thus, the pulse shape discrimination(PSD) method was applied to separate photo-neutron and gamma by obtaining the charge ratio of thewhole and tail of the waveform. The energy of photo-neutron was measured by a time-of-flight (TOF)- 108 -

JAEA-Conf 2020-001

spectroscopy to build up an energy histogram. For PSD and TOF analysis, a Data Acquisition (DAQ)
system based on VME standard was set up with a QDC module to obtain the full charge and tail chargeof signals from NE213, and a TDC to measure the time difference between the incident photon beam
and NE213. The DAQ system also measured the charge of signals from the plastic scintillator.

3. Data Analysis

We obtained the time of flight spectrum of neutron events by the PSD and TOF methods. Theenergy threshold employed for analysis was 0.25 MeVee. This threshold was determined by energy
calibration using gamma radiations of 137Cs, 22Na, and 60Co.
The time-walk effect can significantly affect the TOF measurement as well as neutron energy.
For time measurement, we used the constant-fraction-discriminators (CFDs). Although the time-walkeffect was small when using CFDs, this effect was still essential in the energy measurement of photoneutron. To minimize the time-walk effect, the correction of TDC with the fitting function as thecorrelation of TDC and QDC was applied.
Figure 3 shows neutron gamma separation with total vs slow gates (a) and TOF vs the ratio of
slow to total (b). We employ the latter correlation, for neutron and gamma separation.
The neutron energy was calculated by using TOF method. The efficiency of the neutron detector
was estimated by 252Cf experiment and SCINFUL-QMD simulation [6] to evaluate the total number ofneutrons emitted at 90 degrees from target. The energy spectrum was normalized by the solid angle, thenumber of incident photons and the number of target atom. Photon attenuation in the target was taken
into account using PHITS calculation. The double-differential cross-section of the photo-neutron
reaction was obtained by using the normalized data. In data analysis, the time resolution of gamma peak
was 0.76 ns, and the energy resolution of neutron was less than 10%.

(a)- 109 -

JAEA-Conf 2020-001(b)neutronFigure 3. Neutron gamma separation with total vs slow gates (a)and TOF vs the ratio of slow to total (b).

4. Calculation

PHITS (version 3.02) [7] was used to calculate the double-differential cross section (DDX) of
photo-neutron produced by the 197Au(γ,sn) reaction for 17 MeV polarized photon. A cylinder Au target
whose diameter and thickness were 5 μm was chosen to reduce the self-attenuation thickness effect of
the target. The DDX was obtained on a detector ring, whose width is 5 cm, the ring placed at 90 degrees
and 100 cm away from Au target.
Figure 4 shows the geometry of the DDX calculation in the PHITS code. Neutron production
from the nuclear reaction was calculated by the Generalized Evaporation Model (GEM) model. Figure
5 indicates neutron flux distribution in the PHITS calculation.
200
150

100
50
0

0

150
100
50
Figure 4. Geometry in PHITS.

- 110 -

200

JAEA-Conf 2020-001

Figure 5. Neutron flux distribution.

5. Result
Figure 6 shows DDXs obtained from the experiment and the PHITS calculation. In this figure,closed circles indicate the experimental data and the solid line is the result of the calculation. The
experimental spectrum shows two components; evaporation component and direct component. Theenergy region of the evaporation and direct component were less and more than 4 MeV, respectively. In
contrast, the calculation result shows only the evaporation component, as expected. Thus, a model toreproduce the direct component should be included in the simulation for the photonuclear reaction. To
develop the model, experimental data of DDX are strongly desired for various targets and energies.

Figure 6. Experimental and calculation DDX.

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JAEA-Conf 2020-001

References[1] Y.Kirihara, H.Nakashima, T.Sanami, Y.Namito, T.Itoga, S.Miyamoto, A.Takemoto, M.Yamaguchi
and Y.Asano, Neutron emission spectrum from gold excited with 16.6 MeV linearly polarizedmonoenergetic photons, Journal of Nuclear Science and Technology (2019),
https://doi.org/10.1080/00223131.2019.1691073.
[2] K.Horikawa, S.Miyamoto, T.Mochizuki, et al., Neutron angular distribution in (γ, n) reactions withlinearly polarized γ -ray beam generated by laser Compton scattering, Phys. Lett. B., (2014).
[3] A.Agodi, Nuovo Cimento, 5, 21, (1) (1957).
[4] Y.Asano, S.Miyamoto, Shielding design of laser electron photon beamlines at Sring-8, Prog. Nucl.
Sci. Tech. (2014).
[5] S.Miyamoto, Y.Asano, S.Amano, et al., Laser Compton back-scattering gamma-ray beamline on
NewSUBARU, Radiat. Meas. (2007).
[6] D.Satoh, S.Kunieda, Y.Iwamoto, et al., Development of SCINFUL-QMD code to calculate
the neutron detection efficiencies for liquid organic scintillator up to 3 GeV, Nucl. Sci. Tech. (2002).
[7] T.Sato, Y. Iwamoto, S. Hashimoto, T. Ogawa, T. Furuta, S. Abe, .T.Kai, P.Tsai, N. Matsuda, H.
Iwase, N. Shigyo, L. Sihver and K. Niita, Features of Particle and Heavy Ion Transport code System(PHITS) version 3.02, J. Nucl. Sci. Technol. (2018),
https://doi.org/10.1080/00223131.2017.1419890.

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JAEA-Conf 2020-001

22 Measurement of neutron total cross-section of93
Nb at J-PARC MLF ANNRI
Shunsuke Endo, Atsushi Kimura, Shoji Nakamura, Osamu Iwamoto, Nobuyuki Iwamoto,
Yosuke Toh, Mariko Segawa, Makoto Maeda and Masayuki Tsuneyama
Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195
endo.shunsuke@jaea.go.jp

Abstract
The neutron total cross-section of Nb was measured by a transmission method to examine theresonance data by J-PARC MLF ANNRI. During the measurements, a diffraction effect of the Nb
sample was remarkably observed. The diffraction effect of Nb was also studied since its effect cannot
be ignored in deriving the neutron total cross-sections in a low energy region. As a result of the
discussion by assuming a simple model including structure, it is clear that the diffraction effect is notignored below 0.2 eV. This article gives results of the neutron total cross-section measurement of Nb
at J-PARC MLF ANNRI and discussions on the effect of diffraction.

1

Introduction

Niobium(Nb)-93, whose natural abundance is 100%, is one of material elements to strength-en
stainless steel and is used in structural materials of nuclear reactors. Since Nb is also used as anelement of superconductor alloys, it has been used in fusion reactors and accelerators. Niobium causes
long-lived activities in stainless steel due to the long half-life of 94 Nb, about 20 thousand years [1].
Determination of resonance parameters with high accuracy are therefore required to estimate theactivation. The cross-sections of Nb were measured by Wang et al. [2] , Meaze et al. [3], Drindak
et al. [4] and so on. The resonance parameters determined by them have discrepancies as shown in
Table 1. In particularly, there are discrepancies larger than uncertainties between the results of Wanget al. and those of Drindak et al.. In order to determine which resonance parameters are more likely,
measurements of neutron total cross-sections were performed at Accurate Neutron-Nucleus Reaction
measurement Instrument (ANNRI) of Material and Life Science Experimental Facility (MLF) in Japan
Proton Accelerator Research Complex (J-PARC).

Table 1: Resonance parameters for the first resonance of 93 Nb in past measurements. Γγ , Γn
and g are the gamma width, the neutron width and the statistical weight factor, respectively.
Reference
Wang+(2011)[2]Meaze+(2011)[3]
Drindak+(2006) [4]
Sharapov+(1999) [5]Saplakoglu+(1958)[6]Mughabghah 6th(2018)[7]JENDL-4.0(2010)[8](=ENDF/B-VIII.0(2018)[9])Resonance Energy [eV]35.994 ± 0.0026
35.92 ± 0.02
35.922 ± 0.004
35.9
35.9
35.9 ± 0.1

Γγ [meV]224.61 ± 20.82
215.57 ± 17.98
191 ± 8205 ± 51
218 ± 182gΓn [meV]
0.0772 ± 0.0037
0.078 ± 0.041
0.125 ± 0.001
0.11 ± 0.01
0.15 ± 0.01
0.125 ± 0.010

35.922191

0.125- 113 -

JAEA-Conf 2020-001

2

ExperimentTransmission measurements were performed at the ANNRI of MLF in J-PARC in June of 2019. In
the MLF [10], neutrons were produced by spallation reactions of mercury targets. The 3-GeV rapid
cycling synchrotron, producing pulsed proton beam, was operated in double-bunch mode and 500-kW
beam power with a repetition rate of 25 Hz at that time. The ANNRI is one of beamlines in the MLF,
where germanium gamma-ray detectors and Li-glass neutron detectors are installed to measure nuclear
reactions. Cross-section measurements have been performed at the ANNRI [11].
Figure 1 shows the experimental scheme. Two types of Li-glass detectors were installed at ANNRI,
one is 6 Li-enriched (GS20) and the other one is 7 Li-enriched (GS30). GS20 can detect γ-rays as well as
neutrons in the reaction: 6 Li +n → 3 H + α. Time of flight (TOF) spectrum ascribed to only neutron
can be obtained by subtracting the GS30 spectrum from the GS20 one. These Li-glass detectors wereplaced at 28.2-m flight length. Signals from detectors were processed by a V1720 module made in
Costruzionl Apparecchiature Elettroniche Nucleari S.p.A. (CAEN) [12], and pulse height and TOF
were recorded in each event. Neutron total cross-sections can be obtained by neutron transmission
ratio. In order to determine the transmission ratio, measurements for beam open (blank) and Nb
sample-in were performed for 10 hours and 12 hours, respectively.
The size of used Nb target is shown in Table 2. Table 3 shows the impurities contaminated in the
Nb target. Since tantalum has similar chemical properties to Nb, it is included a little in Nb sample. In
this article, when cross-section is derived, effects of Ta are not corrected because this effect is estimatedto be the order of about 10−3 b in a region lower than 1 eV.

Table 2: The size of natural Nb target
Surface area
653.3 ± 5.7 [mm2 ]Thickness
10.2834 ± 0.0002 [mm]Weight56.1112 ± 0.0003 [g]Surface density (5.567 ± 0.049) × 10−2 [/b]Figure 1: Experimental scheme

Table 3: Impurities contaminated in the Nb target
ElementTaFeO
C N others
Content [ppm] 2000 200 100 50 50
35

3Analysis

3.1

Corrections

Figure 2 shows pulse-height spectra by GS20 and GS30 in the measurement of the Nb sample.
GS30 has a slight sensitivity to neutron since 6 Li is contained by about 0.01%. In order to remove
noises and low-energy γ-rays backgrounds, a threshold of pulse-height was set at 210 channel. Eventswere used when the pulse-height is higher than the threshold. Dead-time corrections were made using
a dead-time of 625 ns determined by the setting of the V1720 module.
In the present measurements, there are two types of backgrounds due to neutrons of frame-overlap
and to γ-rays. First, the frame-overlap neutrons were estimated by following function:
t − 37f (t) = a exp −
+ c,
(1)
t1

- 114 -

JAEA-Conf 2020-001

where t represents TOF in ms and a, t1 and c are fitting parameters. Frame-overlap effect was correctedby fitting this function to the TOF spectrum between 37 ms and 40 ms. Since the background eventsby this effect were much smaller than true events, this correction has little effect on the derivation ofcross sections.
Second, γ-rays backgrounds were removed by subtracting TOF spectrum of GS30 from that ofGS20 as mentioned in Section 2. Figure 3 shows the TOF spectra of GS20 and GS30 and estimatedframe-overlap correction component.

Figure 2: Pulse height spectra of GS20 (redFigure 3: TOF spectra by Li-glass detec-line) and GS30 (black line) with threshold line
(pink line)

tors A (blue line) and B (red line) and frameoverlap correction component (black line)

3.2

Derivation of total cross-sections

Transmission ratio (T ) was obtained by dividing the Nb sample-in spectrum (Ysample ) by the blank
spectrum (Yblank ). The cross-section can be derived from
σ(E) = −1
1 Ysampleln= − ln T (E),n
Yblankn(2)where n is the surface density. The obtained cross-sections are plotted in Fig. 4 and Fig. 5 incomparison with other past measurements and JENDL-4.0. The effects of resolution function anddoppler broadening were taken into account on JENDL-4.0 for comparison with experimental results.
The resonances at 4.3 eV and 10.4 eV are ascribed to 181 Ta. The cross-sections are fluctuated in a low
energy region because of neutron diffraction. We thought this diffraction effects were observed clearly
due to small capture cross-section of Nb (about 1.1 b at thermal energy). Wang et al. used spallation
neutron sources as we did, while the other past group used nuclear reactor and fast chopper. Both oftheir results also are fluctuated in the same way as ours.

4

Estimation of neutron diffraction

4.1

The formalism of neutron diffraction

In the Nb measurements, neutron diffraction was clearly observed in a low energy region, andtherefore we examined this neutron diffraction effect. Since the effect of neutron diffraction is due tocrystal structure, it is not appropriate to discuss nuclear reactions in a single nucleus. For this reason,
polycrystalline model (abbreviate this to PCM hereinafter) was used here for estimation of neutron
diffraction.
In PCM, the total cross-section for scattering processes including neutron diffraction can be written[13] as:inel (E) ,elel (E) + S inel (E) + σ b
b
(3)
(E)
+
S
S
Scoh
σs (E) = σcoh
incoh
incoh incohcoh- 115 -

JAEA-Conf 2020-001

Figure 4: Neutron total cross-sections by the

Figure 5: Neutron total cross-sections by thepresent and other past measurements at low
energypresent and other past measurements at resonance regionb and σ bwhere each function S(E) contains the information of sample system; σcoh
incoh are the boundcoherent and incoherent cross-sections relating to cross-section σ:

2
A+1
σb =
σ,
(4)
A
where A = M/m; M and m are target mass and neutron mass, respectively. For Nb, incoherent effects
can be ignored since σinc is negligibly small - about 0.038% against coherent effects. That is why
the first term in Eq. (3) is only needed for estimation of neutron diffraction. The elastic coherent
contribution of a polycrystalline solid can be given [13] by:el (E) =Scoh

π 2 h̄22mEv0

τhkl <2kn


τhkl



3h̄2 φ1 (Θ) 2w(τhkl )
exp −
τhkl
τhkl
2M kB θD



,

(5)

where kn , v0 , h̄, kB and θD are neutron wave vector, the volume of the unit cell, the reduced Plank
constant, Boltzmann constant and Debye temperature, respectively; w(τhkl ) is the multiplicity of reciprocal lattice vectors τhkl :2π  2
τhkl =
h + k 2 + l2 ,(6)awhere a is lattice constant and h, k, l represent crystal planes. The φm term is described as:
 1
φm (Θ) =dε εm (eε/Θ − 1)−1 ,(7)−1where Θ = T /θD and T is the absolute temperature. Here, inelastic coherent contribution involving
phonons can be considered as the same as inelastic incoherent one [14]. The inelastic incoherent
contribution [13-15] can be approximately expressed by the following equation:




2 

k θ
Ak θ
inel (E) = S inel (E) =
Sinc
coh

A
A+1

1+

3
φ3 (Θ) B D
4A
E

−

1
B D
12 Eφ1 (Θ)

1 − exp−12E
φ1 (Θ)
AkB θD,(8)

where the first term represents the total incoherent cross-section, consisted of elastic and inelastic
incoherent processes, and the second term is the elastic incoherent one. Inelastic coherent cross-section
can be obtained by Eq. (8) same as inelastic incoherent one.
Niobium crystal has a body-centered cubic lattice structure. In this case, summation in Eq. (5) isdone when h+k +l is even. Figure 6 shows cross sections calculated with PCM by Nb parameters listed
in Table 4, and experimental results and calculated ones with free gas model (FGM). Cross sections
by FGM are normalized to that by PCM at 8 eV.

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JAEA-Conf 2020-001

Table 4: Niobium parameters used for the PCM calculation
σcoh
σinc
θD
a
6.253 [b] [7] 0.0024 [b] [7] 265 [K] [16] 3.3004 [Å] [17]Figure 6: Cross-sections calculated with PCM and FGM andexperimental results. Coh Ela and Coh Inela mean the coherent elastic and the coherent inelastic scattering cross-sections
calculated by PCM. Total cross-sections of PCM are determined by adding Coh Ela, Coh Inela and absorption crosssections.

4.2

Figure 7: Differences between PCM, FGM
and experimental results. Green band indicates the statistical and systematic uncertainties of experimental results.

Discussion of neutron diffraction and evaluation of nuclear data

Figure 7 shows differences in FGM and PCM from experimental results. Green band indicatesthe statistical and systematic uncertainties of experimental results. It is found that PCM has smaller
discrepancy than FGM and that PCM can consider diffraction effects. Nevertheless, PCM is not enoughto reproduce neutron diffraction effects. This might be due to the distortion of sample. From the region
where the structure effect is not clear in PCM, it was decided not to use the present results below 0.2 eV
where the difference between PCM and experimental results is larger than the experimental uncertainty.
The uncertainty coming from the difference of the measured data might be also added to the systematic
uncertainty.

5

Summary

The transmission measurements at J-PARC MLF ANNRI were performed in order to obtain neutron
total cross-sections of 93 Nb. In the Nb measurements, neutron diffraction was observed in the low
energy region, and then this effect was examined with PCM. It became clear that PCM can consider
diffraction effects, specially the energy occurring the fluctuation by diffraction. Since PCM is notenough to reproduce diffraction effects completely, it was decided not to use the present results below
0.2 eV. The total cross-sections obtained in this work will be analyzed with REFIT code to evaluate
the resonance parameters.

Acknowledgments
The authors would like to thank staffs for their efforts in operating the accelerators and the neutron
production target of J-PARC MLF. This work was supported in part by Grants-in-Aid for Scientific
Research (JP17H01076).

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JAEA-Conf 2020-001

References[1] M. Okoshi. E®ects of Activation on the Dismantlement of Reactors - From the View Point ofRadioactive Waste Management - RADIOISOTOPES. 1998;47:412-423.
[2] T. Wang et al. Measurements of neutron total cross-sections and resonance parameters of niobium
using pulsed neutrons generated by an electron linac. J Radioanal Nucl Chem 2011;289:945-952.
[3] A. K. M. M. H. Meaze and G. N. Kim. Determinations of niobium and palladium resonancecross-sections by time-of-°ight transmission experiment. Indian J. Phys. 2011;85(2):329-338.
[4] N. J. Drindak et al. Neutron Capture and Transmission Measurements and Resonance Parameter
Analysis of Niobium. Nuclear Science and Engineering. 2006;154:294-301.
[5] E. I. Sharapov et al. Search for parity violation in 93 Nb neutron resonances. Phys. Rev. C.
1999;59:1131-1135.
[6] A. Saplakoglu et al. Properties of s-Wave and p-Wave Neutron Resonances in Niobium. Phys Rev
1958;109:1258-1262.
[7] S. F. Mughabghab. Atlas of Neutron Resonances 6th ed, vol. 1. Amsterdam: Elsevier; 2018.
[8] K. Shibata et al. JENDL-4.0: A New Library for Nuclear Science and Engineering. J. Nucl. Sci.
Technol. 2011;48:1-30.
[9] B. A. Brown et al. ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Librarywith CIELO-project Cross Sections, New Standards and Thermal Scattering Data. Nuclear data
sheets 2018;148:1-142.
[10] F. Maekawa et al. First neutron production utilizing J-PARC pulsed spallation neutron sources
JSNS and neutronic performance demonstrated. Nucl Instruments Methods Phys Res Section.
2010;620:159-165.
[11] K. Terada et al. Measurements of neutron total and capture cross sections of 241 Am with ANNRI
at J-PARC. J. Nucl. Sci. Technol. 2018;55:1198-1211.
[12] CAEN web site, https://www.caen.it(Accessed: 2020-3-30).
[13] J. R. Gradana. Total scattering cross section of solids for cold and epithermal neutrons. Z. Naturforsch. 1984;39a:1160-1167.
[14] K. Binder. Total coherent cross section for the scattering of neutrons from crystals. Phys. Stat.
sol. 1970;41:767-779.
[15] G. Placze and L. V. Hove. Interference e®ects in the total neutron scattering cross-section of
crystals. Il Nuovo Cimento. 1955;1:233-256.
[16] C. Y. Ho et al. Thermal Conductivity of the Elements: A Comprehensive Review. Journal of
Physical and Chemical Reference data. 1974;3:1-796.
[17] M. E. Straumanis, and S. Zyszczynski. Lattice parameters, thermal expansion coe²cients anddensities of Nb, and of solid solutions Nb-O and Nb-N-O and their defect structure. J. Appl.
Crystallogr. 1970;3:1-6.

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JAEA-Conf 2020-001

23 High-Energy Measurement of the Neutron Capture
Cross Section of 237Np
G. Rovira1∗ , T. Katabuchi1 , K. Tosaka1 , S. Matsuura1 , K. Terada1 , Y. Kodama1
H. Nakano1 , O. Iwamoto2 , A. Kimura2 , S. Nakamura2 , N. Iwamoto2
1) Laboratory for Advanced Nuclear Energy, Tokyo Institute of Technology
2) Japan Atomic Energy Agency
author: rovira.g.aa@m.titech.ac.jp

Abstract
The neutron capture cross-section of 237 Np has been measured with the Accurate Neutron Nucleus Reaction
Measurement Instrument (ANNRI) at the Materials and Life Science Facility (MLF) of the Japan Proton Accelerator Research Complex (J-PARC) using neutrons with energy ranging from thermal energy to 500 keV. The
absolute cross-section was determined using the values for the whole shape of the ﬁrst resonance from JENDL4.0 and also with the total neutron ﬂux derived from a 197 Au sample measurement in which the ﬁrst resonancewas completely saturated. The results obtained from both techniques present agreement within 2%. In the highenergy region, the cross-section was determined with uncertainty below 4% from 0.5 to 25 keV. Along with thecross section measurement, theoretical calculations were performed to reproduce the present results.

I. Introduction
Accurate nuclear data are of the essence for the nuclear transmutation of minor actinides (MAs)as the nuclear industry is set to tackle the issue of high-level waste (HLW) management. Nuclear
transmutation facilities will take the task of converting HLW into short lived or even stable nuclei.
Current evaluated nuclear data libraries are only suitable for the early stages of the design of nuclear
transmutation systems. However, ﬁnal designs and safety measures demand more precise nuclear datawith a signiﬁcant reduction in terms of their uncertainties [1, 2].
237 Np has a long half-life of 2.14 x 106 years and it is one of the most abundant MAs present inspent nuclear fuel. 237 Np is also one of the main components of the Accelerator-Driven Systems (ADS)
core, a subcritical reactor facility for nuclear transmutation. The region of interest for the core design isfrom 0.5 to 500 keV. Current uncertainties in JENDL-4.0 [3] from 0.5 to 500 keV for the neutron capture
cross section of 237 Np (6-10%) are much higher than the requirements of 3%. Hence, it is of upmost
importance to accurately determine the neutron capture cross section at such energy range to reduce
the uncertainties.
In the region from 0.5 to 500 keV, there are only two sets of data using time-of-ﬂight (TOF) methodthat cover the whole energy range, those of Weston et al [4] and Esch et al [5]; and they present divergences from 15% to 35% . Experimental data by activation method exists in the 100-500 keV range but
they differ from each other about 30-40% [6–8].
In this paper, results of the neutron capture cross section for 237 Np are presented for incident neutron energy ranging from thermal energy to 500 keV with emphasis in the region of 0.5 keV to 500
keV. Moreover, in order to provide more accurate data, two normalization techniques were used to obtained an absolute value for the cross-section. The ﬁrst technique was based on using the whole shapeof the ﬁrst resonance from JENDL-4.0 to determine a normalization factor. In the second technique,the energy-dependent cross-section was normalized using the total neutron ﬂux obtained from a Au
sample measurement in which the ﬁrst resonance was completely saturated. In addition, the results ofthe experiments were complemented using the CCONE code [9] in order to assess the reliability of theresults. Details of the experimental setup and the data analysis are also provided.

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II.

Experimental Procedure

1. Experimental Setup
The experiments were performed using the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Materials and Life Science Facility (MLF) of the Japan Proton Accelerator
Research Complex (J-PARC). Intense pulsed neutrons were produced by the Japanese Spallation Neutron Source (JSNS) in the MLF using the 3 GeV proton beam of the J-PARC facility. The proton pulses
were shot at the spallation target every 40 ms and with a beam power of 400 kW.
A TOF method was employed in the present experiment with a ﬂight path of 27.9 m up to thesample position. Emitted γ-rays from the sample were detected by a NaI(Tl) detector. Detected capture events were stored sequentially with the corresponding TOF and energy value in a computer as a
two-dimensional list format data.
A multi-event time digitizer FAST ComTec MPA4T [10] was employed for fast data acquisition
purposes [10]. The time between the starting spallation trigger event and successive multiple stopfrom the detected γ-rays were digitized. The signal coming from the JSNS proton beam monitor wasused as a trigger signal for the MPA4T module. The energy of the detected γ-rays was measured
using traditional pulse-height method with the signal coming from the dynode of the photomultipliertubes (PMT) from the NaI(Tl) detector. At the same time, in order to reduce the effect of the strong
γ-rays emitted from the spallation reaction on the detected events in the keV region, the signal fromthe anode of the PMT was employed for pulse-width measurement to also determine the energy on the
γ-rays. More information on the pulse width measuring technique employed at ANNRI can be foundhere [11].

2. Samples
A 200 mg (5 MBq) 237 Np sample of 20 mm diameter and 1 mm thickness was utilized for theexperiments. The 237 Np sample consisted of 227 mg of neptunium dioxide (NpO2 ) powder bindedwith 624.5 mg of Al powder. The isotopic purity of 237 Np for the sample was 99.99%. The powders
were packed into a 30 mm diameter Al pellet with 0.4 mm thick walls. An exact replica of the Al
dummy container was used to derive the background induced by the Al case. The incident neutron
spectrum was reconstructed using γ-rays from the 197 Au(n, γ) reaction with a 20 mm in diameter and
1 mm in thickness gold sample and, also, using the 478 keV γ-rays from the 10 B(n, α)7 Li reaction witha boron sample containing enriched 10 B up to 90% and with a diameter of 10 mm and a thickness of
0.5 mm. Background events due to scattered neutrons were derived using a nat C sample with a 10 mmdiameter and 0.5 mm thickness.

III.

Data Analysis

1. Pulse Width to Pulse Height Conversion
Since pulse-width and pulse-height data were simultaneously recorded for all measured γ-ray, aconversion relation can be established by plotting both recorded data in a two-dimension histogram.
More information about the pulse width analysis is described by Katabuchi et al [11].

2. Background Removal
In order to isolate the detected events coming from the 237 Np(n,γ)238Np reaction, different measurements and techniques were applied to the recorded data from the 237Np sample measurement run.
A dead time correction was applied ofﬂine to all measurements in order to estimate the count loss
in the experiment [11]. The main cause for this count loss was the pile-up of two consecutive signals.

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Capture events induced from slow neutrons coming from previous neutron bursts have to be subtracted. Every proton event induces neutron events with a frame length of 40 ms, as the proton beam
repetition is 25 Hz. Overlapping background can be estimated using the operation pattern of J-PARC.
A small part of the proton beam pulses from the 3-GeV synchrotron are injected into the 50-GeV synchrotron ring instead of JSNS. When no proton is shot into the JSNS, the measurable TOF is doubled
since there is no trigger signal. Thus, the overlap background was estimated from the recorded eventsfrom 40 ms to 80 ms.
Blank background is subtracted using the data retrieved from a measurement with no sample.
Likewise, the background events induced due to scattered neutrons at the sample and the events induced by the sample case are removed using the nat C and the TOF spectra obtained from the aluminumcase respectively.

3. Neutron Spectrum
The neutron spectrum was estimated using the gold and boron samples. The obtained TOF spectrum from both runs was divided by the reaction rate simulated using the PHITS program [12]. Figure
1 shows a good agreement of the incident neutron distribution between the two samples except for the
resolved resonance region of gold. Hence, for better performance, only the neutron spectrum from theboron measurement was used in the analysis.

1011

Neutron Spectrum from Boron
Neutron Spectrum from Au

Relative neutron intensity [n/eV/s]

1010
109
107
106
105
103
102
101
100
10-3

10-2

10-1

100

101
102
Neutron energy [eV]

103

104

105

106

Figure 1: Incident neutron spectrum calculated from the gold and boron samples. Over 10 eV the neutron spectrum from theboron measurement offers better results.

4. Data Normalization
A relative capture cross-section can be obtained from the derived neutron capture yield and thethe incident neutron spectrum. However, since the measurement times and reaction rates are different
for the Np and the boron samples, the obtained energy dependence of the cross-section has to benormalized.
The ﬁrst normalization process consisted on using the evaluated data from JENDL-4.0 for the wholeshape of the ﬁrst resonance, from 0.25 eV to 0.7 eV, to normalize the cross-section results. The best ﬁt
for the normalization was obtained by minimizing the following residual:nρ = ∑ [ y i − N · x i ]2
i =1

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(1)

JAEA-Conf 2020-001

being n the number of energy points from 0.25 to 0.7 eV, yi the evaluated cross-section values fromJENDL-4.0, xi the experimental results and N the normalization factor.
Alongside this method, the experimental data was also normalized using the saturated resonancetechnique to assess the reliability of the results. A gold sample used in the experiments was thickenough for the ﬁrst resonance to be completely saturated. Since the resonance is saturated, all neutrons
incoming with the energy of 197 Au(n,γ) ﬁrst resonance (4.9 eV) react with the sample. This assumption
can be seen in ﬁgure 2. However, since the experimental setup has a detection threshold of 920 keV, acorrection has to be made for the capture yield loss due to this threshold. The capture γ-ray spectrum
for the 197Au(n,γ)198Au and the 237 Np(n,γ)238Np reaction were calculated with CCONE at the energyof the 197 Au(n,γ)198Au ﬁrst resonance. The capture yield loss for Au was estimated as the total energy
loss, namely the total γ-ray emitted under the 920 keV threshold, divided by the total excitation energy.
109

Au Capture Yield
Neutron Flux Energy Dependance from Boron

108

107

Counts [1/µs]

106

105

104

103

102
100

101

102

103

104

Time [µs]Figure 2: Energy dependence of the indicent neutron ﬂux (red) normalized to the 197 Au neutron capture yield (blue) at theenergy of the ﬁrst resonance (4.9 eV).

This correction for the saturated resonance method, Nsat can be expressed as:
1−
Nsat =
1−

E N p (0−920)
Np
Bn
E Au (0−920)
BnAu

E

)
/ N pN(totp/Bn
E Au ( tot )
BnAu

·

SN p
1
·
S Au t N p(2)being EN p(0−920) and EN p(tot) the energy sum of the capture γ-rays from 0 to 920 keV and totalNpemitted, BnAu and Bn the binding energies for 198 Au and 238 Np respectively. S N p and S Au are theproton bursts shot during the Np and Au measurements and t N p the Np sample thickness in at/b.

5. Uncertanty analysis
In this experiment, alongside the statistical uncertainties related to the counting rate, several systematic uncertainties have been considered. These systematic uncertainties relate to the normalization,dead-time correction, frame overlap subtraction, self-shielding and multiple scattering corrections andthe neutron ﬂux calculation.
The total error can be seen in ﬁgure 3 for both normalization techniques. The ﬁrst resonancenormalization offers better results as the normalization error is only of 0.5 %, whereas the error forthe saturated resonance normalization is of 2.54%. The results with the ﬁrst resonance normalization
present uncertainties below 3% from 0.5 to about 2 keV and maintain a value of about 4% up to 25 keV.
From this energy point, the effects of the neutron scattering induced by 27 Al are visible as they reducethe incident neutron ﬂux.

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JAEA-Conf 2020-001

Present First Resonance Normalization
Present Saturated Resonance Method Normalization
1

Uncertainty [%]

10

10

0

0

1

10

10
Neutron energy [keV]

2

3

10

10

Figure 3: Total uncertainty.

IV. Cross-section Results
The neutron capture cross section obtained from both normalization techniques can be seen in ﬁgure 4. The results from both normalization agree within uncertainties and offer good overall agreementwith the evaluated data from JENDL-4.0.
104

Present First Resonance Normalization
Present Saturated Resonance Method Normalization
JENDL-4.0

Cross Section [b]

103

102

101

Residuals

100

10-1
1.02
1.01
1
0.99
0.98
10-2

10-1

100

101
102
103
Neutron energy [eV]

104

105

106

Figure 4: 237 Np neutron capture cross-section from 10 meV to 1 MeV using the ﬁrst resonance (red) and the saturated
resonance method (blue) for normalization. The data residuals are plotted underneath.

In the region of interest, from 0.5 to 500 keV, only the results form the ﬁrst resonance normalization
have been considered as they provide less uncertainty.
Figure 5 shows a comparison with previously reported experimental data in the high energy region.
As can be seen, the present data holds better agreement with the experimental data by Weston et al [4]than with the data from Esch et al [5]. In the energy range from 0.5 to 20 keV, where the discrepancies
with JENDL-4.0 are over 15%, data from Esch et al [5] provides even lower values than those fromJENDL-4.0, specially between energies of 1 to 10 keV where the differences with the data by Esch et alamount to 25%. Over 100 keV, only activation data by Lindner et al [7] is consistent with the present
data as activation data from Troﬁmov et al [6] and Stupegia et al [8] overestimate both the present data
and that of Lindner et al and also evaluated data from JENDL-4.0 and ENDF/B VIII.0 [13].

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JAEA-Conf 2020-001

Cross Section [b]102Stupegia (1967)
Lindner (1976)
Ju. N. Trofimov (1983)
Weston (1981)
Esch (2008)
Kobayashi (2002)
Scherbakov (2005)
Present
JENDL-4.0 (n, g)
EDNF/B VIII.0 (n, g)

101

100

10-1
10-1

100

101
Neutron energy [keV]

102

103

Figure 5: Neutron capture cross-section of 237 Np from 0.5 keV to 1 MeV.

V. Theoretical Calculations
The cross-section results of the 237 Np neutron capture were evaluated by means of theoretical
calculations using the CCONE code [9]. Since discrepancies exist with the current JENDL-4.0 datalibrary, the aim of this analysis is not only to assess the reliability of the results but to provide a newtentative evaluation for the neutron capture cross-section of 237 Np.
The 237 Np neutron capture cross-section was derived by changing the E1 transition gamma strength
function. For the rest of the calculations, the same parametrization as in the JENDL-4.0 were usedfrom 0.5 keV to 1 MeV. The ﬁnal results for all reaction channels are shown in ﬁgure 6 together withevaluated results from JENDL-4.0. The present evaluation shows good agreement for the ﬁssion andinelastic channels with JENDL-4.0. For the 237Np neutron capture cross-section, the results of thepresent evaluation provide better agreement with the experimental results in the whole energy rangeof the calculations.
103

Experimetal DataJENDL (n,g)
JENDL (n,f)
JENDL (n,inl)
CCONE (n,tot)
CCONE (n,g)
CCONE (n,f)
CCONE (n,inl)

Cross Section [b]

102

101

100

10-1

10-2

10

-3

102

103

104
Neutron energy [eV]

105

106

Figure 6: Evaluated 237 Np neutron capture cross-section results from the CCONE calculation (red) in comparison with theresults of JENDL-4.0 (blue)

VI.

Conclusions

The neutron capture cross-section of 237 Np was measured with incident neutrons ranging from10 meV to 500 keV. A time-of ﬂight method was employed using the NaI(Tl) spectrometer of the
ANNRI beamline at J-PARC. In the high energy region, the cross-section results were obtained withuncertainty below 4% from 0.5 to 25 keV. However, over that energy the uncertainty increases to over
8% due to an increase of the statistical uncertainty. Nonetheless, from 0.5 to 25 keV the results offer- 124 -

JAEA-Conf 2020-001

a much lower total uncertainty than the uncertainty included in JENDL-4.0 of 6-10%. In comparison
with experimental data, the present data holds better agreement with the experimental data by Weston
et al [4] as the results from Esch et al [5] underestimate the present results. Theoretical calculations
were performed with the CCONE code to reproduce the experimental results. The new calculations
provided by the present analysis offer better agreement with the experimental data than JENDL-4.0,
specially in the region from 0.5 to 20 keV where the differences between JENDL-4.0 and the present
experimental amount to 10-25%.

References[1] H. Iwamoto et al., Nucl. Data Sheets 2014; 118:519-522.
[2] Int. Eval. Co-op., Vol. 26, NEA/OECD, 2015.
[3] K. Shibata et al., J. Nucl. Sci. Technol. 2011; 48:1-30.
[4] L.W. Weston et al., Nucl. Sci. Eng. 1981; 79:184-196.
[5] E-I. Esch et al., Phys. Rev. 2008; C77:034309.
[6] Ju. N. Troﬁmov et al., Conf: 6. All-Union Conf. on Neut. Phys., Kiev, 2-6 Oct. 1983, 1983; 2:142.
[7] M. Lindner et al., J. Nucl. Sci. Eng. 1976; 59:381.
[8] D. C. Stupegia et al., J. Nucl. Sci. Eng. 1967; 29:218.
[9] O. Iwamoto et al., Nucl. Data Sheets, 2016; 131:259-288.
[10] FAST ComTec MPA4T, https://www.fastcomtec.com/products/mpa/mpa4t0/. [Accessed 202001-05.][11] T. Katabuchi et al., Nucl. Ins. and Meth. in Phy. Res. A, 2014; 764:369-377.
[12] T. Sato et al., J. Nucl. Sci. Technol. 2018; 55:684-690.
[13] D.A.Brown et al., Nucl. Data Sheets. 2018; 149:1.

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JAEA-Conf 2020-001

24 New Methods to Reduce Systematic Uncertainties ofCapture Cross Section Measurement Using a Sample
Rotation System
Y. KODAMA1∗, T. KATABUCHI1, G ROVIRA1, H. NAKANO1, K. TERADA1,
A. KIMURA2, S. NAKAMURA2, S. ENDO2
1) Tokyo Institute of Technology
2) Japan Atomic Energy Agency
author: kodama.y.ae@m.titech.ac.jp

Abstract
New methods to reduce systematic uncertainties of capture cross section measurement using a sample rotation
system have been developed. Theoretical and experimental tests of these methods were conducted. Theoretical
study using a Monte Carlo simulation code was performed. The calculated results were compared with test
experimental results. The test experiment was carried out in Japan Proton Accelerator Research Complex.

I.

Introduction

Accurate nuclear data for neutron-induced reactions are necessary for the design of nuclear transmutation system to reduce minor actinides (MA) and long lived ﬁssion products (LLFP) contained innuclear waste. However current uncertainties of nuclear data such as MA and LLFP do not fulﬁll requirement for the design of transmutation facilities. Measurement of the neutron capture cross sectionsof MAs is ongoing at the Accurate Neutron Nucleus Reaction Measurement Instrument (ANNRI) in
the Materials and Life Science Experimental Facility (MLF) of the Japan Proton Accelerator ResearchComplex (J-PARC).
Neutron capture cross section σ [cm2 ] is determined in experiments based on the following equation:
1 ϵγ Nγ,(1)σ=nt ϵn Nnwhere n [atoms/cm3 ] is the atomic density of the sample, t [cm] is the sample thickness, ϵγ andϵn are the efﬁciencies of γ-ray and neutron detectors, and Nγ (Nn ) is the number of detected γ-rays(neutrons) with the γ-ray detector (neutron detector). The capture cross section can be determined fromremeasured Nγ and Nn . In the determination of capture cross section, the systematic uncertainty ofﬁnal cross section is governed by two factors: the incident neutron energy spectrum and normalization
to a standard value at a certain neutron energy.
In ANNRI experiments, the incident neutron energy spectrum is determined by measuring 478 keV
γ-rays from the 10 B(n,αγ)7 Li reaction. Detected γ-ray counts are converted to the numbers of neutrons
using the reaction rate of 10 B(n,αγ)7 Li reaction calculated from the cross section of the 10 B(n,αγ)7 Li
reaction. The energy dependence of the reaction rate depends on the atomic area density of 10 B in
the boron sample because the neutron self-shielding factor increases with the 10 B area density andalso changes with the neutron energy. Thus, boron sample thickness (10 B atomic area density) is very
important to determine the incident neutron energy spectrum.
The saturated resonance method is a commonly-used technique to obtain the absolute cross section[1]. This technique is based on the fact that neutron capture yield becomes equal to the number of theincident neutrons at a strong resonance when the sample is very thick and the resonance is fully- 127 -

JAEA-Conf 2020-001

saturated. Although this technique allows for determining the absolute cross section value without
any nuclear data as standard, it requires a thick sample that is sometimes not available.
In the present work, we suggest two new methods to reduce systematic uncertainties related to thetwo factors above using a sample rotation system.

II.

Methodology

1. Principle
Two methods employ change of the self-shielding effect with sample rotation angle. When a sample
is tilted with respect to the beam axis, effective thickness of the sample becomes larger than the actual
thickness. The reaction yield at the tilted angle θ including the self-shielding effect is expressed asfollows:)
σcap (
1(2)
ϕ 1 − e Ntσtot cosθ ,Yθ = cσtot

where c is correction factor of multiple-scattering effect and, σcap and σtot are the capture and total crosssections. The two new methods suggested below are based on the yield change with sample rotation.

2. Method 1: Boron sample thickness determination
The ﬁrst method is for thickness determination of 10 B sample which is used for measurement of
the incident neutron spectrum in ANNRI experiments. The ratio of the reaction yield at a rotation
angle of 0◦ , Y0◦ to yield at θ, Yθ ◦ is written as:
1

R( T ) =Y0◦
1 − e Ntσtot cosθ=Yθ ◦1 − e Ntσtot(3)

The energy dependence of the yield is measured by the neutron time-of-ﬂight (TOF) method. Thus,
the yield ratio R( T ) is explicitly written as a function of TOF T.
The reaction yields of 10 B(n,αγ)7 Li for different boron thicknesses were calculated using a Monte
Carlo simulation code. A typical calculated result of the yield ratio R(t) of 0◦ to 45◦ is shown in Fig. 1.
The yield ratio R(t) is equal to 1 at low energies (slow TOF) and increases with neutron energy until
1/ cos θ that is 1.41 for 45◦ . The transient TOF region between the two constant values 1 and 1.41
changes with the sample thickness. We deﬁne Thal f as the TOF value where R(t) becomes the half ofthe maximum. Thal f changes with the sample thickness. In other words, the sample thickness can be
determined from Thal f . Figure 2 shows a plot of Thal f vs sample thickness.

Thickness

Thalf

Figure 1: Reaction ratio TOF spectrum

Figure 2: Plot of Thal f vs TOF- 128 -

JAEA-Conf 2020-001

3. Method 2: Capture reaction rate determination
The second method suggested in this work is to enable absolute normalization for capture crosssection measurement even when a sample is not thick enough for the saturated resonance method.
According to Eq. 3, the absolute reaction rate that is the ratio of capture yield to the number of theincident neutrons can be determined from a change of resonance peak height when the sample is tilted
with respect to the neutron beam axis. Rotational change of the peak height of the ﬁrst resonanceof 237Np of the sample used in test experiments described in the next section was calculated. A plot of
calculated resonance peak height vs 1/ cos θ is shown in Fig. 3. The resonance peak height is normalized
to one at an angle of 0◦. The self-shielding factor can be determined from this curve.

Figure 3: Ratio at the resonance peak

III.

Simulation and Experiments

We performed simulation studies using the Monte Carlo simulation code PHITS [2]. In simulation
to test Method 1, the reaction rates of 10 B(n,α)7 Li in 10 B 90% enriched B4 C sample irradiated withneutron beam was calculated. Calculations were made for different sample thicknesses and Thal fdeﬁned in the previous section were derived from the calculations. The obtained results are shown in
Fig. 4. The simulation results are plotted as points. The curve is a ﬁtting to the simulation results. This
plot gives relation between Thal f and the sample thickness that can be used to determine the sample
thickness from Thal f measured in experiments.
In simulation for Method 2, neutron capture reaction rates of 237 Np were calculated. The total
mass and dimensions of the 237 Np sample was the same as the test experiments. Calculations weredone for different tilted angles and the height of the ﬁrst resonance peak was obtained. The peakheight normalized to an angle of 0◦ is plotted as a function of 1/ cos θ shown as Fig. 5.
The sample rotation measurement was carried out at ANNRI in MLF of J-PARC. The 237 Np
sample was tilted with a sample rotation system. A 237 Np sample (5.2 MBq) was irradiated with a neutron beam from a spallation neutron source of MLF. A TOF method was employed in this experimentwith a neutron ﬂight path of 27.9 m. Neutron capture γ-rays emitted from the sample were detectedwith a NaI(Tl) detector. After background subtraction, the peak heights of the ﬁrst resonance of 237 Np(En = 0.49 eV) were derived for different tilted angles. The experimental results are plotted in Fig. 5.
The simulation and experimental results do not agree well.

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JAEA-Conf 2020-001

Figure 4: Calibration curve for the boron sample thickness

Figure 5: Reaction yield ratio at effective thickness of eachtilted angle θ

IV. Summary
In order to reduce systematic uncertainties of neutron capture cross section measurement, two new
methods using a sample rotation system were proposed. In the ﬁrst method, the thickness of a boron
sample used to measure the incident neutron energy spectrum can be determined precisely. In thesecond method, the absolute capture reaction rate can be determined even when the sample is not
thick enough for the saturated resonance method. Monte Carlo simulations were carried out and, forthe second method, compared with results of a test experiment. Results showed that the calculationand the experimental results do not agree well. The cause of the disagreement is not yet clear. Moreinvestigation and study are necessary to improve agreement.

References[1] Richard L. Macklin et al. ,Nuc. Ins. and Meth. 164, 1, 213-214(1979).
[2] Tatsuhiko Sato et al. ,J. Nucl. Sci. Technol. 55, 684-690(2018).

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25

Proton Spectra with Low-Energy-Threshold from 40- and 70-MeV
Proton-Induced Reactions
Yuji YAMAGUCHI1, Toshiya SANAMI2, Yusuke KOBA3 and Yusuke UOZUMI1
1

Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University
744 Motooka, Nishi-ku, Fukuoka-shi, Fukuoka-ken 819-0395 Japan
2

High Energy Accelerator Research Organization

1-1 Oho, Tsukuba-shi, Ibaraki-ken 305-0801 Japan
3

National Institute for Quantum and Radiological Science and Technology
4-9-1 Anagawa, Inage-ku, Chiba-shi, Chiba-ken 263-8555 Japane-mail: yyamaguchi@nucl.kyushu-u.ac.jp

We have developed a low-energy-threshold detector consisting of a Bragg curve counter (BCC),
two silicon-surface barrier detectors and a bismuth germanate scintillator to obtain experimental
double-differential cross section (DDX) data on (p, p’x) reactions for incident energies of several tens of
MeV. The BCC can identify charged particles by itself and can do so for protons down to 1 MeV. The
developed detector is used to measure proton spectra from natC, 27Al, natCu, natAg, natTa, and 197Au targets atthe detection angles in 60 – 150° with incident energies of 40 and 70 MeV. The minimum energy in the
spectrum data is 1.3 MeV, which gives spectra in the whole energy range. Measured spectra are comparedwith the calculation of intranuclear cascade plus evaporation process.

1.

Introduction
Energy and angular distributions of secondary particles from energetic proton-nucleus reactions

are required to estimate the spatial distributions of energy deposition and radiation damage for theengineering design of an accelerator driven system and a particle radiation therapy facility. Because theestimation is performed using a Monte Carlo simulation code such as PHITS [1], the nuclear reaction
models must have high predictive power in terms of energy and angular distributions. A two-stage model,
which consists of an intranuclear cascade (INC) model [2, 3] and an evaporation model [4, 5], is generally
used to describe the emission of secondary particles from proton-nucleus reactions up to several hundredsof MeV. In recent studies [6, 7], the INC model followed by the generalized evaporation model (GEM) has
been improved for (p, p’x) reactions on light-to-heavy targets below 100 MeV. According to Ref. [7], thecalculation results of the INC model covering high-energy range mainly above 10 MeV reproduce theenergy and angular distributions of experimental data for targets from 12C to 209Bi at angles between 15°
and 130°, but the GEM results covering low-energy range mainly below 10 MeV show discrepancies forheavy targets (A ൒ 120). To improve the GEM, systematic experimental data covering low-energy range

for (p, p’x) reactions are required for not only heavy targets but also light and medium targets at various
angles from forward to backward.

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Bertrand et al. [8] and Harada et al. [9] have reported experimental data on light-to-heavy targets
for incident proton energy Ep < 100 MeV. Bertrand et al. obtained proton spectra for Ep = 29, 39, and 62
MeV, while Harada et al. obtained spectra for Ep = 42 and 68 MeV. However, the spectra on a 197Au target
exhibit different tendencies around 55° and 120° in the secondary energy range of 2 – 6 MeV between thedata for Ep ~ 65 MeV. In addition, both sets of data are insufficient below 10 MeV for Ep ~ 40 MeV. The
discrepancy at 2 – 6 MeV should be investigated and insufficient coverage of spectra for Ep ~ 40 MeV
should be updated for improvement of the GEM. Thus, new experimental data of (p, p’x) reactions aredesired for secondary energies down to 2 MeV for Ep = 40 and 70 MeV.
In this study, we take the data which satisfy the requirement above with various targets (natC, 27Al,
nat

Cu, natAg, natTa, and 197Au). For the measurement, we develop a low-energy-threshold detector consistingof a Bragg curve counter (BCC) [10], two silicon-surface barrier detectors (SSDs) and a bismuth
germanate (BGO) scintillator.

2.

Experiment
The experiment was performed at the cyclotron facility of National Institute of Radiological

Sciences. The experimental setup and the low-energy-threshold detector are described in detail in Ref. [11].
Outlines of the setup and the detector are described herein. The plan view of the experimental setup is
shown in Fig. 1. A scattering chamber was connected directly to the beam duct of the cyclotron andevacuated to less than 10-3 Pa. Incident protons hit a thin-film target located inside the scattering chamberand entered a Faraday cup consisting of a stainless-steel pipe and a graphite beam dump. The targets fornat

C, 27Al, natCu, natAg, natTa, and 197Au were mounted on a target changer. Thicknesses of 11, 5, 3, 3, 3, and

0.5 m were chosen for polyethylene, aluminum, copper, silver, tantalum, and gold foils, respectively. For
the natC target, a 134-m-thick graphite foil was also used to remove the proton component scattered byhydrogen in the polyethylene. Energy spectra of secondary particles emitted from the target were measured
at 60°, 120°, and 150° in the laboratory system using the low-energy-threshold detector.
The schematic drawing of the low-energy-threshold detector is shown in Fig. 2. This detector

Figure 1. Plan view of experimental setup. Incident protons come from the left side of this view.

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JAEA-Conf 2020-001

consists of a BCC, two SSDs, and a BGO scintillator. The BCC is a parallel-plate gas ionization chamberwith a grid. The BCC can determine the energy EBCC and the atomic number of a low-energy charged
particle from the integral and peak height, respectively, of the anode signal, which reflects the charge
distribution along the trajectory of the charged particle [10]. The Ar + 10 % CH4 counting gas is kept
inside the cylindrically shaped chamber using O-rings. The electrodes of the BCC consist of a cathode,nine field shaping rings, a grid, and an anode. The cathode-grid and grid-anode distances are 150 mm and 5mm, respectively. The cathode is a stainless-steel disk with a central hole 10 mm in diameter covered witha 2.2-m-thick aluminized Mylar film. The Mylar film is electrically connected to the stainless-steel diskon the aluminized surface and is thin enough for low-energy charged particles to enter the BCC. The fieldshaping rings and the grid are arranged at equal intervals and connected with a chain of 1-M resistors tomaintain a uniform electric field. The anode plate is a stainless-steel disk with a central hole 32 mm indiameter and covered with a 5.7-m-thickaluminum foil. Thus, the anode enables
energetic secondary particles to penetrate withsmall energy loss. The energetic particles are
identified using E – E method with the BCC
and rear detectors. In this case, the BCC works
as a transmission detector.
The measured data were corrected toremove the effects of background component
using data obtained without a target. The

Figure 2. Schematic drawing of low-energy-thresholdenergy losses in the target, entrance window,

detector. Secondary particles come from the left side ofand anode were compensated for by shifting

this drawing and pass through the entrance window.

the energy scale by calculation using the SRIM
code [12]. Finally, DDXs were obtained by
d� 𝜎𝜎
��𝐸𝐸�
�
,
dΩd𝐸𝐸 𝑠𝑠𝑠𝑠𝑠𝑠𝑠𝑠𝑠(1)

where Y(E) is the number of charged particles, s is the number of target atoms per unit area,  is thenumber of incident protons,  is the solid angle, and E is the energy bin width. The solid angle wasdetermined experimentally by -particle counting using an 241Am check source placed instead of thetarget.

3.

Results and discussion
Figure 3 shows the plot of EBCC versus Bragg peak height (atomic number) obtained using 53.3

kPa counting gas. In this figure, hydrogen and helium isotopes are identified. On the high-energy side,
those particles that penetrated the anode are observed within dashed circles. In the energy range from 1
MeV to 3 MeV, the hydrogen isotopes of protons, deuterons and tritons are clearly identified because the

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JAEA-Conf 2020-001

Bragg peak is characterized by the mass number of the charged particle besides the atomic number. Protons
down to 1.1 MeV are separated from other particles in Fig. 3. Thus, the threshold energy of 1.1 MeV wasachieved for proton identification.
Figure 4 shows the plot of E (BCC) versus E (first SSD) obtained using 53.3 kPa counting gas.
The hydrogen isotopes are also identified in this figure because of the satisfactory energy resolution of the
BCC. Although the identified proton line overlaps partially with the folding-back lines of the hydrogenisotopes that penetrated the first SSD, the penetrating events can be removed using the signal of the second
SSD in offline analysis.
Figure 5 shows proton spectra for 40- and 70-MeV incident protons on natC, 27Al, natCu, natAg,
nat

Ta, and 197Au targets at 60 – 150°. The measured spectra were obtained in a wide energy range. The

minimum energy of the measured data is 1.3 MeV, which is 0.2 MeV above the particle identificationthreshold because of compensating for the energy losses in the target foil and the entrance window. In
comparison of the natCu(p, p’x) spectra with the natTa(p, p’x) spectra below 10 MeV for Ep = 70 MeV, thenatCu(p, p’x) spectra show a broad peak whereas the natTa(p, p’x) spectra show monotonic decrease as the

proton energy decreases, which reflects the difference in the Coulomb barrier. This target dependence isalso observed in comparison of natAg(p, p’x) spectra with the 197Au(p, p’x) spectra for Ep = 40 MeV.
Calculation results and spectra of JENDL-4.0/HE reproduce measured spectra for natC, 27Al andnat

Cu targets below 10 MeV. However, calculations and JENDL spectra overestimate measured natTa(p, p’x)

spectra for Ep = 70 MeV. The 197Au(p, p’x) spectra of the INC plus GEM have a threshold at 6 MeV for Ep= 40 MeV whereas there are measured data below 6 MeV. This discrepancy is due to classicalapproximation in GEM calculation.

Figure 3. Plot of EBCC versus Bragg peak height. Figure 4. Plot of E (BCC) versus E (first SSD).
The particles identified as proton, deuteron, triton, The particles identified as proton, deuteron, triton,
3

He and  are shown.

3

He and  are shown.

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JAEA-Conf 2020-001

Figure 5. Proton spectra for 40- and 70-MeV incident protons on natC, 27Al, natCu, natAg, natTa, and 197Au
targets at 60 – 150°. Measured spectra are shown using closed circles with bar indicating statisticaluncertainties. Calculations of INC plus GEM in Ref. [7] and that in PHITS as the default model [2] areshown with solid and dashed histograms, respectively. Smooth curves show spectra of JENDL-4.0/HE.

4.

Conclusion
We developed a low-energy-threshold detector consisting of a BCC, two SSDs, and a BGO

scintillator to obtain experimental data on (p, p’x) reactions covering the low-energy range. We used the

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JAEA-Conf 2020-001

detector and obtained proton spectra down to 1.3 MeV at 60 – 150° for targets of natC, 27Al, natCu, natAg,
nat

Ta, and 197Au using 40- and 70-MeV protons. The resultant spectra were compared with results of the

INC plus GEM and spectra of JENDL-4.0/HE. Below 10 MeV, calculation results and JENDL spectra
reproduced measured spectra for natC, 27Al and natCu targets, but overestimated for the natTa target.

Acknowledgement
This work was conducted at Joint-use Research Facility for Collaborative Project with
NIRS-Cyclotron.

References[1] T. Sato, Y. Iwamoto, S. Hashimoto, et al. Features of particle and heavy ion transport code system (PHITS)
version 3.02. J. Nucl. Sci. Technol. 2018;55(6):684-690.
[2] A. Boudard, J. Cugnon, J. C. David, et al. New potentialities of the Liege intranuclear cascade model forreactions induced by nucleons and light charged particles. Phys. Rev. C 2013;87:014606.
[3] Y. Sawada, Y. Uozumi, S. Nogamine, et al. Intranuclear cascade with emission of light fragment code
implemented in the transport code system PHITS. Nucl. Instrum. Methods Phys. Res. B 2012;291:38-44.
[4] S. Furihata, T. Nakamura. Calculation of nuclide productions from proton induced reactions on heavy targets
with INC/GEM. J. Nucl. Sci. Technol. 2002;39(2):758-761.
[5] J. Benlliure, A. Grewe, M. de Jong, et al. Calculated nuclide production yields in relativistic collisions offissile nuclei. Nucl. Phys. A 1998;628:458-478.
[6] Y. Uozumi, T. Yamada, S. Nogamine, et al. Intranuclear cascade model including collective excitations andtrajectory deflections for (p, p’x) reactions around 50 MeV. Phys. Rev. C 2012;86:034610.
[7] Y. Uozumi, T. Yamada, M. Nakano. Intranuclear cascade model for 50-MeV-region (p, p’x) reactions over awide target mass range. J. Nucl. Sci. Technol. 2015;52(2):264-273.
[8] F. E. Bertrand, R. W. Peelle. Complete hydrogen and helium particle spectra from 30- to 60-MeV proton
bombardment of nuclei with A = 12 to 209 and comparison with the intranuclear cascade model. Phys. Rev.
C 1973;8:1045-1064.
[9] M. Harada, Y. Watanabe, Y. Tanaka, et al. Light charged-particle production in proton-induced reactions on
C, 27Al, 58Ni, 90Zr, 197Au, and 209Bi at 42 and 68 MeV. J. Nucl. Sci. Technol. 2002;39(2):393-396.

12[10] T. Sanami, M. Hagiwara, T. Oishi, et al. A Bragg curve counter with an active cathode to improve theenergy threshold in fragment measurements. Nucl. Instrum. Methods Phys. Res. A 2008;589:193-201.
[11] Y. Yamaguchi, T. Sanami, Y. Koba, et al. Low-energy-threshold detector for measuring proton spectra atseveral tens of MeV using Bragg curve spectroscopy. Nucl. Instrum. Methods Phys. Res. A
2020;953:163158.
[12] J. F. Ziegler, M. D. Ziegler, J. P. Biersack. SRIM – The stopping and range of ions in matter (2010). Nucl.
Instrum. Methods Phys. Res. B 2010;268:1818-1823.

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26 Charged particle emission reactions induced by 100-MeV/u 12C ionsKazuhito YOSHIDA1*, Yuji YAMAGUCHI1, Ryota IMAMURA1, Tatsuki KAKEBAYASHI1,
Mizuki YAMADA1, Yuki FUKUDA1, Tomoe KATAYAMA1, Toshiya SANAMI2,
Yosuke IWAMOTO3, Naruhiro MATSUFUJI4, Yusuke KOBA4, Zviadi TSAMALAIDZE5,Petr EVTOUKHOVITCH5, and Yusuke UOZUMI1
1

Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University
2

High Energy Accelerator Research Organization
3

4

Japan Atomic Energy Agency

National Institute for Quantum and Radiological Science and Technology
5

Joint Institute for Nuclear Research, Dubna, Russia
e-mail: yoshida@nucl.kyushu-u.ac.jp

The measurement was performed at the Heavy Ion Medical Accelerator in Chiba (HIMAC),
National Institute for Radiological Sciences, Japan. The carbon-ions were accelerated to 100 MeV/u andbombarded the target (C, Al, and Co). Emitted charged particles were detected by counter-telescopes installedat the PH2. Light particles (p, d, t. 3He and α) were detected with two E-E telescopes comprising two siliconsurface-barrier detectors (SSDs), a GSO(Ce) crystal and four PWO crystals. Particles heavier than α weredetected with two E-E telescopes consisting of two SSDs and a CsI(Tl) crystal. Thanks to their high energy
resolutions, isotope separation was achieved for many of detected particles. This report describes
experimental results obtained during 2018-2019 at HIMAC.

I.

INTRODUCTION

With respect to a role in radiotherapy, carbon-ions offer excellent advantages [1]. One is the highrelative biological effectiveness (RBE) thanks to its high linear energy transfer (LET). Another advantage ofcarbon-ion therapy is that it provides highly conformal dose distributions due to the Bragg peak, and therefore
it is possible to deliver a large and uniform dose to the target while sparing normal tissues. However, there isa certain potential that they induce undesired dose outside the primary beam field. The undesired dose arises
from secondary particles produced through beam interactions with the patient's body. These dose distributions
provide a low integral dose to surrounding healthy tissues. The low dose exposure of the normal tissue raised
concerns about the occurrence of secondary malignancies as the long-term effects.
When a primary carbon ion traverses the patient body, it can undergo nuclear fragmentation.
According to the breakup model [2] and the abrasion-ablation model [3,4], highly energetic particles
including clusters, which have approximately the same velocities as that of the projectile, are observed in an
extreme forward cone in the laboratory angles. Therefore, they are responsible for the dose tail observed after- 137 -

JAEA-Conf 2020-001

the distal edge of the Bragg Peak [5,6]. In order to understand the dose tail, many experiments were carriedout. The energy-angle double differential cross sections (DDXs) of charged particle productions were alsomeasured [7,8] at angles smaller than 20 degrees including zero degrees. In contrast, little attention has been
paid to DDXs of the lateral angles. The charged composite particles, which have high LET and large RBE,
emitted laterally have been supposed to have low energies [3,4], and to be stopped within the primary beam
diameter. Then, the importance of lateral doses was expected to be negligible. In recent years the increasing
concern about low dose exposure requires accurate estimation of lateral doses. However, DDXs of lateralemission of charged composite particles were not reported at larger angles than 20 degrees.
The purpose of this work was to measure DDXs of charged particle emissions to lateral directions
from interaction between clinical carbon-ion beam and elements constituting the human body. The
measurement covered charged particles from protons to 12C, and DDXs were determined for most of theisotopes. In this report, we describe the experimental equipment and method, and present some preliminary
DDX data measured at the beam energy of 100 MeV/u.

II. EXPERIMENTS
Table 1 Targets used in this work.
27

Thickness [mm]

Al

nat

59

0.1 and 2.0(for calibration)

0.1

0.05

C

Co

Experiments were carried out at the Heavy Ion Medical Accelerator In Chiba (HIMAC), National
Institute for Radiological Sciences, Japan. 12C beam was focused on a target in a spot of about 5-mm diameter.

The target was set at the center of a vacuum chamber of 50-cm diameter. The targets used were 12C, 27Al, and59

Co, which are all natural metals. Their thicknesses are listed in Table 1.

Fig. 1 Telescope for p, d, t, 3He, and 

Fig. 2 Telescope for particles heavier than - 138 -

JAEA-Conf 2020-001

Fig3. A system for measurement the light

Fig4. A system for measurement the heavy

particles

particles

The energy of emitted particles was measured with two stacked scintillator spectrometer which are shown
schematically in Fig. 1. Light particles (p, d, t. 3He and α) were detected with two E-E telescopes comprising
two silicon-surface-barrier detectors (SSDs), a GSO(Ce) crystal and four PWO crystals. Particles heavier
than α were detected with two E-E telescopes consisting of two SSDs and a CsI(Tl) crystal. The first andsecond SSD were E detectors of 100 m and 2 mm in thickness, respectively. The GSO(Ce) crystal was
cubic with an edge length of 43mm. The PWO crystal was cuboidal with 50 × 50 × 199mm. The CsI(Tl)scintillator was cubic with an edge length of 28mm.

Fig. 5 Two dimensional scatter plot of first and second SSD
At first, DDX of the Al(C,px) reaction was determined in a proton energy range of 70-120 MeV at
30 and 45 degrees. The thick target (2-mm-thick 27Al plate) and the low intensity beam were used in thesemeasurements, where the number of incident 12C ions was counted directly by a plastic scintillator detector.
Second, experiments with a thin target (0.5-mm-thick 27Al plate) and the high-intensity beam were
performed for the Al(C,px) reaction to measure proton spectra over a range of 10-500 MeV. The spectra wereconverted to DDXs by normalizing to the thick target data. During these measurements, a beam monitor
consisting of a plastic scintillator detector was used to count the secondary particles scattered from a thinmetal foil. The readout of the beam monitor was also normalized to that of the counter used for thick-target
experiments. Lastly, the DDX values for other reactions were measured with a thin target, a high intensity
beam, and the beam The telescope which used for measuring light particles (p, d, t. 3He and α) had sufficient- 139 -

JAEA-Conf 2020-001

depth to stop 480-MeV protons and 1200-MeV -particles. Figure 5 shows example data measured by thetelescope for the heavy particles. The telescope which was used for measuring particles heavier than  had
sufficient depth to stop 730-MeV 6Li ions and 1200-MeV 12C ions. One of the faces of the GSO(Ce)scintillator and the PWO scintillator were each connected to a photomultiplier tube (PMT) and that of theCsI(Tl) scintillator was connected to a photo diode to convert scintillation light into electric signals andamplify the signal. Electric signals from detectors were fed into a spectroscopic amplifier via a preamplifier.
Their pulse heights were analyzed by an amplitude-to-digital converter. The digitized data were transferred
to a PC through the CAMAC system and recorded on hard-drive. Energy calibration was performed byreferring to proton energy which was calculated by the Bethe formula. The spectrometers which were used
for measuring light particles was installed outside the chamber. The telescopes which were used formeasuring the heavy particles were installed into the vacuum chamber. Energy spectra of light particles were
measured at laboratory angles from 30◦ to 120◦. Energy spectra of the heavy particles were measured atlaboratory angles from 15◦ to 20◦.
High-energy particles may undergo nuclear reactions before being stopped or may scatter out of the crystal.
The ratio between the number of stopped particles and the total number of particles is referred to as the peakefficiency of the spectrometer and is necessary to determine absolute cross sections. Thus far, the peakefficiency has been determined as a function simulated by PHITS calculation. The DDXs are determined by
𝑑𝑑 � 𝜎𝜎𝑌𝑌�,
𝑑𝑑𝑑𝑑𝑑𝑑𝑑 𝑃𝑃𝑃𝑃� 𝜙𝜙𝜙𝜙𝜙𝜙𝜙

where ε is the energy bin width required for data reduction,  is the solid angle of the detector, P is thepeak efficiency, St is the surface density of the target, and  is the total charge of the incident 12C beam.
The number of emitted particles in , which is represented by Y, is obtained by the particle identification(PI) technique. The quantity PI is given by�PI � 𝐸𝐸���� �𝐸𝐸��� � 𝜙𝐸𝐸�� ,where Etot is the total energy deposited in the detector and E is the energy measured in the E detectors.
The parameter b is set 1.73, which is the optimum value to separate the isotopes. A two dimensional plot of PI
versus emission energy of Be and B was shown in Fig.6.

B
Be

Fig. 6 Two-dimensional plot of PI versus particle energy (Be and B).

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JAEA-Conf 2020-001

III. RESULT AND DISCUSSION
Preliminary data have been obtained for the reaction of 100-MeV/u 12C ion on the 27Al target. The
spectra of DDXs for 100-MeV/u 27Al(12C, px) and 27Al(12C, x) reactions at 30, 45, 60, 120 degrees areshown in Fig.7 and Fig.8, respectively. The spectra of DDXs for 100MeV/u 27Al(12C, 7Bex) reactions at 15
and 20 degrees are shown in Fig.9. The maximum energy of proton was 210 MeV at 60degrees. The highest
observed energy of -particle was 230 MeV at 60 degrees. The highest observed energy of 7Be was 490 MeV
at 20 degrees. -particle of 230 MeV has the range of 28.5 mm in water. Therefore, healthy tissues aroundthe beam axis to cancerous tissues can be affected by such high-energy particles. As shown in Fig.6, 470
MeV -particles were measured at 30degrees. In previous study, 285 MeV proton was observed for 400 MeV
nat

C(, px) reactions at 30 degrees. Considering secondary nuclear reactions, the influence of charged particle

emission over 30 degrees for a human cannot be ignored.

Fig. 7 DDXs of 27Al(12C, px) reaction

Fig. 8 DDXs of 27Al(12C, x) reaction

Fig. 9 DDXs of 27Al(12C, 7Bex) reaction- 141 -

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IV. CONCLUSION
We measured DDXs of charged particle emission reactions induced by 100MeV/u 12C ions on threetargets (C, Al, Co) at 15, 20, 30, 45, 60, and 120 degrees at HIMAC, NIRS. DDXs of emitted particles (p, d,t, 3He,  6Li, 7Li, 7Be, 9Be, 10Be, 10B, 11B, 10C, 11C, 12C) were determined in wide ranges of emission energies.
The notable point is that high-energy particles are observed at large angles. Particularly, particles above 100
MeV show certain values of DDXs. This fact suggests that the high-energy heavy-ions can have a long range
over the beam diameter, and the healthy tissues around the initial beam axis may have late effects low-dose
exposure of heavy ions.

ACKNOWLEDGMENTS
This work was conducted under the research project with heavy ions at NIRS-HIMAC, program
number H350. This work was supported by JSPS KAKENHI Grant Number JP17H03522.

REFERENCES[1] D. Schardt, T. Elsässer, and D. Schulz-Ertner, “Heavy-ion tumor therapy: Physical and radiobiologicalbenefits,” Rev. Mod. Phys. 82, 383 (2010).
[2] J. R. Wu, C. C. Chang, and H. D. Holmgren, “Breakup of α Particles in the Fields of Nuclei,” Phys. Rev.
Lett. 40, 1013 (1978).
[3] R. Serber, “Nuclear reactions at high energies,” Phys. Rev: 72; 1114–1115, (1947).
[4] J. Hüfner, K. Schӓfer, B. Schürmann, “Abrasion-ablation in reactions between relativistic heavy ions,”
Phys. Rev. C12, 1888 (1975).
[5]N. Matsufuji, A. Fukumura, M. Komori et al. “Influence of fragment reaction of relativistic heavy charged
particles on heavy-ion radiotherapy,” Phys. Med. Biol., 48. 1605-1623, (2003).
[6] N. Matsufuji, T. Kanai, N. Kanematsu, T. Miyamoto, M. Baba, T. Kamada, H. Kato, S. Yamada, J. Mizoeand H. Tsujii, “Specification of Carbon Ion Dose at the National Institute of Radiological Sciences (NIRS),”
J. Radiat. Res: 48; Suppl. A:A81-6 (2007).
[7] M. De Napoli et al., “Carbon fragmentation measurements and validation of the Geant4 nuclear reaction
models for hadrontherapy,”, Phys. Med. Biol. 57, 7651-7671 (2012).
[8] M. Toppi et al., “Measurements of 12C ion fragmentation on thin carbon target from the FIRST
collaboration at GSI,” Jour. Phys. Conference Series 590, 012035 (2015).

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27 Measurement of Neutron Energy Spectra of
345 MeV/u 238U Incidence on a Copper Target
Kenta SUGIHARA1,2 , Nobuhiro SHIGYO1 , Atsuko AKASHIO2 and Kanenobu TANAKA2
1. Department Applied Quantum Physics and Nuclear Engineering, Kyushu University
Motooka, Nishi-ku, Fukuoka, 819-0395, Japan
2. RIKEN Nishina Center, RIKEN, 2-1 Hirosawa, Wako, Saitama 351-0198, Japane-mail: k.sugihara@kune2a.nucl.kyushu-u.ac.jp
Thick target neutron yields produced by 345 MeV/u 238 U incidence on a copper target
were measured at 45◦ and 90◦ with the time-of-flight method. The experimental results werecompared with those calculated using the JAERI Quantum Molecular Dynamics (JQMD)
model ver.1.0 followed by Generalized Evaporation Model implemented in PHITS ver.3.10.
It was demonstrated that the JQMD model underestimates measured data.

1

Introduction

Radioactive Isotope Beam Factory (RIBF) at RIKEN is an accelerator facility[1]. RIBF has an ability
to accelerate heavy ions (e.g. Ca, Kr, Xe, U) up to 345 MeV/u through superconducting ring cyclotron(SRC). It is in progress to conduct a lot of researches with the utilization of Radioactive Isotope (RI)
beams which are generated by reactions between primary beams and RI generating targets. A numberof neutrons are produced at the target and a beam dump. Therefore, areas around the target and thebeam dump is a high dose rate region.
Radiation shielding around the high dose rate area was basically designed by the Moyer model[2].
As a source term of the model, neutron thick target yield (TTY) by 345 MeV/u 238 U beam is essential.
There is an available data of TTY for 1 GeV/u 238 U incidence on iron measured at Gesellschaft für
Schwerionenforschun (GSI)[3]. Neutron energy spectra for forward directions were measured in the study.
As RIBF is located in the underground, measured data for side directions are of great importance. Thereis no measured data of TTY from 345 MeV/u 238 U incidence on copper, which is beam dump material
at RIBF.
Particle transport code PHITS[4] has been recently applied for radiation shielding design. Measured
data for examining the accuracy of physics models in the code is not sufficient. Experimental data ofneutron TTY is highly desired for verifying neutron emissions to optimize radiation shielding design.
In this study, neutron TTYs at 45◦ and 90◦ were measured from 345 MeV/u 238 U incidence on
copper. Simulation TTYs are calculated by JAERI Quantum Molecular Dynamics (JQMD)[5] modelver.1.0 followed by Generalized Evaporation Model[6] implemented in PHITS ver.3.10. These TTYs arecompared to validate how accurate the physics model is.

2

Experiment

The experiment was carried out at Zero Degree Spectrometer of RIBF. The experimental arrangement
is illustrated in Figure 1.
The incident energy of 238 U beam was 345 MeV/u. The beam intensity was about 106 pps. As radio
frequency (RF) of SRC was 18.5 MHz, the beam pulse came every 54 ns. A copper target with a thicknessof 10 mm was positioned at the F10 chamber. The target was three times as thick as the range of the238
U beam.
Neutrons produced in the target were measured with NE213 liquid organic scintillators. The light
output of the scintillators was recorded by charge sensitive analog-to-degital converters (ADCs). The
timing of incident particles on the detectors was registered in a time-to-degital converter (TDC). The- 143 -

JAEA-Conf 2020-001

˚

N

˚

6#FBN

 QQT$V5BSHFU
NNU

Figure 1 Illustration of the experimental setup for the TOF measurement. 345 MeV/u 238 U beamirradiated a copper target. Neutron and veto detectors were located at 45◦ and 90◦ from the beam axis.
Both flight path lengths were 4.0 m.
neutron detectors were located at 45◦ and 90◦ with respect to the beam axis. Both the diameter andlength of the detectors are 12.7 cm. The flight path lengths from the target to each neutron detector were
4.0 m as shown in Figure 1. Kinetic energies of detected neutrons were determined by the inverse timeof-flight (TOF) method[7]. Start and stop signals came from neutron detectors and RF signals of SRC,
respectively. Energies of neutrons overlapped with preceding beam pulse were less than 17.6 MeV. The
threshold level in the analysis was set to 12.3 MeVee which was two times higher than the Compton edgeof 6.13 MeV γ-ray from a 244 Cm-13 C neutron source. This corresponds to the maximum light output of
18.4 MeV neutron. These low energy neutrons were automatically removed in data analysis because theirpulse heights were smaller than the bias level. To eliminate charged particle events from total events,
an NE102A plastic scintillator of 2 mm in thickness (veto detector) was used in front of each neutron
detector.
The background neutrons scattered from the floor were measured by an iron shadow bar of 1 m long,which was installed between the F10 chamber and each neutron detector.

3000

1SPNQUĕSBZ

ADC total [ch]

2500
2000
1500
1000
500
00

500 1000 1500 2000
TDC [ch]Figure 2 TDC-ADC two-dimensional plot for the NE213 scintillator at 90◦ . Prompt γ-rays appear
about 1400 ch. Slow neutrons, the channel of which was less than about 160 ch, were not measured dueto the RF frequency. Therefore, the maximum time difference between neutrons and prompt γ-rays was
34 ns because the time resolution of TDC was 0.0274 [ns/ch].

- 144 -

JAEA-Conf 2020-001

3Analysis

Neutron energy spectra were derived from the flight time and data of pulse height distribution of
the neutron detectors[7]. Neutron events were identified with the two gate integration method[8]. After
neutron events were extracted, kinetic energies of the neutrons were obtained from the time difference
between the neutrons and prompt γ-rays in Figure 2. Neutron detection efficiency was calculated by the
SCINFUL-QMD code[9][10]. Finally, measured data were converted into neutron production TTYs.

4

Results and Discussion

TTY [n/sr/MeV/source]

The measured data of neutron TTYs were compared with the JQMD model ver.1.0 implemented in
PHITS ver.3.10. The measured and calculated neutron TTYs for 345 MeV/u 238 U bombardment on
copper are shown in Figure 3. Only statistical error in the measured data was included in Figure 3.

10-2

˚

10-3

10-4
-5

10

10-6

˚

Exp.
PHITS

102
103
Neutron Energy [MeV]

Figure 3 Neutron TTYs from 345 MeV/u 238 U incidence on copper. Blue triangles and black solidlines represent the experimental data and the results of PHITS calculation using JQMD ver.1.0 model,
respectively.
Integral of the experimental and calculated energy spectra above 40 MeV is shown in Table 1. Statistical errors are less than 1 % for both the experiment and PHITS calculation. According to Table 1, thenumber of produced neutrons for PHITS was 0.61 time as large as that of our experimental data at 45◦ .
The ratio for 90◦ was 0.44.
Table 1

Angular distribution of neutron production integrating neutron energy spectrum above 40 MeV.
Exp.
PHITS

5

45◦ [n/sr/source]
2.8 × 10−1
1.7 × 10−190◦ [n/sr/source]
2.7 × 10−2
1.2 × 10−2

Conclusion

Neutron TTYs at 45◦ and 90◦ were measured from 345 MeV/u 238 U incidence on a copper target with
the TOF method. The PHITS calculations underestimate the measured data. The experimental data
will be useful for the estimation of the source term in the Moyer model.

- 145 -

JAEA-Conf 2020-001

6

Acknowledgement

This experiment was performed at RI Beam Factory operated by RIKEN Nishina Center and CNS,
the University of Tokyo. The PHITS calculation was performed using the supercomputer system of
RIKEN.

References[1] K. Tanaka, N. Inabe, K. Yoshida et al. Evaluation of radiation levels and comparison with PHITS
calculations for the BigRIPS separator in Radioactive Isotope Beam Factory. Prog. in Nucl. Sci.
Technol. 2014;4:201-204.
[2] B. J. Moyer. Method of Calculation of the Shielding Enclosure for the Berkeley Bevatron. in Proc.
1st Int. Conf. Shielding around High Energy Accelerators, Presses Universitaires de France, Paris
1962;65.
[3] O. Yordanov, K. Gunzert-Marx, P. Adrich et al. Neutron yields from 1 GeV/nucleon 238U ion beams
on Fe target. Nucl. Instr. Methods B. 2005;240:863-870.
[4] T. Sato, Y. Iwamoto, S. Hashimoto et al. Features of Particle and Heavy Ion Transport code System(PHITS) version 3.02. J. Nucl. Sci. Technol. 2018;55:684-690.
[5] K. Niita, S. Chiba, T. Maruyama et al. Analysis of the (N, xN’) reactions by quantum molecular
dynamics plus statistical decay model. Phys. Rev. C. 1995;52(2620):2620-2635.
[6] S. Furihata. Statistical analysis of light fragment production from medium energy proton-induced
reactions. Nucl. Instr. Methods, B. 2000;171:251-258.
[7] D. Satoh, D. Moriguchi, T. Kajimoto et al. Measurement of neutron-production double-differential
cross-sections on carbon bombarded with 290-MeV/nucleon carbon and oxygen ions. Nucl. Instr.
Methods A. 2011;644:59-67.
[8] M. Moszynski, G.J. Costa, G. Guillaume et al. Identification of different reaction channels of high energy neutrons in liquid scintillators by the pulse shape discrimination method. Nucl. Instr. Methods,
A. 1994;343:563.
[9] D. Satoh, S. Kunieda, Y. Iwamoto et al. Development of SCINFUL-QMD code to calculate theneutron detection efficiencies for liquid organic scintillator up to 3 GeV. J. Nucl. Sci. Technol.
2002;39(sup2):657-660.
[10] T. Kajimoto, N. Shigyo, T. Sanami et al. Measurement of absolute response functions and detection
efficiencies of an NE213 scintillator up to 600 MeV. Nucl. Instr. Methods A. 2011;665:80-89.

- 146 -

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- 147 -

JAEA-Conf 2020-001

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JAEA-Conf 2020-001

Acknowledgement
This work was supported by JST-OPERA Program Grant Number JPMJOP172, JAPAN, Grant-in-Aid for
Scientific Research (B) (No.16H03906) and JSPS Fellows (No.19J12649) from the Japan Society for the
Promotion of Science (JSPS) and Kyushu University Platform of Inter/Transdisciplinary Energy Research (Q-PIT).
References[1] S. Manabe et al., IEEE Trans. Nucl. Sci., 65:8, 1742-1749 (2018).
[2] W. Liao et al., IEEE Trans. Nucl. Sci., 65:8, 1734-1741 (2018).
[3] D. F. Measday, Phys. Rep., 354, 243-409 (2001).
[4] S. E. Sobottka et al., Phys. Rev. Lett., 20:12, 596-598 (1968).
[5] S. Abe et al., J. Nucl. Sci. Technol., 54:1, 101-110 (2016).
[6] K. Niita et al., Phys. Rev. C, 52:26, 2620-2635 (1995).
[7] S. Furihata et al., Japan Atomic Energy Research Institute; 2001. (JAERI-Data/Code 2001-015).

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7HFKQRO    - 158 -

JAEA-Conf 2020-001

30 Integration test with a Gaseous Detector and a Solenoidal
Magnet for the Precise Neutron Lifetime Measurement
Kodai Yano1∗ , Yasuhiro Makida2 , So Makise1 , Kenji Mishima2 , Hidetoshi Otono3 ,Naoyuki Sumi1 and Tamaki Yoshioka3
1

Department of Physics, Kyushu University, Fukuoka 819-0395, Japan
KEK, High Energy Accelerator Research Organization, Tsukuba 305-0801, Japan
3
Research Center for Advanced Particle Physics (RCAPP), Kyushu University, Fukuoka 819-0395,
Japan
2

E-mail: yano@epp.phys.kyushu-u.ac.jp
The neutron lifetime (τn ) is an important parameter for particle physics and cosmology. The neutron
lifetime was measured by two major methods. One of them is called the bottle method, and the otherone is called beam method. Though, there is an 8.6 sec (4.0 σ) deviation between these two measured
results. To solve this problem, a new type of method, the electron counting method, is implemented at
BL05 MLF J-PARC using a pulsed cold neutron beam to measure the true neutron lifetime with thesame or better accuracy than the previous measurement. A Time Projection Chamber (TPC) records
both the electrons from neutron β decay and protons from the 3 He neutron capture reactions to estimate the neutron flux that enters the TPC. However, electron background signals require the largest
correction and they are the source of uncertainty for this experiment. It is confirmed by Monte Carlosimulation that a uniform magnetic field generated by a solenoidal magnet along the neutron beam
can greatly reduce this background. Hence, we proposed another experiment (LiNA experiment) using a solenoidal magnet to validate this simulation result. The detector has been produced and theintegration test with a magnet has been finished. The status of progress is reported in this paper.
KEYWORDS: neutron lifetime, time projection chamber, solenoidal coil1.

Introduction

1.1

Motivation
The neutron lifetime is one of the most important parameters for Big Band Nucleosynthesis (BBN).
The light elements, such as helium and lithium, were produced by a combination of neutrons and protons. Therefore, the number of light elements produced in the early universe depends on the number ofneutrons that are left. Moreover, since neutron decays have the advantage of no nuclear physics uncertainties, neutron lifetime is also an important parameter for the Cabibbo-Kobayashi-Maskawa (CKM)
matrix element Vud .
The averaged neutron lifetimes are 879.4 ±1.0 sec [1]. The recent neutron lifetime measurements
have been performed by two diﬀerent experimental methods. One of them is called the bottle method[2], in which neutrons are stored in a special material or magnetic bottle, and neutrons that survive after
a certain time are measured. Another method is the beam method [3], where neutrons are measured witha flux monitor while protons from neutron β decay are measured using another detector. Though, there
is an 8.6 sec (4.0 σ) deviation between the results from these two methods. A new type of measurementwith the same accuracy is therefore required to resolve the diﬀerence.
1.2

A new type of beam method
We have discussed a new type of beam method, the electron counting beam method, which gives
diﬀerent systematic errors from previous experiments. In this new method, neutron lifetimes are derived- 159 -

JAEA-Conf 2020-001

by simultaneous measurement of electrons from β decay and from 3 He neutron capture reactions recorded
by a Time Projection Chamber (TPC). As a neutron source, we use a high-intensity pulsed neutron beamprovided at the Japan Proton Accelerator Research Complex (J-PARC). This method was originallydeveloped by Kossakowski et al. [4]. The neutron lifetime τ is calculated by the following equation,)(
1 S 3 He /ε3 He
τ=
(1)
ρσν
S β /εβwhere ρ is the 3 He density, σ is the cross section of neutron capture by 3 He and ν is the neutron velocity.
Since the cross section is inversely proportional to neutron velocity, σν could be treated as constant(σν = σ0 ν0 ), so we can use the thermal velocity ν0 = 2200 m/s and cross section σ0 = 5333 ± 7 barn for
all neutron velocities. S and ϵ are the number of signals and the selection eﬃciency for each reaction,
respectively. We aim to measure the neutron lifetime with O(0.1%)(∼1 sec) accuracy using this method.
1.3

Largest correction
The largest two corrections for this experiment are the estimation of scattered background neutrons in
the beam by the TPC operation gas and selection eﬃciency for β decay electrons. The scattered neutrons
will be captured by the detector material and emit the γ ray, in then produce the electron via Compton
scattering. Because this electron has similar space, energy and time distributions to β decay electrons,
we cannot reject them by signal selection nor discriminate using the method of time of flight. Lithium
fluoride (6 LiF) can suppress the number of emitted prompt γ rays down to 0.01%. However, it is estimatedby Monte Carlo simulation that these backgrounds remain 4% of the β events, which causes a large
systematic uncertainty. We can improve the purity of the β decay signal with a tighter selection process,but larger eﬃciency corrections are required. In any case, minimizing corrections is essential for precise
measurement.

2.

Methodology

Figure 1 shows a schematic view of the experimental apparatus. The uniform magnetic field is applied along the neutron beam axis to separate β decay electrons from a background [5]. Moreover, β decay
electrons remain in the signal region. Better signal eﬃciency and a lower level correction requirement
can be achieved by decreasing these backgrounds. The magnet we use for this experiment is a superconducting solenoidal coil that was originally prepared for the BESS experiment [6]. The drift direction of
the TPC is vertically upward.

3.

Performance

We have evaluated the performance of background reduction with a Monte Carlo simulation basedon Geant4 [7]. Figure 2 shows particle tracks in the vacuum chamber as a projection to the orthogonalplane to the beam axis. Each figure in Fig. 2(a) corresponds to β decay and background with and without amagnetic field. The central boxes indicate the signal region. One can see that while only a few backgroundtracks remain (right bottom plot) under the non-magnetic field, all β decay tracks remain in the signal
region (left bottom plot) under the magnetic field. Figure 2(b) shows that the background is suppressed
to a few % compared to the case without the field. Moreover, the magnetic field recovers the β decay
signal eﬃciency, because β decay electrons do not reach the inner wall and kept in the signal region. A
magnetic field to 400 mT is enough to decrease the correction size to O(0.1%) for the neutron lifetime
measurement.

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JAEA-Conf 2020-001

Fig. 1. A schematic view of the experimental apparatus. In this setup, TPC, vacuum chamber and solenoidal
magnet are installed in order from the center. The bunched neutron beam shorter than TPC length passes throughthe center of TPC.

(a) Tracks in the vacuum chamber(b) Fraction of remain backgroundFig. 2. Background reduction performance with the solenoidal magnet was evaluated by Monte Carlo simulation. (a) Tracks in the vacuum chamber with and without magnetic field (600 mT). Central box indicate signal
region for neutron β decay signals. (b) Background reduction performance in the central signal region.

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JAEA-Conf 2020-001

4.

New TPC and integration test

4.1

Production
To distinguish the central signal and other background signals, we produced a diﬀerent type of TPC,
the TPC that has multiple layers. Chip condensers and resistances were soldered on the circuits. Drift andanode wires (Be-Cu ϕ100 µm and Au-W ϕ30 µm ) are mounted to the circuits. Figure 3 shows the final
form of new TPC. The new TPC consists of three layers and non-magnetic material, such as Aluminum,
stainless and polycarbonate resin. Most of the components such as the Aluminum frame and circuits
were designed and produced at Kyushu University.

Fig. 3. A picture of TPC we have constructed. The TPC consists of 3 drift layers, and the signal region is in the
middle layer.

4.2

Integration test with a solenoidal magnet
Figure 4 shows the overall view of the setup. We have carried out an integration test with this setup at
KEK to evaluate several detector capabilities, for instance, background reduction capabilities and energy
resolution. Let us show the result of the background exclusion test. Figure 5 shows the reduction of
background by applying a magnetic field. We incident a γ ray irradiated from a radiation source (60 Co,
137 Cs, 152 Eu) from outside the vacuum chamber to produce an electron assumed as background events.
With a magnetic field, most of the background events cannot invade the signal region, compared tothe one without the magnetic field. The background events excluded down to only a few %, which isconsistent with simulation results.

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JAEA-Conf 2020-001

Fig. 4. A picture of TPC installed inside the vacuum chamber and solenoidal magnet. The integration test withthis setup has carried out at KEK.

Fig. 5. A plot of the event entering the signal region. We used a radiation source(60 Co) for emitting an electron
assumed as a background. Left ((a) and (d)) and middle plots ((b) and (e)) show the entry events with only anenvironment (purple plot with black lines) and the environment with background events (pink plot without blacklines), with and without a magnetic field, respectively. The right plot ((c) and (f)) shows the entries of background
events. The background events are dramatically reduced in the presence of a magnetic field. The plot above ((a),(b) and (c)) and below ((d), (e) and (f)) show the experimental and simulation results, respectively.

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5.

Summary and prospect

We aim to measure the neutron lifetime with O(0.1%)(∼1 sec) accuracy by using the electron beam
method. The background in the central region can be suppressed by using a multi-layered TPC andsolenoidal magnetic field. The two largest corrections can be decreased to O(0.1%) of the neutron lifetime
according to the Monte Carlo simulation. A new multi-layered TPC was produced, and an integration
test with a solenoidal magnet at KEK has been finished. The current setup at J-PARC MLF BL05 will be
replaced by this magnet setup, and we will prepare for neutron beam operation.

Acknowledgement
This work was supported by JSPS KAKENHI Grant Number 18H01231 and 18J10106.
References[1] M. Tanabashi et al., (Particle Data Group), Phys. Rev. D 98, 030001 (2018) and 2019 update.
[2] A. P. Serebrov et al., ”Neutron lifetime measurements using gravitationally trapped ultracold neutrons”, Phys.
Rev. C 78, 035505 (2008).
[3] J. S. Nico et al., ”Measurement of the neutron lifetime by counting trapped protons in a cold neutron beam”,
Phys. Rev. C 71, 055502 (2005).
[4] K. Schreckenbach, G. Azuelos, P. Grivot, R. Kossakowski, and P. Liaud, ”Neutron decay measurements witha helium-filled time projection chamber”, Nuclear Instruments and Methods in Physics Research A 284, 120(1989).
[5] H. Otono, ”LiNA-Lifetime of neutron apparatus with time projection chamber and solenoid coil”, Nuclear
Instruments and Methods in Physics Research Section A (Volume 845, 2017, Pages 278-280).
[6] Y. Makida et al., ”Ballooning of a thin superconducting solenoid for particle astrophysics”, IEEE Transactions
on Applied Superconductivity (Volume: 5, Issue: 2, June 1995).
[7] S. Agostinelli et al., ”Geant4 - A Simulation Toolkit”, Nuclear Instruments and Methods A 506 (2003) 250303.

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JAEA-Conf 2020-001
2020-001

31

HUNS Upgrade; Simulation of the Beam Profile
for a Thermal Neutron Source in HUNS-2

Kaoru Y. HARA1)*, Fumiaki SATO2), and Takashi KAMIYAMA1)
1) Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Japan
2) School of Engineering, Hokkaido University, Sapporo 060-8628, Japan*e-mail: hara.kaoru@eng.hokudai.ac.jp

The upgrade of HUNS-2 is ongoing to increase the intensities of the electron andneutron beams. The profiles of neutron beam have been calculated using the PHITS
code to design a TMRA of thermal neutron source in HUNS-2. In the calculation, a trial
TMRA enhanced the intensity of thermal neutron beam as compared with the originalone.

1. Introduction
The electron linear accelerator in the Hokkaido University Neutron Source (HUNS[1]) facility was recently upgraded toward HUNS-2 to increase the electron beam
current. The HUNS-2 is an accelerator-driven neutron source that is equipped with acold-, thermal- and fast-neutron type of sources. Experiments such as small angle
neutron scattering [2], neutron transmission imaging [3] and neutron irradiation test
have been currently conducted using the HUNS-2 for studies of material science anddevelopments of measurement method.
In this work, the neutron beam profile for the HUNS-2 thermal neutron source wascalculated with the Monte Carlo simulation code PHITS [4] (version 2.80/ 3.02) toimprove a target-moderator-reflector-assembly (TMRA) of the original thermal neutron
source system. The TMRA consists mainly of a neutron production target with atungsten disc, lead blocks and water channels, a polyethylene moderator, graphitereflectors [5]. Through the simulations, a position and size of TMRA have been designed
for increasing the intensity of thermal neutron beam.

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2. Original HUNS thermal neutron source
In the HUNS facility, the thermal neutron source is located on the “CENTER” beamcourse. The horizontal cross-section view of the TMRA of the original HUNS thermal
neutron source is illustrated in Fig. 1. Neutrons are produced by irradiation of theneutron production target with the electron beam, where the neutron production target
is composed of a tungsten plate, lead blocks, and stainless water channels as shown in
the inset of Fig. 1. First, in the neutron production target, the bremsstrahlung photons
are generated and the photonuclear reactions are induced when the electron beam is
stopped. Second, the neutrons, which are emitted from the photonuclear reaction, arethermalized through the polyethylene moderator and graphite reflectors. Finally, thethermalized neutron beam is collimated into the perpendicular direction with respect tothe electron beam axis. To reduce the contamination of “gamma flash” on the neutron
beam course, in addition, three lead blocks are placed near the neutron production
target.
For a typical transmission experiment with the thermal neutron source, a distance
between a neutron detector and the polyethylene moderator (L) is approximately 7 m.
The integrated intensity of neutron beam that was measured at L = 7 m was 103

n/cm2/pulse in the neutron energy range of 0.01−0.1 eV [6]. The energy, current, andrepetition rate of the electron beam were 34 MeV, 35 A, and 50 Hz, respectively, at themeasurement for the neutron intensity. The schematic view of experimental setup is
shown in Fig. 2, where a 6Li glass detector [7] (or 3He gas counter) was used. An
aperture size of B4C collimator near the detector was 1 cm × 1 cm.
The original HUNS thermal neutron source is involved in a composite source system
which can alternatively produce the neutron or X-ray beam. The composite source
system has been used for developing a complementary imaging technique [3]. In thesimulation for the thermal neutron beam, therefore, the geometry of composite source
system is defined in the PHITS input file as shown in Fig 3. The neutron and X-ray
sources are located respectively at the upside and downside of the composite source
system. A more information about the configuration and usage of the composite source
system is provided in the references [3, 5]. Additional walls for radiation shields and an
up-and-down stage are ignored in the simulation while these are at outside of thecomposite source system in the actual situation. The calculated energy distribution of
the thermal neutron beam is shown in Fig. 4 by the red line. In our previous work [5], a- 166 -

JAEA-Conf 2020-001
2020-001

neutron point source with the Maxwell-Boltzmann distribution of kT = 1.3 MeV wasdefined at the neutron production target as an initial source. In this work, an electron
beam was defined as a more realistic initial source to produce photoneutrons. The
initial source is described in the next section.
In the simulation, the geometry of the X-ray source was remained the same because
the focus of this work was the neutron source. For the X-ray beam production, as usual,the 9-MeV electron beam will irradiate a cupper plate to generate bremsstrahlung
photons. The threshold energies in the 63,65Cu(,n) reactions are above 9 MeV. The peakenergy of the calculated energy distribution of X-ray beam is 100−200 keV.

3. Simulation to design the HUNS-2 thermal neutron source
The beam profile of thermal neutron source for a trial TMRA was simulated using
the PHITS code, where the geometry was based on the original TMRA but somewhat
different configuration. The particle transport calculation was divided into two steps byusing a “DUMP option” in order to reduce the computation time. At the first step,
kinematical information of neutron particles was output to a dump file after apencil-like electron beam of 1 cm in diameter irradiated the neutron production target
at 34 MeV. The neutron production target was only defined in an air space instead of
TMRA. A rectangular tally for the dump covered the neutron production target and had
almost same size as the target. At the second step, the dumped neutron data was used
for each trial TMRA as the initial source because the first step needed a longcomputation time. The evaluated nuclear data of JENDL-4.0 [8] was utilized for thesimulation with respect to the neutron-induced nuclear reactions. The neutrons werecounted by a tally at the distance of L = 1 m from the moderator.
An improved TMRA which enhances the intensity of thermal neutron beam is shownin Fig. 5. The points are (1) positions of neutron production target and polyethylene
moderator, (2) a size of polyethylene moderator, and (3) an additional graphite reflector
with an entrance. For the improved and original TMRA, the calculated energy
distributions of neutron beam are shown in Fig. 4 by the blue and red lines, respectively.
The energy distribution for the improved TMRA is similar to the one for the original
TMRA while the integrated intensity of neutron beam in the energy range of 0.01−0.1
eV is increased up to 1.6 times. In the both calculations, here, the density of graphitereflectors (= c) was 2.3 g/cm3 (theoretical density) although the graphite reflectors with- 167 -

JAEA-Conf 2020-001
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c = 1.6 g/cm3 was used for the actual original TMRA.

4. Conclusion
In order to increase the intensity of thermal neutron beam in HUNS-2, theconfiguration of TMRA for the thermal neutron source was investigated using the
PHITS code. The improved TMRA was obtained through the simulation of the energy
distribution of thermal neutron beam. The TMRA of thermal neutron source will be
upgraded based on the results of this work.

Neutron production target
Moderator(polyethylene)



 




��
 �� 
��

Shieldn


Reflector(Graphite�

Neutron beam


n


� 
� 

�� 



Electron
Shield

beam
 Pb

Shield

Water

SUS


TungstennElectron
beam


Fast

Lead
neutrong
Bremsstrahlung (,n) reaction


Figure 1: Original TMRA of the thermal neutron source in the HUNS facility.
L=7m
2m
0.5 m
Reflector

TargetShieldShieldShieldShieldElectron beam

Moderator

3.5 m
Flight tube

0.7 m

Collimator
ShieldCollimator

ShieldSlit

Neutronbeam

Neutronbeam
Concrete wall

Filter

Detector

Figure 2: Experimental setup in the measurement of the neutron beam intensity.

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JAEA-Conf 2020-001
2020-001

Stainless
Lead
Aluminum
CopperGraphite
Water
Tungsten
Polyethylene
Boric acid resin

Electron beam

Neutron beam

Figure 3: Geometry of the composite source system of neutron and X-ray sources.
2

10

Original
Improved

0

Intensity [a.u.]

10
10

-2

10

-4

10-6
10-8
-10

10

10

-8

-7

10

10

-6

-5

10

10

-4

-3

10

En [MeV]

10

-2

10

-1

0

10

10

1

Figure 4: Energy distribution of thermal neutron beam.
Moderator(polyethylene)


Neutron production
target

Reflector(Graphite�

ShieldShieldShieldShieldNeutron beam

L = 1 mElectron
Shield

beam
 Pb

Shield

Tally


Figure 5: Improved TMRA of the thermal neutron source.

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2020-001

References[1] Furusaka M, Sato H, Kamiyama T, Ohnuma M, Kiyanagi Y, Activity of Hokkaido
University Neutron Source, HUNS. Phys. Proc. 2014;60:167−174.
[2] Ishida T, Ohnuma M, Seong BS, Furusaka M, Intermediate-angle Neutron
Scattering Instruction for Quantitative and Non-destructive Characterization of
Nanostructure in Steel and Other Alloy, ISIJ Int. 2017;57:1831−1837.
[3] Hara KY, Sato H, Kamiyama T, Shinohara T, Measurement and simulation for acomplementary imaging with the neutron and X-ray beams, EPJ Web Conf. 2017;146:
03032-1−4.
[4] Sato T, Iwamoto Y, Hashimoto S, Ogawa T, Furuta T, Abe S, Kai T, Tsai P, Matsuda
N, Iwase H, Shigyo N, Sihver L, Niita K, Features of Particle and Heavy Ion
Transport code System (PHITS) version 3.02, J. Nucl. Sci. Technol. 2018;55;684−690.
[5] Kamiyama T, Hara KY, Sato H, Taira H, Convertible Source System of Thermal
Neutron and X-ray at Hokkaido University Electron Linac Facility, Nuovo Ciemento
2015;38C:187−192.
[6] Uehara Y, Hara KY, Asako M, Sato H, Kamiyama T, Shinohara T, AESJ 2017 Fall
Meeting Poster, Hokkaido University, Sapporo;2017 Sep11−13.
[7] Kino K, Furusaka M, Hiraga F, Kamiyama T, Kiyanagi Y, Furutaka K, Goko S,
Harada H, Harada M, Kai T, Kimura A, Kin T, Kitatani F, Koizumi M, Maekawa F,
Meigo S, Nakamura S, Ooi M, Ohta M, Oshima M, Toh Y, Igashira M, Katabuchi T,
Mizumoto M, Measurement of energy spectra and spatial distributions of neutron
beams provided by the ANNRI beamline for capture cross-section measurements at
the J-PARC/MLF, Nucl. Instur. Methods. 2011;A626−627:58−66.
[8] Shibata K, Iwamoto O, Nakagawa T, Iwamoto N, Ichihara A, Kunieda S, Chiba S,
Furutaka K, Otuka N, Ohsawa T, Murata T, Matsunobu H, Zukeran A, Kamada S,
Katakura J, JENDL-4.0: a new library for nuclear science and engineering, J. Nucl.
Sci. Technol., 2011;48:1−30.

- 170 -

JAEA-Conf 2020-001

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7QBHX h?2 2tT2`BK2MiH /i BM  7Q`K Q7 `2+iBQM `i2 r2`2 +QKT`2/ rBi? +H+mHiBQMb#v mbBM; i?2 S>Aha +Q/2 bBKmHiBM; i?2 2tT2`BK2MiH +QM/BiBQMX b i?2 `2bmHib- Bi rb
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7Q` KMv v2`b (jĜN)- i?2`2 Bb biBHH M Bbbm2 iQ +QM}`K b i?2 +QKT+i M2mi`QM bQm`+2 T2`7Q`KM+2 mbBM;
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mbBM; i?2 7 GBUT- MV7 "2 `2+iBQM b Bib ++QKTHBb?K2Mi i2bi Q7 i?2 bvbi2K ++2TiBM; i?2 T`QiQM #2K QM
i?2 GB i`;2iX
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TQbbB#H2X- 171 -

JAEA-Conf 2020-001k

1tT2`BK2Mi

hrQ 2tT2`BK2Mib r2`2 +QM/m+i2/ iQ Q#iBM `2+iBQM `i2b #v i?2 K2i?Q/b rBi? i?2 3 >2 HQM; +QmMi2` UBV
M/ BM/BmK 7QBH +iBpiBQM UBBV b i?2 }`bi 2tT2`BK2Mi QM M2mi`QM ~mt Q7 _La@AAX _2bQMb 7Q` TTHvBM;i?2 irQ K2i?Q/b r2`2 b 7QHHQrbc i?2 hP6 K2i?Q/- r?B+? Bb bbmK2/ iQ #2 i?2 KQbi 2{+B2Mi 7Q` i?2
M2mi`QM bT2+i`mK K2bm`2K2Mi- Bb MQi 2z2+iBp2 i i?2 _La@AA b b?QrM BM 6B;m`2 R- #2+mb2 M
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M/ 115 AMUM- γV116m AM `2+iBQMbXUBV 3 >2 HQM; +QmMi2`h?2 +QM};m`iBQM Q7 3 >2 HQM; +QmMi2` Bb b?QrM BM 6B;m`2 kX M GL. k8kkN8 3 >2 +QmMi2` rb mb2/X
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mbBM;  jXe J"[ 252 *7 bQm`+2X h?2 /BbiM+2b #2ir22M i?2 bQm`+2 M/ i?2 /2i2+iQ` r2`2 Ryy- 8y- M/ R8+KX L2ti- M2mi`QM `2+iBQM `i2b r2`2 K2bm`2/ rBi? T`QiQM #2KbX h?2 /BbiM+2 7`QK  GB i`;2i iQi?2 3 >2 HQM; +QmMi2` rb k33Xd +KX h?2 T`QiQM #2K +m``2Mib r2`2 #2ir22M yXyee M/ yXRy ƒX

6B;m`2 R*QM};m`iBQM Q7 _La@AA

6B;m`2 k*QM};m`iBQM Q7 3 >2 HQM; +QmMi2`

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6Qm` BM/BmK 7QBHb UAMORĜAMO9V r2`2 TH+2/ QM  +`/#Q`/ b?22i i  TQbBiBQM Q7 RXe +K 7`QK i?2 i`;2iX
6B;m`2 j b?Qrb i?2 ``M;2K2Mi Q7 i?2 7QBHbX h?2 bBx2 M/ r2B;?i Q7 BM/BmK 7QBHb M/ i?2 /BbiM+2b 7`QK
AMOR `2 b?QrM BM h#H2 RX h?2 T`QiQM #2K rBi? M p2`;2 +m``2Mi Q7 jXN ƒ B``/Bi2/ 7Q` R8 KBMmi2bX
7i2` i?2 B``/BiBQM- i?2 7QBHb r2`2 +QQH2/ M/ i?2 jje@F2o /2+v γ@`vb Q7 i?2 115 AMUM- M� V115m AM `2+iBQM
M/ i?2 RkNj@F2o /2+v γ@`vb Q7 i?2 115 AMUM- γV116m AM `2+iBQM r2`2 K2bm`2/ #v M >S:2 /2i2+iQ`XAMORĜAMOj r2`2 TH+2/ i kXd +K 7`QK i?2 bm`7+2 Q7 i?2 >S:2 /2i2+iQ` r?BH2 AMO9 rb TH+2/ i yXky
+K 7`QK i?2 bm`7+2X h?2 `2HiBp2 2{+B2M+v Q7 i?2 >S:2 /2i2+iQ` rb Ry WXjaBKmHiBQM

_2+iBQM `i2b r2`2 +H+mHi2/ mbBM; S>Aha@kX33X 1L.6@"foAAXy (RR) rb mb2/ 7Q` i?2 7 GBUT- MV`2+iBQM +`Qbb b2+iBQMX C1L.G@9Xy (Rk) rb mb2/ 7Q` +`Qbb b2+iBQMb Q7 M2mi`QM BM/m+2/ `2+iBQMX  T`QiQM#2Kb rBi?  /BK2i2` Q7 RXy +K M/ rBi? M mMB7Q`K /Bbi`B#miBQM rb BM+B/2Mi QM i?2 +2Mi2` Q7  GB
i`;2i U/BK2i2` 8 +K- i?B+FM2bb Rdy ƒKV iQ ;2M2`i2 M2mi`QMbX h?2 MmK#2` Q7 M2mi`QMb ;2M2`i2/ 7`QK
i?2 GB i`;2i rb RXkR × Ry−4 T2` T`QiQMXUBV 3 >2 HQM; +QmMi2`- 172 -

JAEA-Conf 2020-001

h#H2 R

6B;m`2 j

1tT2`BK2MiH b2imT Q7 7Qm`
BM/BmK 7QBHb

AMOR
AMOk
AMOj
AMO9

aT2+B}+iBQMb Q7 BM/BmK 7QBHb

aBx2 (+K×+K×+K)

q2B;?i (;).BbiM+2 7`QK AMOR (+K)

RX3 × RXd × yXky
RXd × RX3 × yXky
RX3 × RXN × yXky
RXd × RXN × yXky

9Xee
9XNN
9XN3
9XNy

y
kX3
8X8
8X9

6B;m`2 9 BHHmbi`i2b i?2 +QKTmiiBQMH ;2QK2i`v mbBM; S>AhaX PMHv i?2 _La@AA i`;2i b?B2H/ M/i?2 3 >2 HQM; +QmMi2` r2`2 bbmK2/ #2+mb2 b+ii2`2/ M2mi`QMb 7`QK i?2 ~QQ`- i?2 rHHb- M/ i?2 `QQ7
r2`2 +QMbB/2`2/ iQ #2 M2;HB;B#H2X L2mi`QM i`+F H2M;i?b BM i?2 3 >2 2z2+iBp2 `2 r2`2 +H+mHi2/ #v 
h@i`+F iHHvX //BiBQMHHv- i?2 +H+mHiBQMb Q7 M2mi`QMb 2KBii2/ 7`QK  252 *7 bQm`+2 r2`2 T2`7Q`K2/X
h?2 2M2`;v /Bbi`B#miBQM /2}M2/ #v CAa w 98kR (Rj) rb TTHB2/ b M BMBiBH bQm`+2 BM i?2 bBKmHiBQMXUBBV AM/BmK 7QBH +iBpiBQM
6B;m`2 8 /2TB+ib i?2 ;2QK2i`v Q7 i?2 bBKmHiBQMX AM/BmK 7QBHb r2`2 TH+2/ QM i?2 bK2 TQbBiBQMb b i?2
2tT2`BK2MiX PMHv i?2 _La@AA i`;2i b?B2H/ M/ i?2 BM/BmK 7QBHb r2`2 bbmK2/X L2mi`QM i`+F H2M;i?b
BM i?2 7QBHb r2`2 Q#iBM2/ #v h@i`+F iHHB2bX_2+iBQM `i2b r2`2 +H+mHi2/ mbBM; i?2 h@i`+F `2bmHibX AM i?2 +H+mHiBQMb- C1L.G@9Xy rb mb2/ 7Q`>2UM- TV +`Qbb b2+iBQM Q7 i?2 3 >2 /2i2+iQ`- M/ C1L.G /QbBK2i`v }H2 NN (R9) rb mb2/ 7Q` 115 AMUM- M� V
M/ 115 AMUM- γV `2+iBQM +`Qbb b2+iBQMb Q7 i?2 BM/BmK 7QBH +iBpiBQMX

3

Li target
10
10

10
10

55

202

x [cm]
x [cm]y [cm]

55

201

00

00

200

-5
<5

108

203

ne

103

<5
-5

109

108

106

104

105

101
102 200

900

104

108

106
103

900

<10
-10
<10
-10

<5
-5

00
55
x [cm]
x [cm]

Air
Al
In#1
In#2
Polyethylene

6B;m`2 99:2QK2i`v Q7 +H+mHiBQM 7Q` i?2 3 >2
HQM; +QmMi2`

6B;m`2 8

-10
<10
<4
<2
0
2
4
-4
-2
z0
[cm] 2 4z [cm]
109

109

10
10

Cu
In#3
vacuumH2O
In#4:2QK2i`v Q7 +H+mHiBQM 7Q` i?2 BM/BmK
7QBH +iBpiBQM

_2bmHib

UBV 3 >2 HQM; +QmMi2`
h#H2 k b?Qrb i?2 K2bm`2/ M/ +H+mHi2/ `2+iBQM `i2b rBi? i?2 252 *7 bQm`+2X 1``Q`b BM i?2 i#H2b
+Q``2bTQM/ iQ biiBbiB+H 2``Q`bX h?2 +H+mHiBQM iQ 2tT2`BK2Mi `iBQ U*f1V mbBM; i?2 252 *7 bQm`+2 rb
rBi?BM i?2 `M;2 Q7 RXde iQ RXNyX h?2 Q`B;BM Q7 i?2 /Bb+`2TM+B2b Bb mM/2` BMp2biB;iBQMX SQbbB#H2 `2bQMb
7Q` i?2 /Bb+`2TM+B2b `2 mM+2`iBMiB2b BM i?2 ;b T`2bbm`2 M/ b2MbBiBp2 `2 Q7 i?2 3 >2 HQM; +QmMi2`X 6Q`
2tKTH2- r?2M i?2 T`2bbm`2 Bb R iK- i?2 `2+iBQM `i2 i Ryy +K Bb 8RXj b−1 - r?B+? Bb 9R W bKHH2` i?M
i?i 7Q` Ry iKX h#H2 j b?Qrb i?2 K2bm`2/ M/ +H+mHi2/ `2+iBQM `i2b rBi? i?2 T`QiQM #2KX 6Q`
i?2 K2bm`2K2Mib i i?2 T`QiQM #2K +m``2Mi Q7 yXyeeĜyXRy ƒ- h?2 *f1 rb 3X93Ĝ3XddX h?2`27Q`2- i?2
S>Aha +H+mHiBQM BM h#H2b k M/ j Qp2`2biBKi2/ i?2 2tT2`BK2MiH pHm2bX
UBBV AM/BmK 7QBH +iBpiBQM

- 173 -

JAEA-Conf 2020-001

h#H2 k*QKT`BbQM #2ir22M i?2 K2bm`2/ M/+H+mHi2/ `2+iBQM `i2b rBi? 252 *7
bQm`+2h#H2 j*QKT`BbQM #2ir22M i?2 K2bm`2/ M/+H+mHi2/ `2+iBQM `i2b rBi? i?2 T`QiQM#2Kb.BbiM+2 (+K)

1tTX (fb)

S>Aha (fb)*f1*m``2Mi (ƒ)

1tTX (fb)

S>Aha (fb)

*f1

Ryy
8y
R8

dRXy ± yXd
key ± R
kjRd ± 8

Rk8
9N8
9ke9

RXde
RXNy
RX39

yXyee
yXy38
yXRy

kRd ± R
kd3 ± kjkN ± j

R39k
kjdN
k339

3X93
3X8e
3Xdd

h#H2 9*QKT`BbQM #2ir22M i?2 K2bm`2/ M/+H+mHi2/ `2+iBQM `i2b 7Q` i?2 115 AMUMγV116m AM `2+iBQM i jXN ƒ

AMOR
AMOk
AMOj
AMO9

1tTX (fb)

S>Aha (fb)*f1

8X8 × Ry
8Xj × Ry−17
9XN × Ry−17
9XR × Ry−17

kXj × RyRX3 × Ry−16
RX3 × Ry−16
RX3 × Ry−16

9Xk
jX9
jXd
9X9

−17

−16

h#H2 8*QKT`BbQM #2ir22M i?2 K2bm`2/ M/+H+mHi2/ `2+iBQM `i2b 7Q` i?2 115 AMUMM� V115m AM `2+iBQM i jXN ƒ
1tTX (fb)

AMOR
AMOk
AMOj

U8X8 ± yXRV × Ry
UjX9 ± yXRV × Ry−20
URXe ± yXkV × Ry−21−20

S>Aha (fb)*f1

9X9 × Ry
RXN × Ry−20
RX3 × Ry−21

3Xy
yX8e
RXR

−19

h#H2b 9 M/ 8 b?Qr i?2 `2+iBQM `i2b Q7 i?2 115 AMUM- γV116m AM M/ 115 AMUM- M� V115m AM `2+iBQMb`2bT2+iBp2HvX 1``Q`b BM i?2 i#H2 +Q``2bTQM/ iQ biiBbiB+H 2``Q`bX h?2 *f1 7Q` i?2 115 AMUM- γV116m AM`2+iBQM rb jX9Ĝ9X9 r?BH2 i?i 7Q` i?2 115 AMUM- M� V115m AM `2+iBQM rb yX8eĜ3XyX PM2 Q7 i?2 TQbbB#H2+mb2b 7Q` i?2 /Bz2`2M+2b Q7 *f1 #2ir22M i?2 irQ `2+iBQMb Bb i?i i?2 T`QiQM #2K TQbBiBQM #2ir22M i?2
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i2M/ iQ #2 2KBii2/ BM i?2 7Q`r`/ /B`2+iBQMX h?2`27Q`2- `2+iBQM `i2b 7Q` /Bz2`2Mi #2K TQbBiBQMb r2`2
+H+mHi2/ #v S>AhaXh#H2b e M/ d b?Qr i?2 `2+iBQM `i2b Q7 i?2 115 AMUM- γV116m AM M/ 115 AMUM- M� V115m AM `2+iBQMb`2bT2+iBp2Hv- r?2M TQbBiBQM Q7 T`QiQM #2K Bb KQp2/ RXk +K iQr`/ AMOkX h?2b2 `2bmHib b?Qr i?i i?2`2+iBQM `i2b 7Q` i?2 115 AMUM- M� V115m AM `2+iBQM H`;2Hv /2T2M/b QM i?2 T`QiQM #2K TQbBiBQMXh#H2 e_2+iBQM `i2b 7Q` i?2 115 AMUM- γV116m AM`2+iBQM i jXN ƒ r?2M i?2 TQbBiBQM Q7
i?2 T`QiQM #2K Bb KQp2/ RXk +K iQr`/
AMOk

AMOR
AMOk
AMOj
AMO9

8

1tTX (fb)

S>Aha (fb)*f1

8X8 × Ry
8Xj × Ry−17
9XN × Ry−17
9XR × Ry−17

kXk × RykXy × Ry−16
RX3 × Ry−16
RXe × Ry−16

9Xy
jX3
jXd
jXN

−17

−16

h#H2 d_2+iBQM `i2b 7Q` i?2 115 AMUM- M� V115m AM`2+iBQM i jXN ƒ r?2M i?2 TQbBiBQM Q7
i?2 T`QiQM #2K Bb KQp2/ RXk +K iQr`/AMOk
1tTX (fb)

AMOR
AMOk
AMOj

U8X8 ± yXRV × Ry
UjX9 ± yXRV × Ry−20
URXe ± yXkV × Ry−21−20

S>Aha (fb)*f1

kXj × RyRXk × Ry−19
9Xk × Ry−21

9Xk
jX8
kXe

−19

.Bb+mbbBQM

h?2 S>Aha +H+mHiBQM Qp2`2biBKi2/ i?2 `2+iBQM `i2b 7Q` #Qi? 2tT2`BK2MibX "b2/ QM i?2 `2bmHibr2 +QMbB/2`2/ i?`22 TQbbB#H2 +mb2b Q7 i?2 /Bz2`2M+2c T`QiQM #2K 2M2`;v UAV- T`QiQM #2K T`Q}H2 UAAVM/ i?2 7 GBUT- MV +`Qbb b2+iBQM UAAAVXUAV S`QiQM #2K 2M2`;v- 174 -

JAEA-Conf 2020-001

Cross Section [barns]

0.8

ENDF/B-VII.0
L.Damone(2018)
K.K.Sekharan(1976)
J.H.Gibbons(1959)
H.W.Newson(1957)
R.Taschek(1948)

0.7
0.6
0.5
0.4
0.3
0.2
0.1
0

6B;m`2 d

Differential Cross Section [barns/sr]

Neutron yields [/proton]

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q?2M i?2 T`QiQM 2M2`;v Bb +?M;2/ 7`QK kX9N iQ
kXR J2o- M2mi`QM vB2H/b /2+`2b2 iQ #Qmi Rf9X
×10−3
0.2
PM2 Q7 i?2 TQbbB#H2 +mb2b Q7 HQr M2mi`QM vB2H/
0.18
0.16
BM i?2 2tT2`BK2Mi Bb i?i BM+B/2Mi T`QiQM 2M2`;v
0.14
0.12
Bb HQr2` i?M 2tT2+i2/X Ai Bb `2[mB`2/ iQ K2bm`2
0.1
i?2 T`QiQM #2K 2M2`;v /B`2+iHvX
0.08
0.06
UAAV S`QiQM #2K T`Q}H2
0.04
aBM+2 i?2`2 rb  TQbbB#BHBiv i?i i?2 #2K
0.02
2
2.2
2.4
2.6
2.8
3
rb b?B7i2/ 7`QK i?2 +2Mi2` Q7 i?2 GB i`;2i BM
Proton Energy [MeV]i?2 +iBpiBQM 2tT2`BK2Mib- Bi Bb T`2bmK2/ i?i
6B;m`2 e L2mi`QM vB2H/b 7`QK GB i`;2i 7Q` BM+B/2Mi
HH BM+B/2Mi T`QiQMb /B/ MQi ?Bi i?2 GB i`;2iX AM
T`QiQM 2M2`;v
Q`/2` iQ K2bm`2 T`QiQM #2K T`Q}H2- M //B@
iBQMH 2tT2`BK2Mi Bb THMM2/X
UAAAV 7 GBUT- MV +`Qbb b2+iBQM

1.6

1.8

2
2.2 2.4 2.6
Proton Energy[MeV]

2.8

3*`Qbb b2+iBQM Q7 i?2 7 GBUT- MV7 "2 `2+@iBQM b  7mM+iBQM Q7 i?2 T`QiQM 2M2`;v

6B;m`2 3

0.4
ENDF/B-VII.0

0.35
0.3

H.W.Lefevre (1969)

0.25

F.Gabbard (1959)

0.2
0.15
0.1
0.05
0

1.6

1.8

2
2.2 2.4 2.6
Proton Energy[MeV]2.8

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7Q` i?2 +iBpiBQM i2+?MB[m2 BM T`iB+mH` +QmH/ #2 z2+i2/ #v i?2 #2K TQbBiBQM #2+mb2 i?2 BM/BmK 7QBHrb TH+2/ i  p2`v M2` TQbBiBQM M/  +QMbB/2`#H2 H`;2 bQHB/ M;H2 rBi? `2bT2+i iQ i?2 i`;2i +2Mi2`X>Qr2p2`- i?2 T`2b2Mi K2bm`2K2Mi bvbi2K ?b  HBKBi iQ B/2MiB7v i?2 +Q``2+i TQbBiBQM Q7 i?2 T`QiQM #2KX
h?mb r2 BMp2biB;i2/ Bib 2z2+i p2`v i2MiiBp2Hv b /2b+`B#2/ BM a2+iBQM 9 UBBVX- 175 -

JAEA-Conf 2020-001

e

amKK`v

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KQ/2`iQ` M/ i?2 7QBH +iBpiBQM i2+?MB[m2 7Q` i?2 115 AMUM- M V115m AM M/ 115 AMUM- γV116m AM `2+iBQMb
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kyykcRj99,RĜRjjX

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JAEA-Conf 2020-001

33 Development of a neutron beam monitor for nuclear datameasurement using spallation neutron source
Hideto NAKANO1∗, T. KATABUCHI1, G. ROVIRA1, Y. KODAMA1, K. TERADA1
A. KIMURA2, S. NAKAMURA2, S. ENDO2
1) Laboratory for Advanced Nuclear Energy, Tokyo Institute of Technology
2) Japan Atomic Energy Agency
e-mail: nakano.h.ag@m.titech.ac.jp

In neutron capture cross section measurement, the number of the incident neutrons is necessary toderive the neutron capture cross section. However, in measurement with ANNRI, neutron monitoring
detection has not been employed. To avoid possible failure of the proton pulse counting method
and make measurement with ANNRI more robust, an additional neutron beam monitor is underdevelopment. A plastic scintillator and 6 Li are adopted for a detector. A test detector system was built
to study the feasibility of the method. Neutron detection is conﬁrmed in experiment results.

I.

Introduction

Highly precise neutron nuclear data are required in nuclear transmutation research of long-lived
minor actinides (MA) in nuclear waste. It has been difﬁcult to measure neutron-induced reaction
cross sections of MAs due to large background of the decay gamma-rays from radioactive samples. In
recent years, with the advent of spallation neutron sources, the qualities of cross section measurementsof MAs were signiﬁcantly improved. The Japanese Spallation Neutron Source (JSNS) in the Japan
Proton Accelerator Research Complex (J-PARC) was started in operation in 2008. In order to utilize ahigh-intensity pulsed neutron beam from JSNS for nuclear data measurement, the Accurate Neutron
Nucleus Reaction Measurement Instrument (ANNRI) was built and has been used for the past ten
years [1] .
In neutron capture cross section measurement, the number of the incident neutrons is necessary toderive the neutron capture cross section. To normalize the detected gamma-ray yield to the number ofthe incident neutrons, the neutron count is usually monitored by detecting the incident neutrons with aneutron detector. However, in measurement with ANNRI, neutron monitoring detection has not been
employed and, instead, the number of proton beam pulses injected into the spallation target has beenused based on the assumption that the number of proton beam pulses is proportional to the numberof incident neutrons. This assumption is mostly plausible but could fail when the conditions of the
proton accelerator or the neutron source change. To avoid possible failure of the proton pulse counting
method and make measurement with ANNRI more robust, an additional neutron beam monitor isunder development.

II.

Detector Design

To develop a neutron beam monitor for ANNRI, there are two issues to overcome. First, very highintensity neutron beam from JSNS requires a fast detector system that can process signals at a highcounting rate. Second, gamma-ﬂash, an intense gamma-ray burst produced when the proton beampulse bombards the spallation target, can paralyze a detector. Thus, gamma-ray sensitivity of the neutron monitor should be low. In order to fulﬁll the requirements, a thin sheet-type plastic scintillator- 177 -

JAEA-Conf 2020-001

combined with a thin 6 Li layer on a Mylar ﬁlm is adopted for the present neutron monitor. The incident
neutrons react with 6 Li and the 6 Li(n,t)4 He reaction occurs. The emitted particles, tritons and alphas,
are detected with the plastic scintillator. The short ranges of tritons and alphas allow for using a thinplastic scintillator ﬁlm, and the thin detector leads to low gamma-ray sensibility. Another requirement
for fast detection is achieved by the fast response property of plastic scintillator. Simulation studies using Monte Carlo simulation code PHITS [2] was performed to optimize the detector design, especiallythickness of the 6 LiF layer. In Fig. 1, the overall geometric design is shown.

Figure 1: Detector geometric design.

III.

Experiments

A test detector system was built to study the feasibility. LiF was deposited on a Kapton ﬁlm
by a vacuum deposition method. The LiF layer was thin enough for tritons and alpha particles topenetrate and reach the plastic scintillator. Li was not isotopically enriched. The isotopic ratio of 6 Li
was the natural abundance of 7.6%. The photomultiplier tube is HAMAMATSU R1306-22ASSY. Test
experiments were carried out at ANNRI. The detector was placed at a ﬂight length of 29 m. Twomeasurements with and without LiF foil were conducted. In Fig. 2, time-of-ﬂight (TOF) spectra areshown. TOF measurement was started by a signal from the J-PARC accelerator and stopped by PMT
signal. In Fig. 2, a strong peak is observed in the TOF region from 1 to 2.5 µs. This peak is created by
gamma-ﬂash from the spallation target. No events are observed in the TOF region from 3 to 7 µs. Thiswas caused by dead time of the system. The intense gamma-ray burst was detected in each TOF cycle
and paralyzed the system for the system dead time. In Figs. 3 and 4, pulse-height (PH) spectra areshown. One is for full TOF range and the other is for TOF larger than 3 µs. A peak between 300 and
1200 ch is observed only in full TOF range and disappears in larger than 3 µs. The peak is observedin both with and without LiF. This suggests that the peak is created by the gamma-ﬂash events. The
6 Li(n,t)4 He events are a broad continuum component observed only in Fig. 3.

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JAEA-Conf 2020-001

Figure 2: TOF spectra with and without LiF.

Figure 3: PH spectra with LiF of full TOF range and larger
than 3 µs.

Figure 4: PH spectra without LiF of full TOF range andlarger than 3 µs.

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JAEA-Conf 2020-001

IV. Summary
A new neutron detector system was designed and built. Signal from the 6 Li(n, t)4 He reaction was
clearly observed. In the future plan, an enriched 6 Li will be used and the system will be optimized. A
waveform acquisition system is also planed to use to reduce the dead time. The new system will betested at J-PARC and its performance will be evaluated.

References[1] Kino, K., et al. Nucl. Instrum. Methods Phys. Res. A 626, pp. 58-66, 2011.
[2] Tatsuhiko S, et al. Journal of Nuclear Science and Technology, Vol. 55, No. 6, pp. 684-690, 2018.

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Model Analysis of Isotope-Production Cross Sections
for Proton- and Deuteron-Induced Reactions on 93Zr
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JAEA-Conf 2020-001&KLQD

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JAEA-Conf 2020-001
2020-001

35

Systematical calculation of probabilities of beta-delayed neutron
emission and fission in the entire region of nuclear chart
Hiroyuki KOURA
Advanced Science Research Center, Japan Atomic Energy Agency
Shirakata 2-4, Tokai-mura, Ibaraki 319-1195, Japan
Fumitaka Endo
Department of Physics, Tohoku University, Aoba, Sendai, Miyagi 980-8578, Japane-mail: koura.hiroyuki@jaea.go.jp

In the very neutron-rich nuclei, the beta-decay occurs with some accompanying processes as
neutron emissions. The beta-delayed fission also takes place in the superheavy nuclear mass
region. The beta-decayed neutron process is an important phenomenon for manipulating nuclear
reactor, and the beta-delayed fission critically affects the termination of the r-process
nucleosynthesis in stars. We performed a systematical calculation of probabilities of beta-delayed
neutron and beta-delayed fission in the entire region of nuclear chart with the improved grosstheory of beta-decay and the KTUY mass model.

1. Introduction
The − decay occurs with a weak interaction in the neutron-rich nuclear mass region andnucleus releases electron, gamma-ray and anti-neutrino. Many neutron-rich nuclei can emit
neutrons through  decay, called delayed neutrons. Delayed neutrons play an important role inallowing nuclear reactors to be safely controlled, due to the delay that causes between
neutron-induced fission events and their eventual emission. Moreover, in the superheavy nuclear
mass region, some of nuclei occur fission process due to the repulsive Coulomb force between
protons.

Regarding the beta decay, beta-delayed fission can occur if the fission barrier ofdaughter nucleus is lower than the beta-decay Q-value of parent nucleus. These nuclei are
expected to be very neutron-rich and superheavy nuclear mass region.
In order to estimate  -decay rate, delayed neutron probability, and delayed fission probability
theoretically, calculation of the nuclear matrix elements of the  decay is required. The grosstheory is a type of macroscopic model to describe various -decay properties. It is based on sum
rules of the -decay strength function, and treats the transitions to all the final nuclear levels in a

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statistical manner. The gross theory has been successful in describing  decay for the entire
nuclear mass range [1–6]. The results of the gross theory provide a guide to experiments on 
decay, especially, for newly measured nuclear data of very neutron-rich nuclei, and for purely
theoretical nuclear data extremely far from known nuclei. Due to its statistical treatment, thegross theory only describes macroscopic features. Recently, we have introduced a treatment ofspin and parity to the gross theory, and have improved decay rate in especially the forbidden
transition [7-9].
In this report, we perform a systematical calculation of probabilities of beta-delayed neutron
and beta-delayed fission in the entire region of nuclear chart with the improved gross theory ofbeta-decay and the KTUY nuclear mass model [10] and fission barrier calculation [11].
2. Beta decay
The decay constant of the -decay can be divided by the types of -decay operators, , and the
decay constant is obtained as the sum of partial decay constants, λ. If we take into account theallowed and first forbidden transitions, the total -decay rate is expressed as𝜆𝜆� � 𝜆𝜆F � 𝜆𝜆GT � 𝜆𝜆1st � �,
where the right-hand represents the
decay rates of the Fermi transition,
Gamow-Tellertransition,

and

the

first-forbidden transition. The first two
terms are the allowed transitions.
Under

the

usual

approximation,

each decay rate can be written with thenuclear matrix elements, |M(E)|, which
can be calculated in the framework of thenuclear physics, and the integrated Fermi
function, f , which represents a distortionof wave functions due to the Coulomb
force. The actual expressions for theallowed transitions are

Figure1:

Illustration

of

beta

decay

and

beta-delayed neutron emission. (Left): Level
scheme of beta decay. (Right): Strength function ofbeta decay. The (black) histograms indicate theintensity of beta decay, which is equal to squatted
nuclear matrix element, |M(E )|2 . The purple
line is the strength function averaged with eachintensity.

Pn:

Beta-delayed neutron emission

probability. This decay is available with QSn>0.

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JAEA-Conf 2020-001
2020-001𝜆𝜆� ���� � ���� ℏ�

𝜆𝜆�� �

�|𝑔𝑔� |� ��� |𝑀𝑀� �𝐸𝐸�|� 𝑓𝑓��𝐸𝐸�𝑑𝑑𝑑𝑑,

��� � ���� ℏ�

��

|𝑔𝑔� |� ∙ 3 � ��� |𝑀𝑀�� �𝐸𝐸�|� 𝑓𝑓��𝐸𝐸�𝑑𝑑𝑑𝑑,
�Here, the coefficients are composed of the coupling constant of the weak interaction and thephysical constants as the mass of electron, me, the light velocity, c, the Planck constant per 2 𝜋𝜋, ℏ.
The coupling constant of the weak interaction has two types: the vector type as 𝑔𝑔� and the axial

vector type as 𝑔𝑔� , respectively. The integral is performed from −Q to 0 and Q is the total(maximum) decay energy from the ground-state of parent to daughter nuclei, or -decay Q-value.

The emission of delayed neutrons is a phenomenon that accompanies  decay. Figure 1 shows aschematic illustration of the  decay and delayed neutron emission processes. In  decay, a nucleus
decays from the parent ground state to a particular daughter state. If the neutron separationenergy for the daughter nucleus, Sn, is smaller than Q, the nucleus can emit a neutron with anenergy from 0 to Q − Sn, measured from the ground state of the parent nucleus. This process
occurs in the transmission process of the quantum mechanics. The situation of beta-delayed fissionis the same as the delayed neutron (Fig. 2).
However, the fission process occurs in thetunneling penetration, which is rather different
from the neutron emission. The decay widths ofdelayed neutron emission and delayed fission
probabilities are calculated by the following
expressions:Γ� �𝐸𝐸� �Γ� �𝐸𝐸� �

Figure 2: Illustration of beta-delayed neutron
emission and beta-delayed fission.

����1 1 2𝑀𝑀𝑀𝑀�𝑔𝑔�𝜌𝜌∗ �𝐸𝐸 � �� � 𝜀𝜀�𝜀𝜀 𝑑𝑑𝑑𝑑,
2𝜋𝜋 𝜌𝜌�𝐸𝐸� ℏ��������� ∗ �𝐸𝐸� �fiss � 𝜀𝜀�
𝜌𝜌
1 1
�
𝑑𝑑𝑑𝑑,
2𝜋𝜋𝜋𝜋
2𝜋𝜋 𝜌𝜌�𝐸𝐸� ��
1 � exp ���ℏ𝜔𝜔where, M is the mass of the nucleus, R is the radius, g is the weight, (E) is the level density ofthe daughter nucleus,  is the width of the fission barrier, and Bfiss is the fission barrier height of

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JAEA-Conf 2020-001
2020-001

the daughter nucleus.
With the use of these decay widths and that of gamma transition, Γ� , the probabilities ofbeta-delayed neutron and beta-delayed fission are calculated as
𝑃𝑃� �

𝐶𝐶 �
Γ�
� 𝑆𝑆� �𝑑𝑑� 𝑓𝑓��𝑑𝑑�
𝑑𝑑𝑑𝑑,
𝜆𝜆 ���Γ� �Γ� � Γ�where, k is indication of the decay

100‘f’ for delayed fission). The coefficient80

C is a constant related to the

decay constant of the beta decay, 𝜆𝜆,

60

80
40

probabilities

increasewithincreasing of neutron-rich can be
seen.

This trend continues

even in the superheavy nuclear
mass region.
Thebeta-delayed

fissionprobabilities are also shown in
Fig. 4.

In the landscape offission-barrier height with the

200

220

240

260

40

Proton number Z

20

P≤2n (%)
20

40

60

80

0
100

120
140
160
Neutron number N

180

200

220

60

240

260

100
80

40

60
40

20

20

P≤3n (%)
0

20

40

60

80

P≤3n (%)

the

180

60

0

80

Generally,

120
140
160
Neutron number N

100

20

nuclearappears.

0
100

80

100

along N=82, 126, 184, and 228

80

40

the shell closures from the KTUY
discontinuity of the probabilities

60

60

Due tomodel,

40

80

calculated beta-delayed neutron

mass

20

100

Figure 3 shows examples ofemission probabilities.

20

P≤1n (%)

P≤2n (%)

3. Results

40

0Proton number Z

improved gross theory [7-9].

60

20

and the beta-decay strength function,𝑆𝑆� �𝐸𝐸� , are calculated from the

100
P≤1n (%)

coefficients outside the integrals. The

Proton number Z

process (‘n’ for delayed neutron, and

0
100

120
140
160
Neutron number N

180

200

220

240

260

Figure 3: Examples of beta-delayed neutron emission
probabilities for one-, two- and three- neutron emissions.
P≤1n: Probability of one-neutron emission excluding two or
more neutrons. P≤2n: Probability of two-neutron emission
excluding three or more neutrons. P≤3n: three-neutron
emission excluding four or more neutrons. Black squareindicates the long-lived nuclei and gray one shows
estimated proton and neutron-drip lines.

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JAEA-Conf 2020-001
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KTUY mass model in super- andextremely

super-heavy

mass

region, the “conventional” islandof stability is found near nuclei
with Z=114 and N=184, while
there is “peninsula” along N=228,
which appeared in Ref. [11]. The
N=228 peninsula is caused bysingle-particle shell gap of N=228
as described in Ref. [12]. These

Figure 4: Beta-delayed fission probabilities.

properties affect the beta-delayed

points indicate experimentally-identified nuclides until

fission probabilities. In Figure 4,

2018, from the chart of the JAEA nuclides 2018 [13].

there

are

lower

The blackprobabilities

around Z=126 and N=228. This trend comes from higher fission barrier of the KTUY modelaround the same region. Below the region of N=228, beta-delayed fission process is limited only
along beta-stable mass region, which locates nearly Z=114 and N=184, and Z=126 and N=228.
Above the region of N=228, beta-delayed fissioning region dominates. The beta-delayed fission
affects the abundance of the rapid neutron capture process (r-process) in stars. The fission acts
as termination and probably recycling of synthesis of nuclides.
Acknowledgment
This work was financially supported by the Japan Society for the Promotion of Science
KAKENHI (Grants No. 18K03631 and No. 17H06090) as a program ‘Effects of fission to isotopic
abundances on the r-process synthesis in star’ and ‘Study of the origin of heavy elements using an
innovative mass spectrograph’.
References[1] K. Takahashi, M. Yamada, Prog. Theor. Phys. 41, 1470 (1969)[2] S. Koyama, K. Takahashi, M. Yamada, Prog. Theor. Phys. 44, 663 (1970)[3] K. Takahashi, M. Yamada, T. Kondoh, Atom. Dat. Nucl. Dat. Tab. 12, 101 (1973)[4] T. Kondoh, T. Tachibana, M. Yamada, Prog. Theor. Phys. 74, 708 (1985)[5] T. Tachibana, M. Yamada, Y. Yoshida, Prog. Theor. Phys. 84, 641 (1990)[6] T. Tachibana, M. Yamada, Proc. Inc. Conf. on Exotic Nuclei and Atomic Masses, ENAM95,

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JAEA-Conf 2020-001
2020-001(Editions Frontueres, Gif-sur-Yvette, 2015), p. 763, and references therein.
https://wwwndc.jaea.go.jp/nucldata/beta-decay-properties.pdf (accessed 2020/03/24).
[7] H. Koura, T. Yoshida, T. Tachibana, S. Chiba, EPJ Web of Conferences 146, 12003 (2017)[8] H. Koura, S. Chiba, Phys. Rev. C 97, 06430 (2017)[9] F. Endo, H. Koura, Phys. Rev. C 99, 034303 (2019)[10] H. Koura, T. Tachibana, M. Uno, M. Yamada, Prog. Theor. Phys. 113, 305 (2005)
http://wwwndc.jaea.go.jp/nucldata/mass/KTUY04_E.html (accessed 2020/03/24).
[11] H. Koura, Prog. Theor. Exp. Phys. 2014, 113D02 (2014)[12] H. Koura, S. Chiba, J. Phys. Soc. Jpn. 82, 014201 (2013)[13] H. Koura, J. Katakura, T. Tachibana and F. Minato, JAEA-Conf 2019-001, 151-156 (2019)

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JAEA-Conf 2020-001

36 Development of multi-group neutron activation cross-section
library from JENDL/AD-2017
Chikara KONNO
Japan Atomic Energy Agency
Email: konno.chikara@jaea.go.jp
JENDL Activation Cross Section File for Nuclear Decommissioning 2017 (JENDL/AD-2017) was released
in 2018. Then a multi-group neutron activation cross-section library (MAXS/AD-2017) with the sameformat as MAXS-2015 by Dr. Okumura has been developed from JENDL/AD-2017 by using PREPRO2018 for activation calculations in nuclear facility decommissioning. MAXS/AD-2017 will be converted to
ORIGEN libraries and be tested with the JPDR decommissioning data. In future MAXS/AD-2017 will be
released.
1. Introduction
JENDL Activation Cross Section File for Nuclear Decommissioning 2017 (JENDL/AD-2017) [1] wasreleased in 2018. This file includes the data of neutron-induced nuclear reactions for 311 nuclides from 10-5
eV to 20 MeV. Dr. Okumura et al. developed a multi-group neutron activation cross-section library(MAXS2015) based on the nuclear data libraries JENDL-4.0 [2] and JEFF-3.0/A [3] for activation
calculations in nuclear facility decommissioning [4]. A multi-group neutron activation cross-section library(MAXS/AD-2017) with the same format as MAXS-2015 has been developed from JENDL/AD-2017 in orderto make it possible to use the new JENDL file for activation calculations in nuclear facility decommissioning.
2. How to make MAXS/AD-2017
JENDL/AD-2017 includes total production cross sections (MF3) of radioactive and stable nuclides,
branching ratios (MF9) and partial production cross sections (MF10) for the ground and isomer states of
nuclides. JENDL/AD-2017 has the following four versions;l MF3, MF9 and MF10 at 0 K,
l MF3, MF9 and MF10 at 293.6 K,l MF3 and MF10 at 0 K (for NJOY [5] processing),
l MF3 and MF10 at 293.6 K (for NJOY processing).
MAXS-2015 was produced with the NJOY2012 [5] code. However it was found that GENDF files
produced with the groupr module in NJOY2012 did not include production cross sections to isomer states.
Then the PREPRO 2018 [6] code was adopted for producing a group-wise file of JENDL/AD-2017 (MF3,
MF9 and MF10 at 0 K). The following modules in PREPRO 2018 were used; ENDF2C, LINEAR, RECENT,SIGMA1, ACTIVATE, FIXUP, DICTIN, GROUPIE. The calculation conditions are as follows;l Temperature : 300 K,- 193 -

JAEA-Conf 2020-001

l Group structure : 199 groups (VITAMIN-B6 [7]),l Weighting spectrum : Maxwell + 1/E + Fission,l Infinite dilution cross section.
The produced group-wise file of JENDL/AD-2017 was converted to MAXS/AD-2017 of the MAXS format[4] with a small program. Figure 1 shows the data of 59Co in MAXS/AD-2017 as an example. Figure 2 plots
the capture cross section (red line) of 59Co in MAXS/AD-2017 with the continuous energy one (blue line),
where the red line represents the blue line well.
The following issues were pointed out in this processing.
l No information of decay data (MF8) in the capture reaction of 187W and 193Os à Add
l The MT number of the (n,t) reaction of 6Li is changed from 105 to 107 for ORIGEN-S [8] because
ORIGEN-S cannot treat the (n,t) reaction.
l The MAXS format includes no data for the (n,n’) reactionà MAXS/AD-2017 includes the data for the(n,n’) reaction, though ORIGEN-S cannot treat the (n,n’) reaction.
A similar procedure for a DCHAIN-SP [9] library was also established, and was provided to the PHITS group.
Users can use the DCHAIN-SP library of JENDL/AD-2017 in the latest PHITS (PHITS3.16) [9].
3. Summary
A multi-group neutron activation cross-section library (MAXS/AD-2017) with the MAXS format wasdeveloped from JENDL/AD-2017 for activation calculations in nuclear facility decommissioning. Next
MAXS/AD-2017 will be converted to ORIGEN-S libraries and be tested with the JPDR decommissioning
data [10]. Then MAXS/AD-2017 will be released.
4. References[1] https://wwwndc.jaea.go.jp/ftpnd/jendl/jendl-ad-2017.html (current as of Jan. 2020).
[2] K. Shibata, O. Iwamoto, T. Nakagawa et al., JENDL-4.0: “a new library for nuclear science andengineering”, J. Nucl. Sci. Technol. 48, 1 (2011).
[3] J-Ch. Sublet, A.J. Koning, R.A. Forrest, and J. Kopecky, “The JEFF-3.0/A Neutron Activation File—
New Features, Validation, and Usage”, AIP Conference Proceedings 769, 203 (2005).
[4] K. Okumura, K. Kojima, K. Tanaka, “Development of multi-group neutron activation cross-section
library for decommissioning of nuclear facilities”, Proc. of 2014 Symposium on Nuclear Data, p. 43,
JAEA-Conf 2015-003(2016).
[5] R.E. MacFarlane, D. W. Muir, R. M. Boicourt, A. C. Kahler, “The NJOY Nuclear Data Processing
System, Version 2012”, LA-UR-12-27079, Los Alamos National Laboratory (2012).
[6] https://www-nds.iaea.org/public/endf/prepro2018/ (current as of Jan. 2020).
[7] J.E. White, et al., “Production and Testing of the Revised VITAMIN-B6 Fine-Group and the
BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived From ENDF/B-VI.3
Nuclear Data”, ORNL/TM-6795/R1 (2000).

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JAEA-Conf 2020-001[8] I.C. Gauld, O.W. Hermann, R.M. Westfall, “ORIGEN-S: SCALE SYSTEM MODULE TO
CALCULATE FUEL DEPLETION, ACTINIDE TRANSMUTATION, FISSION PRODUCTBUILDUP AND DECAY, AND ASSOCIATED RADIATION SOURCE TERMS”, ORNL/
TM-2005/39 Version 6 Vol. II, Sect. F7 (2009).
[9] https://phits.jaea.go.jp/News.html#news0030 (current as of Jan. 2020).
[10] N. P. Kocherov (Ed.), “International benchmark calculations of radioactive inventory for fission reactor
decommissioning”, INDC(NDS)-355 (1996).

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Fig. 1 Example of MAXS/AD-2017 (59Co).

JAEA-Conf 2020-001

JAEA-Conf 2020-001

Fig. 2 Capture cross section of 59Co in JENDL/AD-2017 (Red line : MAXS/AD-2017).

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JAEA-Conf 2020-001QuantiﬁcationofIntegralDataEﬀectiveness
Using the Concept of Active Sub-Space in Nuclear Data TestingGo CHIBA, Daichi IMAZATO
Faculty of Engineering, Hokkaido University
Kita-13 Nishi-8, Kita-ku, Sapporo, Hokkaido, 060-8628 Japane-mail: go chiba@eng.hokudai.ac.jp¹
Severalmethodsbasedontheconceptoftheactivesub-spaceareproposedtoquantifyintegraldataeﬀectiveness,andtheyareadoptedtoexisting32integraldatawithmulti-groupcrosssectionrepresentation. Dimensions of sub-spaces spanned by integral data in a nuclear data space canbe quantiﬁed with the proposed method using orthogonal projections of each integral data vector
to the sub-space. In addition, a method to choose a minimum independent integral data set is
proposed, and it is demonstrated that this method can properly choose a wide variety of integraldata among a set of dependent integral data.

1 IntroductionIn the ﬁeld of nuclear and radiation engineering, a huge amount of experimental data relatedto the reactor physics and the radiation shielding have been obtained at various facilities in theworld,andsomeofthemhavebeenreleasedtothepublicasopendatatovalidatenumericaltoolssolvingreactorphysicsandradiationshieldingproblems. Experimentaldatawhichcanbeutilized
tovalidatenucleardataarereferredtoasintegraldataintheﬁeldofnucleardataengineering,and
somanyintegraldatahavebeenaccumulatedthroughinternationalprojectssuchasICSBEPand
IRPhEP. Those integral data have been eﬃciently utilized to test evaluated nuclear data ﬁles.
As described above, the number of available integral data has become enormous now, so it is
important to choose a proper set of integral data when testing evaluated nuclear data ﬁles with
them. To do so, dependency among integral data should be carefully examined, so a procedure
how to choose proper integral data is desired. We have proposed to adopt the concept of theactive sub-space to this problem in our previous study and have tested our method with a set of
existing integral data with one-energy group approximation[1] and a set of ﬁctitious integral datawith multi-group treatment[2] in the past.
Inthepresent work, weproposeanewproceduretoquantify independency ofasetofintegraldata and a new method to choose a minimum set of independent integral data among a hugeamount of integral data. These methods are adopted to actual integral data with multi-group
representation.

2 Theory
2.1 Basic concept of the proposed methodWeregardeachofnucleardata(ND)isavectorwhichisorthogonaltootherNDvectors,anda nuclear data space can be deﬁned from a set of these ND vectors. A sensitivity of integral datawithrespecttoNDisregardedasavectorinthenucleardataspace,andasetofsensitivityvectors
can span a sub-space, which we call an integral data space. Dimension of the integral data spacecanbedeﬁned, andisequaltoorsmaller thanthenumberofsensitivity vectors. Anorthonormalset of basis vectors of the integral data space can be also derived.

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JAEA-Conf 2020-001

If a ND vector exists on the integral data space, this means that this ND can be independently
validated with this set of the integral data. On the other hand, if a ND vector does not exist on the
integral data space, it is impossible to independently validate this ND with this set of the integral
data. In our method, possibility of the independent validation of each ND is quantiﬁed by a normof an orthogonally-projected vector of this ND vector to the integral data space. This is illustrated
in Fig. 1 where a two-dimensional integral data space and a ND vector are presented.

Fig. 1: Relation of an integral data space and a nuclear data vector

2.2

On quantiﬁcation of integral data space dimension

An orthonormal basis set of an integral data space can be obtained by the singular value
decomposition of a sensitivity matrix SI×J = (s1 s2 · · · sJ ) where sj is a sensitivity vector of the
�T�
dpj dpj
dpj···, where pj is a parameter of thejth integral data and is deﬁned as sj =dσ1 dσ2dσI
jth integral data and σi is the ith nuclear data. The superscript T is for vector transposition. The
numbers of nuclear data and integral data are denoted to as I and J here. To obtain dimension andan orthonormal basis of the integral data space, the singular value decomposition of S is carriedout as S = UDVT where UI×I = (u1 u2 · · · ul · · · uI ) and⎛ 2⎞
σ1 0 · · · 0 0 · · · 0⎜ 0 σ22 · · · 0 0 · · · 0 ⎟
⎜
⎟
⎜ ..
⎟
..
..
⎜ .
⎟
.
.
0
0
·
·
·
0
⎜
⎟
2
⎟,
0
0
0
σ
0
·
·
·
0
DI×J = ⎜
(1)
l
⎜
⎟
⎜ 0
⎟
0
0
0
0
·
·
·
0
⎜
⎟
⎜ .
⎟
..
..
..
.. . .
⎝ ..
.
.
.
.
.
0 ⎠
0
0
0
0 0 ··· 0
where l is the dimension and ui (1 ≤ i ≤ l) are orthonormal basis of the integral data spaceconsidered here.
Figure 2 shows singular values distributions of several sensitivity matrices obtained in theprevious study[1]. Generally, the same number of singular values as the number of integral data is
numerically calculated, so we need to determine a threshold value for singular values to separate
meaningful components from noise (meaningless) components. As this ﬁgure suggests, however, it
is very diﬃcult to do this since there is no clear boundary showing drastic change in the behavior.
In the present work, we propose the following procedure to determine the dimension of theintegral data space:
1. A set of orthonormal basis vectors is obtained by the singular value decomposition of asensitivity matrix. Set n=1.
2. Choose n principal basis vectors corresponding to the largest singular values, and construct
a n-dimensional sub-space spanned by these n basis vectors.

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JAEA-Conf 2020-001

101

7 data10 data
14 data
23 data0Singular value

10

10-1

-2

10

10-3

-4

10

5

10

15

20

Fig. 2: Example of singular values distributions of sensitivity matrices.[1] The X-axis corresponds
to the singular value index.
3. Each of sensitivity vectors is orthogonally projected to this sub-space, and norms of the
projected vectors are calculated. Note that sensitivity vectors are normalized so as to maketheir norms unity in this step.
4. If the minimum value of the norms obtained at the step 3 is smaller than a criteria, n isincreased by one, and a procedure from the step 2 is carried out again. If the minimum norm
is larger than this criteria, this procedure is terminated and the dimension of the integral
data space is determined as n.
This procedure is based on rather a physical aspect than a mathematical aspect simply setting acriteria on singular values. In the present work, this criteria is set 0.99. Note that obtained resultswith the proposed method should depend on this parameter setting.

2.3

A method choosing minimum independent data set

There are a huge amount of available integral data, but it is unrealistic to use all of them when
testing evaluated nuclear data ﬁles. Here we propose a method to choose a minimum independent
data set from integral data. The procedure is as follows:
1. The ﬁrst principal basis vector of the integral data space having the largest singular value istaken, and inner products of this vector and each of sensitivity vectors are calculated. After
doing this, a sensitivity vector which gives the largest inner product is chosen as the ﬁrst
integral data.
2. Construct a sub-space spanned by sensitivity vectors of the chosen integral data.
3. Each of non-chosen normalized sensitivity vectors is orthogonally projected to the sub-spaceconstructed at the step 2. If norms of all these projected vectors are larger than a criteria,
this procedure is terminated. If not, a sensitivity vector which gives the minimum norm ischosen, and go back to the step 2.
The criteria at the step 3 is arbitrary, and 0.99 is chosen in the present work.

3

Numerical result

The proposed method is adopted to a set of existing integral data shown in Table 1. All ofthese integral data are reactor physics parameters obtained at fast neutron systems, and intetgral- 201 -

JAEA-Conf 2020-001

data with indices 1 to 14 are for criticality (neutron multiplication factor k), and the others arefor ﬁssion reaction rate ratio at a core center position. F25, F28, F37 and F23 stand for ﬁssion
reaction rates of uranium-235, -238, neptunium-237 and uranium-233, respectively. Sensitivities arecalculated with 70-group cross section data based on JENDL-4.0. Energy mesh structure of this 70group cross section data is the same as the JAERI fast set-3, and a whole energy range is divided
to 70 with constant lethargy width except the ﬁnal group. Forward and (generalized) adjoint
angular neutron ﬂuxes are calculated by a discrete-ordinate neutron transport solver SNR of the
CBZ code system, and sensitivities are calculated with them based on the ﬁrst-order (generalized)
perturbation theory. Sensitivity vectors are normalized so as to make norms of them unity sincesensitivities about diﬀerent reactor physics parameters, k and reaction rate ratio, are considered
here.
Table 1: Integral data with their indices
Index
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16

Name
JezebelJezebel-Pu
Jezebel-233
GodivaFlattop-Pu
Flattop-U
Flattop-233
Big-tenThor (Pu w Th Ref.)
PMF010 (Pu w NU Ref.)
U3MF002-1 (U-233 w HEU Ref.)
U3MF002-2 (U-233 w HEU Ref.)
U3MF003-1 (U-233 w NU Ref.)
U3MF003-2 (U-233 w NU Ref.)
F28/F25 in Godiva
F37/F25 in Godiva

Index
17
18
19
20
21
22
23
24
25
26
27
28
29
30
31
32

Name
F23/F25 in Godiva
F49/F25 in Godiva
F28/F25 in Jezebel
F37/F25 in Jezebel
F23/F25 in Jezebel
F49/F25 in Jezebel
F28/F25 in Jezebel-233
F37/F25 in Jezebel-233
F28/F25 in Flattop-U
F37/F25 in Flattop-U
F23/F25 in Flattop-U
F49/F25 in Flattop-U
F28/F25 in Flattop-Pu
F37/F25 in Flattop-Pu
F28/F25 in Flattop-233
F37/F25 in Flattop-233

As nuclear data, we consider 8 reactions of (n,f), (n,γ), (n,n), (n,n’), (n,2n), μ̄, ν̄ and χ for thefollowing 10 nuclides: uranium-233, -234, -235, -238, plutonium-239, -240, -241, -242, thorium-232
and neptunium-237. Corresponding indices of these nuclear data are listed in Table 2. Since
the number of energy groups is 70, the dimension of the nuclear data space considered here is
8×10×70=5,600.
A normalized-singular values distribution of the sensitivity matrix is shown in Fig. 3. Tworesults with and without sensitivity vectors normalization are shown here. The eﬀect of the sensitivity vectors normalization on the singular values distribution is small in the present case.
The principal basis vectors having the largest singular values are shown in Fig. 4. On the ﬁrst
basis vector, the ﬁrst and second highest peaks are observed in (n,f) cross sections of uranium-235
and neptunium-237, respectively, and signs of these components are opposite to each other. Thiscan be considered due to strong contribution of ﬁssion reaction rate ratio data of F37/F25. On thesecond basis vector, the ﬁrst and second highest peaks are observed in ν̄ and (n,f) cross sections ofuranium-233. This can be also considered due to contribution of criticality data of uranium-233loaded cores. On the third basis vector, the ﬁrst and second highest peaks are observed in ν̄ and(n,f) cross sections of plutonium-239, and this would be due to contribution of criticality data ofplutonium-239-loaded cores.
A dimension of a sub-space spanned by these sensitivities is calculated as 13 with the procedure
mentioned above. Since the total number of integral data is 32, it can be concluded that the present
data set includes highly-dependent integral data. The result obtained by the proposed method is- 202 -

JAEA-Conf 2020-001

Table 2: Nuclear data with their index.
Index
17114121128135142149156163170177184191198110511121119112611331140114711541161116811751182118911961203121012171-

Nuclide
U-233
U-233
U-233
U-233
U-233
U-233
U-233
U-233
U-234
U-234
U-234
U-234
U-234
U-234
U-234
U-234
U-235
U-235
U-235
U-235
U-235
U-235
U-235
U-235
U-238
U-238
U-238
U-238
U-238
U-238
U-238
U-238

ND
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ

Index
22412311238124512521259126612731280128712941301130813151322132913361343135013571364137113781385139213991406141314201427142314411-

NuclidePu-239Pu-239Pu-239Pu-239Pu-239Pu-239Pu-239Pu-239Pu-240Pu-240Pu-240Pu-240Pu-240Pu-240Pu-240Pu-240Pu-241
Pu-241
Pu-241
Pu-241
Pu-241
Pu-241
Pu-241
Pu-241
Pu-242
Pu-242
Pu-242
Pu-242
Pu-242
Pu-242
Pu-242
Pu-242

ND
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ

Index
4481455146214691476148314901497150415111518152515321539154615531-

NuclideTh-232
Th-232
Th-232
Th-232
Th-232
Th-232
Th-232
Th-232
Np-237
Np-237
Np-237
Np-237
Np-237
Np-237
Np-237
Np-237

ND
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)
μ̄
ν̄
χ
(n,f)
(n,g)
(n,n)
(n,n’)(n,2n)μ̄ν̄χequivalent with ignoring components whose normalized singular values are less than around 0.075.
Finally, a minimum set of the independent integral data is also chosen by the proposed method.
As a result, the following 18 integral data are chosen. These are presented in the descending order
based on their priority: 32, 1, 11, 8, 25, 21, 4, 28, 19, 23, 5, 26, 27, 30, 7, 22, 9 and 2. Detailed
information on the ﬁrst ﬁve chosen integral data are provided in Table 3. A wide variety ofintegral data is properly chosen by the proposed algorithm.

4

Conclusion

Several methods based on the concept of the active sub-space have been proposed to quantify
integral data eﬀectiveness, and they have been adopted to existing 32 integral data with multi-groupcross section representation. Dimensions of sub-spaces spanned by integral data in the nuclear data
space can be quantiﬁed with the proposed method using orthogonal projections of each integral
data vector to the sub-space. In addition, a method to choose a minimum independent integral- 203 -

JAEA-Conf 2020-001

010

w normalization
w/o normalization-1

Normalized singular value

10

-2

10

-3

10

-4

10

-5

10

-6

10

-7

10

-8

10

-9

10

0

5

10

15

20

25

30

35

Singular value index

Fig. 3: Singular values distribution of a sensitivity matrix consisting of 32 actual integral data(a) From 1st to 4th(b) From 5th to 8th

0.3

0.3

0.2

0.2
Basis vector

Basis vector

0.1
0
-0.1
-0.2
1st
2nd
3rd
4th

-0.3
-0.4
-0.5

0

1000

2000
3000
4000
Nuclear data index

0.1
0
-0.1

5th
6th
7th
8th

-0.2
-0.3

5000

0

1000

2000
3000
4000
Nuclear data index

5000

Fig. 4: Principal basis vectors spanning the integral data space of 32 actual integral data
Table 3: Detailed information on the chosen integral data
Priority
1
2
3
4
5

Index
32
1
11
8
25

Parameter
F37/F25kk
k
F28/F25

Fuel material
U-233
Pu-239U-233
U-235, -238
U-235

Reﬂector material
U-238
U-235
U-238
U-238

data set has been proposed, and it has been demonstrated that this method can properly choose
a wide variety of integral data among a set of dependent integral data.

References[1] G. Chiba: “Application of active sub-space method to nuclear data integral testing,” the AESJ
spring meeting 2016, (2016).
[2] D. Imazato, G. Chiba: “Quantiﬁcation of integral data eﬀectiveness by active sub-space,”
Proc. of Reactor Physics Asia 2019, RPHA19, Osaka, Japan, Dec. 1-3, 2019, (2019).

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JAEA-Conf 2020-001

38

Uncertainty estimation of conventional neutron-spectrum
unfolding codes with Monte-Carlo based method

Katsumi AOKI1, Tadahiro KIN1, Md Kawchar Ahmed PATWARY1, Kosuke YOSHINAMI1,
Masaya YAMAGUCHI1, Masatoshi ITOH2, Yukinobu WATANABE1
1

Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga,
Fukuoka, 816-8580 Japan

2

Cyclotron and Radioisotope Center, Tohoku University, 6-3 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi,
980-8578 Japane-mail: aoki.katsumi.473@s.kyushu-u.ac.jp
Unfolding process has been applied to neutron spectra derivation of many measurement methods.

Previously, many types of unfolding techniques have been developed. Many codes of them, however, cannot
calculate uncertainty propagation, and thus uncertainty characteristics comparison of some codes has notbeen investigated. To solve this problem, we have developed an uncertainty estimation method based onMonte-Carlo technique. Moreover, we have investigated characteristics of uncertainty propagation of two
unfolding codes, GRAVEL and MAXED. For demonstration, we measured a double-differential thick-target
neutron yield (DDTTNY) of the C(d,n) reaction induced by 20-MeV deuterons with the multiple-foil
activation method at Cyclotron and Radioisotope Center, Tohoku University. The measured data wereanalyzed by GRAVEL and MAXED, and their uncertainty propagation was estimated by using the present
method. As a result, we found that the uncertainty of DDTTNY has neutron energy dependence, and thedependency trend is different between GRAVEL and MAXED codes.

1. Introduction
Accelerator-based neutron source has recently been applied to various fields such as boron neutron
capture therapy (BNCT) [1], radioisotopes (RIs) production [2, 3], nuclear transmutation [4], and nuclear
fusion engineering [5]. Among them, we have been focusing on medical RIs production using neutrons
generated by the (d,n) reaction with a neutron convertor made of thick carbon or beryllium. The generated
neutrons irradiate a raw material, and RI is produced. In the RI production system, 64Cu [6], 92Y [7], and99m

Tc [8] productions were proposed previously.
In research and development of the RI production method, numerical simulation plays an important rolefor estimation of production amount and its purity, the design of irradiation system, and shielding calculation.
For accurate simulation, double-differential thick-target neutron yields (DDTTNYs) should be preknowledge.
In practical RI production applications, the DDTTNYs of the (d,n) reactions with the energy range from 10to 50-MeV deuterons are necessary. However, systematically sufficient experimental data are not available.
Therefore, we have measured the DDTTNYs by the multiple-foil activation method. In this method,- 205 -

JAEA-Conf 2020-001

DDTTNY 𝜙𝜙�� is derived by unfolding process that is to solve an inverse problem expressed by thefollowing equation:

𝑅𝑅�,��
𝑁𝑁�
𝑁𝑁⎛𝑅𝑅�,��⎛ �⎞⎜ ⋮⎜ ⋮ ⎟� ⎜𝑅𝑅�,��
⎜ 𝑁𝑁� ⎟⎜⋮⋮
⎝𝑁𝑁� ⎠𝑅𝑅⎝ �,��

𝑅𝑅�,��
𝑅𝑅�,��

⋯⋯𝑅𝑅�,��
𝑅𝑅�,��

⋯⋯⋮𝑅𝑅�,��

⋯⋮𝑅𝑅�,��

⋱⋯⋮𝑅𝑅�,��

⋱⋯⋮𝑅𝑅�,��

⋯𝑅𝑅�,��
𝑅𝑅�,�� ⎞𝜙𝜙��
⎛ 𝜙𝜙�� ⎞⋮ ⎟⎜ ⋮ ⎟𝜙𝜙�� ⎟ ,𝑅𝑅�,�� ⎟ ⎜
⎜
⎟
⎟
⋮
⋮
𝑅𝑅�,�� ⎠ ⎝𝜙𝜙�� ⎠(1)

where 𝑁𝑁� and 𝑅𝑅�,�� represent the number of produced atoms via the reaction 𝑖𝑖 and the response functionof reaction 𝑖𝑖 with a neutron energy group 𝑗𝑗, respectively. In general, the number of activation reactions ofinterest (𝑛𝑛) is less than the number of neutron energy groups (𝑚𝑚). In the past, many unfolding algorithms

have been developed, for example, an iterative approximation method [9], a maximum entropy method [10],an iterative Bayesian method [11], and an artificial neural network (ANN) method [12]. However,characteristics of the propagated uncertainty of unfolding codes are not known.
To overcome the situation, we have developed a random sampling algorithm of estimating the uncertainty
propagation in DDTTNYs derived by the unfolding process. In our study, we pay attention to GRAVEL [9]and MAXED [10] codes which are conventionally used for neutron spectrum unfolding. These codes are
based on the iterative approximation method and the maximum entropy method, respectively. The GRAVEL
code cannot analytically calculate the propagation of experimental uncertainties. Using the random samplingalgorithm, we have compared the uncertainty propagation in the unfolding process with GRAVEL and
MAXED.
2. Monte-Carlo based uncertainty estimation method
We proposed an algorithm for estimating uncertainty based on Monte-Carlo technique for unfolding
process. The flow chart of the algorithm is shown in Fig. 1.
Firstly, measured number of produced atoms 𝑁𝑁� of interested reaction is randomly changed by Gaussian

distribution having statistical uncertainty. The changed number of atoms 𝑁𝑁�� is calculated by following

equation:𝑁𝑁�� � 𝑁𝑁� � 𝜀𝜀� 𝑁𝑁� 𝑑𝑑,(2)

where 𝑑𝑑 and 𝜀𝜀� represent a random number generated by standard normal distribution and experimental

relative uncertainty of 𝑖𝑖-th reaction. Secondly, we derived a neutron spectrum by an unfolding code with 𝑁𝑁�� .

The process, which 𝑁𝑁�� is generated and neutron spectrum is derived, is conducted iteratively. Then, neutron
spectra give two-dimensional distribution like Fig. 2. In each neutron energy group, we calculate standard

deviation of DDTTNYs. The standard deviations mean propagated uncertainty of DDTTNY. The algorithm
is considered only statistical uncertainty of the measured number of atoms 𝑁𝑁� .

- 206 -

JAEA-Conf 2020-001

Fig. 1 Flow chart of a random sampling algorithm for estimating propagated uncertainty
in unfolding process of deriving DDTTNYs.

Fig. 2 Example of two-dimension distribution of distorted DDTTNYs
for bins of neutron energy and DDTTNY.
3. Experiment
We performed DDTTNY measurement by multiple-foil activation method at Cyclotron and Radioisotope
Center (CYRIC), Tohoku University. Deuterons were accelerated to 20 MeV by an AVF930 cyclotron andbombarded on a thick carbon target (23 mm ൈ 23 mm ൈ 2 mmt). The accelerator-based neutrons weregenerated via the natC(d,n) reaction. The generated neutrons irradiated multiple foils (50 mm ൈ 50 mm). The

foils were made of 27Al (0.025 mmt), natFe (0.02 mmt), 59Co (0.1 mmt), natNi (0.1 mmt), natZn (0.2 mmt), natZr(0.1 mmt), and natMo (0.05 mmt). The foils placed at 1,215 mm downstream to the carbon target at 0 degree- 207 -

JAEA-Conf 2020-001

to direction of deuteron beam. Average deuteron beam current was about 2.0 μA during 19-hours irradiation.

Fig. 3 shows the schematic view of irradiation setup. After the irradiation, we measured gamma-ray emitted
from activated multiple foils with HPGe detector.

Fig. 3 Schematic view of the irradiation setup. The accelerator-based neutrons generated via
the natC(d,n) reaction irradiated multiple foils (27Al, natFe, 59Co, natNi, natZn, natZr and natMo).
4. Results and Discussion
Firstly, the number of produced atoms by activation reactions of interest was derived from counting rateof photo peaks in measured gamma-ray spectra. The derived number of atoms and their statisticaluncertainties are shown in Table 1.
Table 1 Interested activation reaction, the number of produced atoms,
and statistical uncertainty measured in the gamma-ray experiment.
Reaction

Half life

gamma-ray energy

Number of produced atoms

24

Al(n,a) Na

15.0 h1369 keVFe(n,p)56Mn

2.58 h1811 keVCo(n,a)56Mn

2.58 h846.8 keV

5.30�107 � 2.30�105

59

Co(n,p)59Fe

44.5 d1099 keV

58

Ni(n,2n)57Ni

35.6 h1378 keV

27
56
59

64

Zn(n,p)64Cu

12.7 h1346 keV

67

Zn(n,p)67Cu

61.8 h

184.6 keV

Zr(n,2n)95Zr

64.0 d756.7 keVMo(n,p)96Nb

23.4 h568.8 keV9696

1.28�107 � 1.37�105
1.90�107 � 1.12�105

2.68�108 � 1.38�106
1.24�107 � 1.03�105

6.91�108 � 9.56�106
7.59�106 � 7.76�104
7.00�107 � 7.49�105

1.63�106 � 2.47�104

Secondly, we calculated response functions (𝑅𝑅�,�� ) of unfolding process by following equation:
𝑅𝑅�,�� � 𝜎𝜎�,�� � Ω � 𝜌𝜌� � 𝐶𝐶 � 𝐶𝐶� ,(3)2

where 𝜎𝜎�,�� is the cross section [cm ] stored in JENDL-4.0 [13], Ω is solid angle of each foil [sr], 𝜌𝜌� issurface density of atoms [cm-2], 𝐶𝐶 is total charge of deuteron beam [μC], and 𝐶𝐶� is decay correction. Only
cross section is a function of neutron energy, other parameters are constant values. The cross sections areshown in Fig. 4 (a).

Both GRAVEL and MAXED require initial guess spectrum in unfolding process to derive resultant
spectrum. The initial guess spectrum was calculated by deuteron-induced reaction analysis code system- 208 -

JAEA-Conf 2020-001(DEURACS) [14] which is a theoretical calculation code for deuteron induced reactions. The calculated
spectrum is shown in Fig. 4 (b).

Cross section [barns]

101

(a)

100

(b)

10-1
10-2
10-3
10-4

0

5
10
15
Neutron Energy [MeV]20

27Al(n,a)24Na

56Fe(n,p)56Mn

59Co(n,a)56Mn

59Co(n,p)59Fe

58Ni(n,2n)57Ni

64Zn(n,p)64Cu

67Zn(n,p)67Cu

96Zr(n,2n)95Zr

96Mo(n,p)96Nb

Fig. 4 For unfolding process, (a) cross sections plotted as a function of neutron energy for each activation
reaction and extracted from JENDL-4.0. (b) Initial guess spectrum is calculated by DEURACS which istheoretical calculation model.
The propagated uncertainties of each code were calculated by the developed algorithm. The standard
deviation 𝜀𝜀� of eq. (2) was set to the original experimental statistical relative uncertainty. Then, 1,000 spectra
were calculated by random sampling as shown in Fig.1. Fig. 5 shows two-dimensional histograms of the
1,000 DDTTNYs for case of DDTTNY derivation in (a) GRAVEL and (b) MAXED codes. For each code,

we found that uncertainties have neutron energy dependence and the trend is completely different even theyused same original uncertainties. In resultant DDTTNY of GRAVEL has large propagated uncertainty around
1-7 MeV of neutron energy range. In contrast to that, resultant DDTTNY of MAXED has large propagateduncertainty around higher energy range around 7-16 MeV. Moreover, the resultant spectrum is almost sameas the initial guess spectrum in lower energy range (1-5 MeV). That is because our interested reactions were
not occurred in this energy range, and unfolded DDTTNYs never adjusted in the range in the unfolding
process by MAXED which is well-known to have strong initial guess spectrum dependency.
5. Conclusion
We have developed an algorithm for estimating propagated uncertainty in unfolding process of deriving
DDTTNY. The algorithm is based on Monte-Carlo method and can apply to many unfolding codes. In thiswork, we applied it to GRAVEL and MAXED used in unfolding process and investigated difference ofneutron energy dependence of propagated uncertainty between two unfolding codes. For demonstration, we
conducted accelerator-based neutron spectrum measurement by multiple-foil activation method. We derived
the DDTTNY and its uncertainty by using our developing algorithm. From the results, we found neutron- 209 -

JAEA-Conf 2020-001

energy dependency and trend of the dependencies are completely different.
In the future, we will apply the algorithm to other unfolding codes based on the iterative Bayesian andartificial neutral network methods.
(b) MAXED

Frequency

DDTTNY [n/MeV/sr/μC](a) GRAVEL

Neutron Energy [MeV]Fig. 5 Preliminary results about two-dimensional histogram of 1000 DDTTNYs derivedin (a) GRAVEL and (b) MAXED.
References[1]

M. Yoshioka, Review of Accelerator-based Boron Neutron Capture Therapy Machines, in Proceedingsof IPAC 2016, 3171, 2016 (2016).

[2]

Y. Nagai, et al., J. Phys. Soc. Jpn. 82, 064201 (2013).

[3]

T. Kin, et al., J. Phys. Soc. Jpn. 24, 011031 (2019).

[4]

Implications of Partitioning and Transmutation in Radioactive Waste Management, Number 435 inTechnical Reports Series, IAEA, Vienna (2004).

[5]

A. Moeslang, et al., Fusion Engineering and Design 81, 863 (2006).

[6]

T. Kin, et al., J. Nucl. Sci. Technol. 54, 1123 (2017).

[7]

T. Kin, et al., M. Itoh, EPJ Web Conf. 146, 08009 (2017).

[8]

Y. Nagai, et al., J. Phys. Soc. Jpn. 78, 033201 (2009).

[9]

M. Matzke, Report PTB-N-19 (1994).

[10]

M. Reginatto, et al., Nucl. Instrum. Methods, A476, 242 -246 (2002).

[11]

T. Adye, Proc. of the PHYSTAT 2011 Workshop, 313 (2011).

[12]

Y. Sanzen, et al., Proc. 2016 Symp. Nucl. Data, JAEA-Conf 2017-001, 157 (2017).

[13]

K. Shibata, et al., J. Nucl. Sci. and Technol. 48, 1 (2011).

[14]

S. Nakayama, et al., Phys. Rev. C 94, 014618 (2016).

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JAEA-Conf 2020-001

39

The Influence of B and Li Burnable Poison on
Effective Multiplication Factor in the HTGR

Yuki KOGA1, Hideaki MATSUURA1
1
Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University,
744 Motooka Nishi-ku, Fukuoka, 319-0395 Japane-mail: koga_yuki@nucl.kyushu-u.ac.jp

Tritium is required for the engineering tests and the initial DEMO fusion reactor. In order to establish a tritium
supply method, tritium production using the high temperature gas-cooled reactor (HTGR) was proposed. The B
burnable poison (BP) is replaced with Li for tritium production. Li is loaded as a solid unit called Li rod. There isa possibility that difference in an effective multiplication factor (keff) arises by using Li instead of B. The keffdifference in the HTTR is evaluated and discussed. The same evaluation is also conducted for the GTHTR300. It
is revealed how the difference appears and how the HTGR types influence on the results.

1. Introduction
A fusion reactor using T(d,n)α reaction requires tritium for its R&D. Tritium is produced artificially because it
hardly exists in nature, while deuterium is obtained from water. Approximately 400 g of tritium is burned per day
in a 3 GW thermal output power. A sufficient amount of initial tritium is required to start up a fusion reactor. It
was reported that this amount of tritium for the Demo reactor was several 100 g -approximately 27 kg [1,2]. In
addition, 100 g or more of tritium is required for an engineering test using tritium before construction of DEMO
reactors. The tritium consumed on a fusion reactor for the ITER project has been produced by using the Canadian
Deuterium Uranium reactors, although it has not been clarified that the way to supply tritium for the initial DEMO
reactors, in particular Japan. Therefore, we proposed a tritium production method using the high temperature gascooled reactor (HTGR) [3] by loading a Li compound as a burnable poison (BP). The HTGR is composed bymainly graphite (moderator) and He (coolant). They are chemically stable and do not react with the Li compound.
An enough amount of Li can be loaded into the HTGR without 6Li concentration. It is because the HTGR core
size is larger than other types of fission reactor. Normal BP (B rod) is a cylindrical B4C and solid states so that Licompound is loaded as same shape without significantly changing the original core design. We consider Li
compound is loaded as Li rod. The Li rod includes LiAlO2 in a cylindrical alumina case with Ni coated Zr pebbles(tritium absorber) [4]. It can produce and contain tritium during the operation time. We are planning an irradiation
test on the High Temperature engineering Test Reactor (HTTR) [5] in order to confirm the Li rod performance andto demonstrate the tritium production on the HTGR [6].
The absorption cross-sections and loaded amounts are different in Li and B. Accordingly Li rod reactivity may
be different from B rod and the HTGR characteristic may be changed by loading Li rods. In order to operate the
HTGR loaded Li rods, it needs to clarify the characteristic change. The purpose of this paper is the evaluation anddiscussion of the HTGR reactor characteristic difference in the B rod and the Li rod. The effective multiplication
factors (keff) during operation are evaluated as the reactor characteristics. We selected the HTTR with 30 MW
thermal output [5] and the Gas Turbine High Temperature Reactor 300 (GTHTR300) with 600 MW thermal output- 211 -

JAEA-Conf 2020-001[7] for the evaluation. The HTGR characteristic difference by the BP change is also discussed for the two reactors.

2. Analysis model
We assumed the original designs of the B rod for the evaluation. Fig.1 shows the designs of the B rod for the
HTTR [8] and GTHTR300 [7]. We designed the Li rod to fit each BP hole sizes. Fig.2 shows the Li rod design for
the HTTR. Its diameter is 14 mm, height is 450 mm, Al2O3 layer thickness is 1.9 mm, Zr layer thickness is 0.1
mm, LiAlO2 thickness is 2.7 mm and hollow radius is 2.3 mm respectively. The hollow inside the Li rod is full ofZr pebbles coated Ni which diameter is 1 mm. Fig.3 shows the Li rod design for the GTHTR300. Its diameter is
44 mm, height is 950 mm, Al2O3 layer thickness is 6.3 mm, Zr layer thickness is 1 mm, LiAlO2 thickness is 2.5mm and hollow radius is 12.2 mm respectively. The hollow is full of the same Zr pebbles coated Ni.

Fig.1. A schematic view of the normal B rods for the HTTR (a) and the GTHTR300 (b).

Fig.2. A schematic view of the Li rod for the HTTR.

Fig.3. A schematic view of the Li rod for the GTHTR300

In order to evaluate keff and the cumulative weight of produced tritium, we conducted nuclear burning
calculations using the continuous-energy Monte Carlo transport code MVP-BURN [9,10] with a nuclear data
JENDL-4.0 [11]. We assumed a HTTR core system with 300 B rods or 450 Li rods and a GTHTR300 core systemwith 2160 B rods or 2160 Li rods. Therefore, the amounts of loaded 6Li for the HTTR and the GTHTR300 are 193
g and 6674 g respectively. The operation period in these calculations is 360 days and all control rods were assumed
to be pulled out. We set the time steps in the calculations were 0, 1, 5, 30, 60, 120, 180 and 360 days. 600,000
neutrons were generated for each the time steps. The statistical error of keff were less than 0.1 % in the all- 212 -

JAEA-Conf 2020-001

calculation, which is sufficient accuracy. Zr was not included in those calculations because Zr does not influenceon keff and tritium production by its cross-section.

3. Results and discussion
Fig.4 shows the accumulative produced tritium and keff (B and Li rod) by the operation time for the HTTR. The
keff maintained nearly 1.05 during the operation when the B rods were loaded, which corresponded with the keff in
the reference [5]. This is because the B rods suppress excessive reactivity by burning of 10B atoms inside them inthe early period of operation, and the B rods decrease their negative reactivity in the latter period of operation. The
keff decreased straightly during the operation when Li rods were loaded and produced 30.3 g of tritium in 360 days.
6

Li atoms in the Li rods are not burned quicker than 10B atoms and remain more in the late time of operation,

because the cross-section of 6Li is 940 b for thermal neutron whereas that of 10B is 3837 b for it. The keff decreased
immediately after the operation starts in the both of those calculations, by Xe generates.
The accumulative produced tritium and keff (B and Li rod) by the operation time for the GTHTR300 is shown in
Fig.5. When the B rods were loaded, the keff increased until 360 days because the design of the GTHTR300 is
assumed 2 years continuous operation. It was same tendency to the reference [7]. When the Li rods were loaded,
the keff decreased straightly as well as in the case for the HTTR and they produced 818 g of tritium. However, thedifference of keff by the BP change during the operation for GTHTR300 was more than that for HTTR. It means
there is more neutron to suppress by control rod in the GTHTR300 operation.

Fig.4. The accumulative produced T and keff

Fig.5. The accumulative produced T and keffby the operation time for the HTTR.

by the operation time for the GTHTR300.

At first we confirmed the amount of loaded LiAlO2 and the volume of a fuel block in Table 1 to investigate thisreason. The amount of loaded LiAlO2 into the GTHTR300 is about 7 times of that into the HTTR in order toproduce 800 g of tritium, while the difference of fuel block volume is smaller. Therefore, the Li/Block volume
ratio for the GTHTR300 is about 9 times of that for the HTTR. The increase of the ratio decreases 6Li burnability
by self-shielding effect when the LiAlO2 shapes are same.

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JAEA-Conf 2020-001

Table 1. The amount of loaded LiAlO2 into The HTGRs.

Next, we analyzed the outputs of the nuclear burning calculations. Fig.6 shows the amount of burned BP atoms
by the operation time. 10B atoms were burned quicker than 6Li for both of the HTGRs in those calculations, whichfits to the above theory. The amount of burned 6Li atoms for the HTTR was about 10 mol and almost same to thatof 10B atoms (the difference is about 3 %). On the other hand, the amount of burned 6Li atoms for the GTHTR300
was about 270 mol and more than that of 10B atoms, which difference is about 21 %. It was caused by loadinglarge amount of LiAlO2 to produce maximum tritium in the GTHTR300. Fig.7 shows the ratio of remained BP by
the operation time. When the B rods are loaded, the remain ratio of the HTTR was 28 %. It was more than that of
the GTHTR300 slightly. The reason is two B rods are loaded into each HTTR fuel blocks while three B rods areloaded into those of the GTHTR300. 10B atoms in the HTTR is less burn than that in the GTHTR300 by theincrease of Li heterogeneous. when the Li rods were loaded, the remain ratio of the GTHTR300 was 75 %, whereas
the ratio of the HTTR was 69 %. It follows the relation of the Li/Block volume ratio between 6Li burnability by
self-shielding effect. Therefore, excessive load and low burnability of 6Li on the GTHTR300 makes the keff during
the operation bad by neutron efficiency reduction.
It was reported that the GTHTR300 has potential to produce additional 40 % amount of tritium which the Li
rods can produce [12]. However, it does not show the way to improve the keff but one to increase the amount of
produced tritium. The decrease of Li heterogeneous can improve the keff and the remain ratio because self-shielding
effect is weakened and 6Li becomes more burnable. Loading many smaller Li rods is one of the solutions to reduce
Li heterogeneous. We conducted a model calculation under the assumption that 6Li was mixed into the GTHTR300
fuel blocks so as to show the improvement example. Fig.8 is the accumulative produced tritium and keff by theoperation time for the GTHTR300 under this assumption. 4718 g of 6Li was mixed into the fuel blocks in thiscalculation to keep keff more than 1.03 after 360 days operation. All the Li rods for the GTHTR300 included 6674
g of 6Li). The keff during the operation maintained nearly 1.05 until 180 days, which was similar to the case that B
rods were loaded. The ratio of remained 6Li was 63 % consequently the amount of produced tritium achieved 873
g.

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JAEA-Conf 2020-001

Fig.6. The amount of burned BP atoms

Fig.7. The ratio of remained BP

by the operation time.

by the operation time.

Fig.8. The accumulative produced T and keff by the operation time
for the GTHTR300 under the assumption Li are mixed in the fuel blocks.

4. Conclusion remarks
We evaluated effective multiplication factors (keff) during operation of the HTTR and the GTHTR300 for theevaluation and discussion of the HTGR reactor characteristic difference in the B rod and the Li rod. The keff kept
up to a value or increased during the operation when the B rods were loaded into the HTGRs. The keff decreased
straightly during the operation when Li rods were loaded for tritium production because 6Li atoms are not burned
quicker than 10B atoms and more remain in the late time of operation.
The excessive load of LiAlO2 to produce 800 g of tritium using the GTHTR300 increases more the differenceof keff by the BP change than that using the HTTR to produce 30 g of tritium. The GTHTR can improve its keffcharacteristic by decrease Li heterogeneous, which reduces the amount of loaded LiAlO2 and increases 6Li
burnability.

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JAEA-Conf 2020-001

References[1] R. Hiwatari, J Atomic Energy Society Japan [in Japanese] 60 (8) (2018) 52-56.
[2] Y. Asaoka., et al., Fusion Tech. 30 (1996) 853-863.
[3] H. Matsuura, et al., Nucl. Eng. Des.243(2012) 95-101.
[4] H. Matsuura, et al., Fusion Eng. Des. 146 (2019) 1077-1081.
[5] S. Saito, et al., JAERI 1332 (1994).
[6] Y. Koga, et al., Fusion Eng. Des. 136 (2018) 587-591.
[7] T. Nakata, et al., JAERI-Tech 2002-066, (2002) [in Japanese].
[8] N. Fujimoto and N. Nojiri, JAEA-Technology 2005-008, (2006) [in Japanese].
[9] Y. Nagaya, et al., JAEA-Data/Code 2016-019, (2017) [in Japanese].
[10] K. Okumura, et al., J. Nucl. Sci. Tech., 37 (2000), 128-138.
[11] K. Shibata, et al., J. Nucl. Sci. Technol. 48(1) (2011) 1-30.
[12] S. Nagasumi et al., Spring Meeting of AESJ (2016) 2L06.

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JAEA-Conf 2020-001

40 Evaluation of radioactive concentration produced in electric equipments
and materials on the decommissioning of nuclear power plants
Hirokuni YAMANISHI1*, Genichiro WAKABAYASHI1, Takahiro YAMADA1,
Masateru HAYASHI2, Tetsushi AZUMA2, Takayuki IKEDA3
1Atomic Energy Research Institute, Kindai University,
Higashi-Osaka, Osaka, 577-8502, Japan
2Advanced Technology R&D Center, Mitsubishi Electric Corporation,
Amagasaki, Hyogo, 661-8661, Japan.
3Power Plant Engineering & Construction Center, Mitsubishi Electric Corporation,
Osaka, Osaka, 661-8661, Japan.
*E-mail: yamanisi@kindai.ac.jp
A various cables used in nuclear power plants are treated as wastes in decommissioning. If part ofthem is treated as non-radioactive wastes, the amount of radioactive waste can be reduced. Furthermore,valuable resources such as copper can be recycled, and that contributes to the efficient use of resources. In
this study, we have evaluated the produced nuclides in representative PWR cables by neutron irradiation
calculated with the activation cross section data. As the results of evaluation, we have confirmed thepossibility of the Ni-63 concentration exceeding the criterion for clearance near reactor vessel. To improve
the accuracy of inventory evaluation, the uncertainty evaluation of activation cross section will be important.
1. Introduction
Aging nuclear power plants will increase in the coming decades, and most of them are expected to
be shut down. Therefore decommissioning of nuclear power plant is one of the most important issue.
Decommissioning of 1 GW class PWR (Pressurized Water Reactor) is estimated to generate hundreds
kilotons of waste. If the part of them could be treated as non-radioactive wastes, the amount of radioactive
waste could be reduced. Furthermore, valuable resources such as copper which used in electric equipments
can be recycled, and that contributes to the efficient use of resources.
Radioactivity concentration of core component and main structure such as concrete or steel has
been evaluated in several leading plant[1, 2], but sufficient investigation has not been conducted on electriccomponents. Most of the cables laid outside the reactor pressure vessel are considered as non-radioactive
waste due to relatively low neutron flux and small neutron cross section of copper.
In this study, we have evaluated the radioactivity inventory of copper material used in electric
equipments.
2. Method
The evaluation method consists of the following three steps: (1) Chemical composition analysis oftarget material, (2) Neutron flux and spectrum evaluation at target position, (3) Activity inventory calculation.
(1) Chemical composition
analysis(2) Neutron flux, spectrum
evaluation(3) Radioactivity inventory
calculation

 Investigation of structural
materials for each cable type

 Investigate neutron flux andspectrum of typical PWR plant Consideration of cable layout

 Consideration of irradiation andcooling period
 Comparison with criterion value

Fig. 1 Steps of evaluation.
(1) Chemical composition analysis of target material
Cables used in PWR plants are roughly classified into five types: (A) High voltage cable, (B) Low
voltage cable, (C) Control cable, (D) Instrumentation cable, and (E) Special instrumentation cable. Figure 2
shows the cross sectional view of typical cable and table 1 shows the structural material of each layer. Then
we have performed the chemical composition analysis using SEM/EDX (Scanning Electron Microscope /- 217 -

JAEA-Conf 2020-001

Energy Dispersive X-ray Spectroscope) and ICP–OES (Inductivity coupled plasma optical emission
spectrometer) for principal component. In addition, we have performed activation analysis using UTRKINKI to investigate trace components below the detection limit of ICP–OES analysis.

1
2
3
4
5
6

Table 1 Structural material of each layer.
Cable type
Layer
A: High voltage B: Low voltage
C: Control
D: Instrument
E: Special
Conductor
Copper
TPCS
TPCS
TPCS
TPCS
Insulator
CLPE
FREP / FRVFREP / FRVFREP / FRV
FRV / CLPE
Strand
Jute
Jute
Jute
Jute
Jute
ShieldCopper tape
TPCT
TPCT
TPCT
TPCT
Tape
FabricFabricFabricFabricFabric / PS
Sheath
FRV / FRPEFRV / FRPEFRV / FRPEFRV / FRPEFRV / FRPE
TPCS: Tin plated copper strand
TPCT: Tin plated copper tape
CLPE: Cross-linked polyethylene
FREP: Flame retardance EP rubber
FRV: Flame retardance vinyl
FRPE: Flame retardance polyethylene
PS: Polystyrene(2) Neutron flux and spectrum evaluation at target position
Since most of cables are located outside the reactor
pressure vessel in PWR plant, it is considered that thermal neutrons
are dominant at the target position. Therefore, in this study we havefocused on thermal neutrons. However, the thermal neutron flux attarget position depends on the reactor type, then we used 1011 n/cm2/s
at the surface of reactor pressure vessel as the typical value of PWR
plant[2, 3].

3

1

4, 5
2
6

Fig. 2 Cross sectional view oftypical cable.

(3) Radioactivity inventory calculation
In this study, radioactivity inventory calculation was performed using PHITS 3.17 and DCHAINSP[4]. Considering that most of the reactors in Japan will be shutting down in 40 years and start thedecommissioning process, we have set the irradiation period to 40 years. The neutron flux was constantduring irradiation period without any suspension period due to the maintenance or accident. After irradiation
period, the radioactivity concentration of each nuclide was compared with clearance level[5] as criterion.
3. Results
3.1 Chemical composition analysis of target material(1) SEM/EDX measurement
We have performed SEM/EDX measurement to investigate the main component of each layer ofcable. Each measurement samples that selected to cover the structural materials polished the cut surface after
fill the resin. SEM/EDX measurement was performed using Hitachi High Technologies Inc. SU-8000 withthe acceleration voltage was 15 kV. As a results of measurement, metal elements except copper were notdetected from the conductor layer. On the other hand, Na, Mg, Al, Si, Cl, Ca, Zn, Sb and Pb were detectedfrom the other layers.
(2) ICP–OES measurement
We have performed ICP–OES measurement to investigate the trace component. Considering theSEM/EDX results, we have selected element to be measured for the following reasons: Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn: Elements have the potential to produce 63Ni, 60Co and 54Mn
by neutron reaction Mg, Al, Si, Ca, Sn, As, Bi, Pb: Elements have the potential to be contained as a trace element in
conductor
For ICP–OES, each measurement sample was dissolved by adding acid and heating after disassemble andseparated into conductors and layers. ICP–OES measurement was performed using SII Nano Technology Inc.
SPS-3100. Table 2 shows the result of measurement. Ca and Sn were detected as trace element. The detection
limit was approximately several tens of ppm.

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JAEA-Conf 2020-001

Element
Cu
Ca
Sn

Table 2 Detected elements in conductor.
Composition (wt%)
Detection limit
99.95
2.1 ppm
1.192×10−3
8.3 ppm
0.051
21 ppm(3) Activation analysis
To investigate trace elements less than the detection limit of ICP–OES measurement, we have
performed activation analysis using UTR–KINKI. Each measurement sample that was disassemble andseparated into each layer filled in U–8 vessel has irradiated in the central stringer of the reactor. After theirradiation, we have performed the radionuclide analysis produced from trace element by the gamma
spectroscopy using a HPGe detector. As a result of analysis, any significant trace elements were not detected.
Therefore, we have decided the standard chemical composition of cable conductor as shown in table 2

Parallel beam

3.2 Radioactivity inventory calculation
At the beginning, we have calculated average
Sheathneutron flux inside the conductor region using PHITS
3.17 with parallel neutron beam to cable side. Figure 3
Shield, Tapeshows the two dimensional neutron flux.
Insulator
Then we have calculated the radioactivity
concentration in the conductor using DCHAIN-SP withJENDL/AD-2017[6] activation cross section. The
irradiation period was 40 years, and the concentration
Conductor
was calculated from 20 years after the irradiation
Strandstopped. Figure 4 shows the calculation results. As aresult of calculation, 41Ca, 63Ni, 65Zn, 119, 121, 121mSn, 125Sb
Void
and 125mTe was confirmed as a long half-life nuclide.
Compared to the clearance level, radioactive
Fig. 3 Example of PHITS calculation
concentration of 63Ni and 65Zn are greater than the(Two dimensional neutron flux).
119, 121, 121mSn,
clearance criterion at 40 years. Note that
125
125m
Sb,
Te are excluded in this study due to the domestic regulation. Table 3 shows the calculated
radioactivity concentration of 41Ca, 63Ni and 65Zn at 40, 50 and 60 years.

Fig. 4 Calculated radioactivity concentration (left) and ratio to clearance level (right).

Nuclide
41

Ca

63Ni

65

Zn

Table 3 Calculated radioactivity concentration.
Clearance
Radioactivity concentration (Bq/g)
T1/2
level (Bq/g)
40 y
50 y
60 y
1.03×105 y
100
1.85
1.85
1.85
101.2 y100
2.71×103
2.53×103
2.36×103244.06 d
0.1
1.02×105
3.16
9.81×10−5- 219 -

JAEA-Conf 2020-001

4. Discussion
As a result of calculation, 65Zn is a dominant nuclide present in a cable as the irradiation is stopped,
63Ni becomes dominant in a few years due to the short half-life of 65Zn. Moreover, 10 years after
however
irradiation stops, the radioactivity concentration of 63Ni becomes more than 90% of total amount.
In this case, evaluation of 63Ni radioactivity concentration must be required for clearance. However,direct measurement of 63Ni is difficult due to its low energy β ray emission (maximum 0.0659 MeV, average
0.0174 MeV[7]) and no γ ray emission.
Therefore, it is necessary to consider evaluation using calculations including uncertainty evaluation.
For this purpose, uncertainty evaluation of activation cross section is important. Figure 5 shows thecomparison of cross section data between nuclear data libraries. The cross section of 40Ca(n, γ)41Ca reactionin each library is almost same. On the other hand, the cross section of 63Cu(n, p)63Ni reaction has a large
difference of about 10 times between libraries, especially in the thermal neutron region where no
experimental data has been obtained. For this reason, it is considered that the uncertainty of the nuclear data
contributes significantly to the uncertainty of the 63Ni radioactivity inventory evaluation.

Fig. 5 Comparison between activation nuclear data libraries and experimental data of 63Cu(n, p)63Ni
reaction (left) and 40Ca(n, γ)41Ca reaction (right).
5. Conclusion
In this study, we have evaluated the radioactivity inventory of copper material used in electric
equipments by the basis of (1) Chemical composition analysis of target material, (2) Neutron flux andspectrum evaluation at target position and (3) Activity inventory calculation. As a result of evaluation, we
have confirmed the radioactive concentration of 63Ni and 65Zn in the cable near the reactor vessel may exceed
the clearance criterion and the radioactivity concentration of 63Ni becomes more than 90% of total amount.
after 10 years from irradiation stops. Since the direct measurement of 63Ni is difficult, it is necessary to
improve uncertainty evaluation of inventory calculation. For that purpose, the uncertainty evaluation ofactivation cross section is important.
References[1] T. Sukegawa, M. Hatakeyama, S. Yanagihara, “Evaluation of Methodology on Radioactive Inventory
Estimation in the Japan Power Demonstration Reactor Decommissioning Program”, J. Nucl. Sci. Technol.
37, p.367 (2000).
[2] Electric Power Research Institute Inc., “Radiological Characterizations for Reactor Pressure Vessel andInternals Enhanced SAFSTOR”, EPRI Technical report 1008018, (2003).
[3] C. Le Loirec, M. Soulard, G. Ritter, Y. Pénéliau, “Benchmark of DEMAIN, a new calculation package
for decommissioning purposes”, Prog. Nucl. Energy, p.118 (2020).
[4] T. Sato, Y. Iwamoto, S. Hashimoto, et. al., “Features of Particle and Heavy Ion Transport code System(PHITS) version 3.02”, J. Nucl. Sci. Technol. 55, p.684 (2018).
[5] IAEA, “Application of the Concepts of Exclusion, Exemption and Clearance, Safety Standard Series No.
RS-G-1.7”, (2004).
[6] K.Shibata, N.Iwamoto, S.Kunieda, et. al., “Activation Cross-section File for Decommissioning of LWRs”
JAEA-Conf 2016-004, p.47, (2016).
[7] M.G. Stabin and L. Luz, "Decay data for internal and external dose assessment", Health Physics Vol. 83,
p.471-475, (2002).

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2020-001

41

Evaluation of scattered radiation on the contralateral breastin breast-conserving therapy
Rikuto HARA1, Ayano MAKINAGA1

1Department of Radiological Technology, Teikyo University, 6-22, Misaki-machi,

Omuta, Fukuoka 836-8505, Japane-mail: 16RT044e5@stu.teikyo-u.ac.jp
Breast-conserving therapy is a standard local treatment for breast cancer up to stage II after breastconserving surgery. In this study, we evaluated the effect of scattered radiation on the contralateral breast by
the difference in irradiation conditions on the affected breast using Monte Carlo simulation code PHITS. The
contralateral dose was less than the target dose farther away from the inner edge of the affected breast. In
addition, the larger the breast size, the more the effect of scattered radiation increased.

1.

Introduction
Breast-conserving therapy is a standard modality for breast cancer up to stage II after breast conserving

surgery. An irradiation method called “tangential irradiation” that applies irradiation from two directions isusually used as a standard irradiation technique for breast irradiation. A field-in-field method or intensity
modulated radiation therapy (IMRT) is also used to make a uniform dose distribution. In addition, respiratory
gated irradiation may be used to reduce exposure for an organ at risk such as the lung or heart. The standard
treatment requires a treatment period of 4.5~5.5 weeks with 45~50.4Gy/25~28 fractions [1]. In breastconserving therapy, radiation therapy has been reported to significantly reduce local recurrence rates andcontribute to the survival rate. However, radiation therapy has advantages and disadvantages. For example,
one of the early effects is dermatitis. When choosing the dose rate, radiation induced pneumonia andmammary gland hardening should be also considered carefully. Case-control studies and large cohort studies
show that postoperative irradiation of breast cancer increases the incidence of contralateral breast cancer. In
a report by Taylor et al., an irradiation risk ratio of 1.20 was showed for contralateral breast cancer [2]. Oneof the causes has been suggested to be the effect of low-dose exposure to the contralateral breast due toscattered radiation.
In this study, we evaluate the effect of scattered radiation on the contralateral chest using the Monte Carlosimulation code PHITS [3].

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2020-001

4 or 6 MeV

Figure 1 Geometry definition in the PHITS code. A voxel space with a height of 384 cm, a width of 137 cm, anda depth of 299 cm was used. The human voxel phantom, which is 163 cm tall, was placed in the voxel space.

2. Simulation study of radiation therapy using PHITS
To evaluate the effect of scattered radiation dose on the contralateral chest, we used Monte Carlo simulation
code PHITS. Medical Linac and the human voxel phantom (ICPR 110: female phantom) were defined in thecalculation. As seen in Figure 1, an X-ray radiator was constructed with W and Cu material. The collimator
material was defined with 184W-90.5 %, 58Ni-6.5 %, 56Fe-3.0 %, and the flattening filter is made from 63Cu69.0 %, 65Cu-31.0 %. In the calculation, we used typical X-ray energy of 4 MV and 6 MV. The field size ofthe irradiation was 10 cm x 10 cm. To determine the irradiation angle and isocenter for the chest phantom,
we used the calculation algorithm reported in [4]. The isocenter is the area where radiation is mostconcentrated in radiation therapy. The calculated isocenter coordinate was (x, y, z) = (-21 cm, 19 cm, 5 cm).
Irradiation angles of tangential irradiation from two
directions were 60 degrees and 231 degrees (see Figure 2).
We calculated the following dose distributions using PHITS.
(1) Absorbed dose of the affected chest region including thebreast area. (2) Absorbed dose in the contralateral chest
region including the breast area by the scattered radiation.
As described above, in breast-conserving therapy, the entire
conserved breast is irradiated approximately 50 Gy in
divided dose as prescribed dose. We note the present dose
calculation results in Figure 3 show the values before thenormalization to therapeutic prescription doses.

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Figure 2 Simulation condition of the irradiation to thebreast phantom.

JAEA-Conf 2020-001
2020-001

Figure 3 Absorbed dose distributions. (a) The affected chest area including breast area. The horizontal axis isdisplayed as x coordinate. In this case, the range of y-axis is from 0cm to 25cm. (b) the affected chest area including
breast area. The horizontal axis is displayed as y coordinate. The range of the x-axis is from -40cm to 0cm. (c) the
scattered radiation on the contralateral chest area including breast area. The horizontal axis is displayed as x
coordinate. The rage of y-axis is from 0cm to25cm, and (d) the scattered radiation on the contralateral chest area
including breast are. The horizontal axis is displayed as y coordinate. The range of the x-axis is from 0cm to 40cm.

3. Results
As seen in figure 3 (a) and (b), the absorbed dose distribution on the x-direction is larger than that for ydirection. Prescription dose shows an almost overall uniform dose distribution in the chest area. Figure 3 (c)and (d) show the absorbed dose for scattered radiation on the contralateral chest area. Figure 3 (c) shows thatthe effect of scattered radiation was large inside of the contralateral chest area. Numerical results are also
shown in Table 1 and 2.
In this evaluation, we just evaluated the dose distribution of the chest area, not breast itself. So, some low
absorption dose areas in the figure are considered as an absorption in the heart or/and lung region and so on.
Precise analysis of the dose distribution for breast region will be performed in the near future.

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2020-001

Table 1 the absorbed dose of the affected breast

Direction

Dose (Gy/source) 4MV

Dose (Gy/source)

R-L (x-direction)

8.6E-15

1.2E-14

P-A (y-direction)

5.1E-15

7.4E-15

6MV

Table 2 the absorbed dose of the contralateral breast4.

Direction

Dose (Gy/source) 4MV

Dose (Gy/source)

R-L (x-direction)

4.9E-19

1.2E-18

P-A (y-direction)

5.1E-19

6.1E-19

6MV

Summary
Absorbed dose for the scattered radiation on the contralateral evaluated chest region is about 4 orders ofmagnitude smaller than the absorbed dose in the affected chest region. During breast-conservation therapy,
the contralateral breast may be irradiated about 5 mGy. This is equivalent to 2 times of the mammography
dose, suggesting that treatment planning should be done carefully considering low-dose exposure to thecontralateral chest region.

Acknowledgement
We thank Dr. S. Hashimoto (JAEA) for his support of PHITS and Dr. K. Sutherland (Hokkaido Univ.
and Stanford Univ.) for checking the manuscript.

References[1] Japanese Society for Radiation Oncology, Radiation treatment planning guidelines 2016:170-176.
[2] C. Taylor, C. Correa, et al., Early Breast Cancer Trialists’ Collaborative Group. Estimating the risks of breast
cancer radiotherapy: evidence from modern radiation doses to the lungs and heart and from previous randomized
trials, J. Clin. Oncol. 2017;35(15):1641-9.
[3] T. Sato, Y. Iwamoto et al., Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02,
J. Nucl. Sci. Technol. 55, 684-690, 2018.
[4] General Incorporated Association Japan Breast Cancer Society, Guidelines for the Practice of Breast Cancer
Treatment 2018 Edition, 4th Edition: 308-315.

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JAEA-Conf 2020-001

42 Burnup Calculation with Different Fission Yield Data
Atsuya ISO, Yuya YAMASAKI, Taichi TAKEISHI, Satoshi TAKEDA, Takanori KITADA
Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871, Japane-mail : a-iso@ne.see.eng.osaka-u.ac.jp

The fission yield has recently been evaluated in Japan based on the various theories. In order toevaluate the effect of different fission yield data on burnup calculation, burnup chain and/or fission yielddata used in a burnup calculation is to be modified so as to obtain reasonable results because
different/additional isotopes are stored among the data, especially for isomers.
The comparison between JENDL/FPY-2011 and new fission yield evaluated by Tokyo Tech was
performed with SWAT4.0 for the PIE data: BM5 sample of Swiss Beznau-1 PWR, and the results showsthat there is relatively a large impact on the results. This shows the importance of how to handle theburnup chain and/or fission yield data on the comparison among different fission yield data.

1. Introduction
Fission yield is an important physical quantity to evaluate the amount of fission products after
irradiation. The fission yield used in Japanese Evaluated Nuclear Data Library (JENDL) employs
ENDF/B [1] data, but recently the new fission yield was evaluated by Tokyo Tech [2] and the fission yieldis planned to be adopted in the next version of JENDL. The validation of the data is inevitable before theadoption of new fission yield, and the validation is to be performed with different fission yield data
through the comparison of the burnup calculation results to the PIE data. However, there are someconcerning points in the comparison of the burnup calculation results with different fission yield, where
the nuclides stored in the data are different among different fission yield.
Different/additional isotopes are stored among the data, especially for isomers. Therefore, it isnecessary to properly modify burnup chain and/or fission yield data used in a burnup calculation in orderto obtain valid results.
This paper shows the effect of modification of the burnup chain and/or fission yield on burnup
calculation for the comparison between fission yields.

2. Modification of Burnup Chain and Fission Yield
The latest version of JENDL/FPY-2011 [3] (FPY-2011) and fission yields newly evaluated by Tokyo
Tech are discussed in this study. These fission yields were replaced on the cross-sectional library set of- 225 -

JAEA-Conf 2020-001

ORLIBJ40 [4]. Since ORLIBJ40, FPY-2011, and the data of Tokyo Tech are different in nuclides which
are given the fission yield as shown in Table 1, the burnup chain and/or fission yield in ORLIBJ40 need
to be modified.
Table 1 Difference in number of stored nuclides between FPY-2011 and fission yield evaluated by
Tokyo Tech

Number of stored nuclides

ORLIBJ40

FPY-2011

Tokyo Tech yield

1200

1260

1250

Table 1 shows that there are many differences in stored nuclides, and this fact requires the modificationin burnup chain and/or fission yield data. There are two cases to be considered in the modification.
Case-1 : the nuclide which is not used in burnup chain, but of which fission yield is given only for one or
two fission yield data,
Case-2 : the nuclide which is used in burnup chain, but of which fission yield is not given for one or two
fission yield data,
As for Case-1, there are several examples as shown in Figures 1 and 2. 151mCe is not used in burnup
chain of SWAT4.0 [5], but the fission yield is given in FPY-2011, and not given in both ORLIBJ40 andthe fission yield evaluated by Tokyo Tech. Here the fission yield of 151mCe was added to that of 151Ce
which is used in the burnup chain, with the consideration that the half-life of 151mCe (1.02[s]) is similar tothat of 151Ce (1.76[s]), and the daughter nuclide (151Pr) is the same to 151Ce. Another example is for 152m2Pr.
The fission yield of 152m2Pr is not used in burnup chain of SWAT4.0, but the fission yield is given only in
FPY-2011. 152Pr and 152m1Pr are given in the burnup chain, so the fission yield is added to that of 152Pr in
this study.

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JAEA-Conf 2020-001

Fig. 1 Case1-1: In burnup chain, ground state is given, but meta-stable state is not given
Figures show the decay chain around 151Ce before and after the modification of fission yield. Tables
show fission yields of 151Ce and 151mCe before and after modification. The fission yield of the metastable nuclides not used in burnup chain of SWAT4.0 is added to ground nuclide.

Fig. 2 Case1-2: In burnup chain, ground state and meta-stable state are given, but meta-stable2 state
is not given
Figures show the decay chain around 152Pm before and after the modification of fission yield. Tables
show fission yields of 152Pm, 152mPm and 152m2Pm before and after modification.

As for Case-2, there are also several examples as shown in Figures 3. 84mAs is used in the burnup
chain, but the fission yield is not given in FPY-2011 and new fission yield. In this case, we don’t haveany fission yield data for the nuclide, therefore the fission yield for the nuclide is set to be zero, although- 227 -

JAEA-Conf 2020-001

84

As is used in the burnup chain and the fission yield data is given for all fission yield data.

Fig. 3 Case2-1: In burnup chain, ground state is given, but meta-stable state is not given
Figures show the decay chain around 84As before and after the modification of fission yield. Tables
show fission yields of 84As and 84mAs before and after modification. Fission yields of 84mAs not given by
FPY-2011 and fission yield evaluated by Tokyo Tech are 0.

3. PIE analysis with different modification in burnup chain and fissionyield
PIE analysis is usually performed to check the validity of the data used in the calculation such as
fission yield data, and the modification described in the previous chapter brings the difference in the
results as easily expected. Thus the impact of the different modifications on the PIE analysis results isevaluated in this study.
PIE used in the calculation is performed in ARIANE program and the sample of BM5 [6] is the target
fuel to evaluate the nuclide composition after irradiation, as shown in Figure 4. The PIE data is selected
as the experimental results because of small uncertainty among the reported results.

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JAEA-Conf 2020-001

Fig. 4 Fuel assembly model and position of BM5 sample

PIE analysis with and without the modification expressed in Case 1 brings a non-negligible impact
on the PIE analysis results as shown in Figure 5. The results without the modification are evaluated in thecondition that the fission yield of 151mCe is neglected (no addition to the fission yield of 151Ce). Figure
5 shows that the different modification brings about 10% difference in C/E value of 151Sm, 152Sm, 151Eu,
which are the daughter nuclides of 151Cm as shown in Figure 6. These results show that a poor
modification brings the misunderstanding for the validity of the data used in the analysis, and proper
modification is necessary to check the validity of new fission yield data. To avoid poor modification, thefollowing is preferable that all nuclides are used in the burnup chain.

Fig. 5 Impact on the burnup calculation result
Green ellipses show where C/E values are drastically changed by the modification.

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JAEA-Conf 2020-001

Fig. 6 Example of burnup chain around Ce151(Dashed arrows show Ў-decay omitting some nuclides.)4. Conclusion
Recently the new fission yield was evaluated by Tokyo Tech and the fission yield is planned to be
adopted in the next version of JENDL. The validation of the data is inevitable before the adoption of new
fission yield, and the validation is usually performed with different data where the nuclides stored in eachdata are not the same. This fact brings the cases where the misunderstanding of the validity is observedby the poor modification. It is also found that the about 10% difference in the C/E value can be observed
for daughter nuclides through the PIE analysis with proper or poor modification. Proper modification of
burnup chain and fission yield are necessary to obtain reasonable results with the consideration of halflife and decay chain.

References[1] M.B. Chadwick, P. Obložinský, M. Herman, et al., "ENDF/B-VII.0: Next generation evaluatednuclear data library for nuclear science and technology",Nuclear Data Sheet, 107, p.2931, (2006).
[2] K. Tsubakihara et al., To be submitted to JNST; S. Chiba, private communication.
[3] J. Katakura, "JENDL FP Decay Data File 2011 and Fission Yields Data File 2011", JAEAData/Code 2011-025, Japan Atomic Energy Agency (2012).
[4] K. Okumura, K. Sugino, K. Kojima, T. Jin, T. Okamoto, and J. Katakura, "A Set of ORIGEN2
Cross Section Libraries Based on JENDL-4.0: ORLIBJ40", JAEA-Data/Code 2012-032, Japan
Atomic Energy Agency (2012) (in Japanese).
[5] T. Kashima, K Suyama, and T. Takada, "SWAT4.0 - The Integrated Burnup Code System Driving
Continuous Energy Monte Carlo Codes MVP, MCNP and Deterministic Calculation Code SRAC",
JAEA-Data/Code 2014-028, Japan Atomic Energy Agency (2015) (in Japanese).
[6] R.T.Primm III, "ARIANE International Programme Final Report", ORNL/SUB-97-XSV750-1, Oak
Ridge National Laboratory (2003).

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43

Improvement of fuel ion ratio diagnostics performanceusing anisotropic neutron emission spectrum
in NBI heating deuterium-tritium plasma
Tomoki Urakawa and Hideaki Matsuura

Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University,
744 Motooka Nishi-ku Fukuoka 819-0395, Japane-mail : urakawa-0804@nucl.kyushu-u.ac.jp

For the stable operation of the deuterium-tritium (DT) fusion reactor, it is important to know the triton todeuteron fuel-ion density ratio (nT/nD) during the operation. A method to obtain the nT/nD density ratio fromthe emission ratio of 14 MeV neutrons from T(d,n) 4He reactions and 2.45 MeV neutron from D(d,n)3He
reactions has been studied. This method has a serious problem that the detection of DD neutrons is interferedby slowing-down component of DT neutrons. As a solution to this problem, a technic increasing the fractionof the DD to DT reaction rate by NBI heating and utilizing modification of the DT and DD neutron emission
spectra by has been studied. Owing to this technic, the measurement has found to be possible even whennT/nD = 1.0. However, the detectable plasma condition is still limited, and further improvement is required.
In this study, we calculated the neutron spectrum with the detector by introducing a more realistic angle
dependence of neutron production, compared to the previous study. The results showed that the energy region
where DD neutrons can be more clearly measured was found and evaluated the effective in improving
measurement performance. And it was shown that the fuel-ion ratio diagnosis in ITER class plasma wassufficiently established with the improvement of measurement performance.

1. Introduction
Diagnostic to measure the fuel ion ratio in deuterium-tritium (DT) fusion plasma is important to stably
operate and control a fusion reactor. Requirement for fuel ion ratio diagnostics are over the parameter rangeof 0.01 < nT/nD < 10, spatial resolution of a/10, where ‘a’ is the minor radius of the plasma [1]. In the ITER,
plasma operation over the parameter range of 0.1 < nT/nD< 3.0 is assumed. Recently, various methods toknow the fuel-ion density ratio such as gamma-ray measurement [2], charge exchange recombination
spectrometry [3], collective Thomson scattering [4] and neutron measurement [5], have been proposed. The
devices for the above measurement will be installed in ITER for the practical examination of the diagnostic[6]. In fuel ion ratio diagnostic using neutron measurement, the intensity of each spectra of the 14 MeV
neutrons from D(t,n) 4He reactions and 2.45 MeV neutron from D(d,n)3He reactions are measured. Fuel ion
ratio is given by

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JAEA-Conf 2020-001

𝑅𝑅�� � �� ���
𝑛𝑛 �
�
,
𝑛𝑛�
2𝑅𝑅�� � �� ���

�1�

where RDD and RDT are each DD and DT reaction rates, nT and nD are the number density of triton and deuteron,and <σv>DD and <σv>DT are the reaction rate coefficients for DD and DT reaction. As a serious problem ofthis diagnostics, the detection of DD neutron (Signal) is interfered by slowing-down component of DT
neutrons (Noise). Due to the above problem, high measurement accuracy for both neutron production rates
is required to distinguish between the Signal and Noise. And, it is also important to understand the spectrumof the Noise using neutron transport calculation [7]. As a result, in the case of thermal plasma, it has beenreported that the detectable parameter range is nT/nD < 0.6 and Ti > 6 keV [1].
In order to improve the measurement performance, a technic using the risein the fraction of the DD to DT reaction rate and modification of the neutron
emission spectra by NBI-heating has been studied [8]. First, the degree ofenhancement for the DD reaction rate coefficient is much larger than that for
DT due to the cross section of Fig. 1 [9]. Thus, the ratio of DD signal to DT
noise (S/N value) is increased. In beam-injected plasma, the shape of neutron
emission spectra significantly changed. By considering the modification of
the neutron emission spectrum, i.e., shifting the measurement energy region

Fig. 1 Cross section for DD
and DT reactions as afunction of the energy in
the CM frame.

to the higher energy side, we can reduce the Noise to Signal ratio. In Ref [8],possibility of measurement at nT/nD = 1.0 was shown by using the method.
But, evaluation of the detectable parameter range is still insufficient by only
examining Ref [8]. And, further improvement is necessary for measurementof ITER class plasma.
In NBI heating plasma, emission spectrum of the particle which produced
fusion reaction has anisotropy in addition to modified from Gaussian [10].

Fig.2 Differential crosssection for DD reaction.

Due to anisotropic emission, the population of energetic particle is higher in the NBI incident direction. In
the case of D(d,n)3He reaction, anisotropy appears greatly due to differential cross section. Fig. 2 showsdifferential cross section of D(d,n)3He reaction [11]. Solid lines show incident deuteron energy (100, 200,
500, 800, 1000, 1500 and 2000 keV). The modification from Gaussian to the high energy side is even greater
since forward emission is dominant from Fig. 2. Thus, performance of the fuel ion diagnostic using modified
neutron emission spectrum measurement may be more improved due to considered anisotropic.
In this study, we introduce the more realistic angle dependence of neutron production by considering theanisotropy of the neutron emission spectra, and calculate the neutron spectra and S/N value with the detector.
And, we evaluate the energy region where DD neutrons can be more clearly measured and indicate thatconsideration of anisotropy is effective for improving measurement performance.
2. Calculation model
2-1 Double differential neutron emission spectra
In Fig. 3, the double differential emission spectrum for neutron as a function of neutron energy and

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JAEA-Conf 2020-001

emission angle relative to the beam-injection direction in the laboratory system is shown. This neutron
emission spectrum was assumed to be obtained from the ITER like plasma sustained by NBI heating. In Fig.
3, plasma parameters are assumed as follow; ne = 5.0×1019 m-3, Te = 10 keV, PNBI = 33 MW, ENBI = 1.0 MeV,
τE = 3.0 s and nT/nD = 1.0. Here ne is electron density, Te is electron temperature, PNBI is beam power, ENBI isbeam energy, τE is energy confinement time. In this study, we assumed tangential deuteron beam to thetoroidal magnetic field. Beam-particle slowing down is calculated using analytic model described in Ref.
[12]. The velocity distribution functions of the triton and deuteron (bulk component) are assumed to be
Maxwellian. We calculate this neutron emission spectrum using the method described in Ref. [10]. The
double differential neutron emission energy spectrum is written as
𝑑𝑑 � 𝑁𝑁�
𝑑𝑑𝑑𝑑
1
�𝐸𝐸, 𝜃𝜃��� � �� �� �|𝑣𝑣⃗� |������ ��𝑣𝑣⃗���� ��
𝑑𝑑𝑑𝑑𝑑𝑑𝑑𝑑���
𝑑𝑑𝑑𝑑
1 � 𝛿𝛿������ 𝛿𝛿�𝑑𝑑 � 𝑑𝑑� �𝛿𝛿�𝛺𝛺��� � 𝑑𝑑� �𝑣𝑣� 𝑑𝑑𝑑𝑑⃗� 𝑑𝑑𝑑𝑑⃗���� 𝑑𝑑𝑑𝑑 ,where En is the neutron energy in the laboratory system [13];𝑚𝑚 ������
1
𝐸𝐸� � 𝑚𝑚� 𝑉𝑉�� ���� 𝑑𝑑� �𝑚𝑚� � 𝑚𝑚 ������ ������
2
�𝑉𝑉� cos 𝜃𝜃� �

2𝑚𝑚 ������

𝑚𝑚� � 𝑚𝑚 ������

�������� � 𝑑𝑑� � ,

�2�

�3�

where mn(3He,α) is the neutron (3-helium, α-particle) mass, Vc is the center-of-mass velocity of the colliding

particles, θc is the angle between the center-of-mass velocity and the neutron velocity in the center-of-massframe, QDD(DT) is each reactions Q-value, and Er represents the relative energy given by𝐸𝐸� �

1 𝑚𝑚� 𝑚𝑚�����𝑣𝑣⃗ �𝑣𝑣⃗
� .
2 𝑚𝑚� �𝑚𝑚���� � ����

�4�

The θlab represents the angle between the direction of emitted
neutron and that of beam injection in the laboratory system, andΩn is a unit vector in the direction of emission of neutron in thelaboratory system, which is determined using the classicalkinematics as a function of 𝑣𝑣⃗� , 𝑣𝑣⃗���� , and 𝜃𝜃� . In Section 3.1,

we discuss using only the neutron emission spectrum shown in
Fig.1 as a neutron source. In Section 3.2, we calculate the neutron
source using analysis models in Ref. [9] and Ref. [12] for eachplasma parameter and do the following discussion.
2-2 Neutron transport calculation
2-2-1 Blanket with detector modelFig. 3 Double differential emission
spectrum produced by the D(d,n)3He
reactions as a function of neutronenergy and emission angle, θlab,relative to the beam-injection directionin the laboratory system. PNBI = 33
MW, ENBI = 1.0 MeV, Te = 10 keV, andne = 5.0x1019 m-3 are assumed.

Fig. 4 shows the computational schema of the blanket systems with collimator (detector). In Fig. 4, thecollimator direction, φ, was set to be the angle made with respect to the extended line in the NBI incident
direction when viewed from the blanket (upper view). The blanket model was employed as the torus form
which has a 6.2 m major radius and a 2.0 m minor radius. And, blanket material component and each layer

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JAEA-Conf 2020-001

thickness is shown in Table 1 [13]. The collimator is a pillar 10 cm ininner diameter and 20 cm in outer diameter. And the length of collimator
is 225 cm. These conditions were set based on Ref. [8] and Ref. [14].
2-2-2 Signal and Noise calculation
Incident DD neutron spectrum in the detector (Signal) andslowing-down component of DT neutrons (Noise) was calculatedusing the Monte Carlo transport code MVP [15] with theAnd, Signal and Noise are calculated using a torus volume

Fig. 4 The model used for MVP neutrontransport calculations assuming the ITER
condition. The φ shows the angle madewith respect to the extended line in the NBI
incident direction.

neutron source with uniform emissivity [17]. The validity of

Table 1 Blanket component.

JENDL-4.0 [16] nuclear data library. In this study, we adopted
point detector estimators in calculating Signal and Noise.

this calculation method is shown in Ref. [1]. Throughout thesimulation, the measurement energy range of neutron detection
set to Ec±120 keV (Ref. [5]), where Ec is center energy in themeasurement energy range.
3. Result and discussion
3-1 Incident neutron spectrum on detector surface
Fig. 5 shows incident neutron spectrum with the detector. (φ =0°, 25°, 50°). And, the incident spectrum obtained under theassumption of isotropic emission (φ = 0°, 50°) and the slowing
down component of DT neutrons (φ = 0°) is also shown. Since theemission spectrum around the NBI incident direction is measured,
the ratio of energetic neutron increases as φ increase. In particular,when φ = 0°, the peak appears in the high energy region (around 4.0
MeV). This peak is as large as Gaussian formed by thermonuclear
reaction. Fig. 6 shows improvement of count due to anisotropic.
Now, count is the integral value in the range of Ec±120 keV.
Depending on the collimator direction, φ, the measurement energy

Fig. 5 Incident neutron spectrumwith the detector. φ shows theangle made with respect to theextended line in the NBI incident
direction. PNBI = 33 MW, ENBI = 1.0
MeV, Te = 10 keV, and ne = 5.0x1019
m-3, nt/nd = 1.0 are assumed.

region where count is improved differs. In the case of φ = 50°, countis improved about 50% in the 2.4-2.5 MeV region. On the other hand,in the case of φ = 50°, count is improved about 8.0 times in the highenergy region (around 4.0 MeV). And, the intensity of the incident
spectrum obtained under the assumption of isotropic emission isdifferent between φ = 0° and φ = 50° due to the effect of line of sight.
It can be confirmed that the above results are appropriate from theinfluence of the collimator direction on the intensity of neutron flux
and the S/N value discussed in Ref. [7].

229 - 234

Fig. 6 Improvement of countdue to anisotropic. ‘S’ is theintegral value in the range of
Ec±120 keV.

JAEA-Conf 2020-001

3-2 Improvement of S/N value and distinction of Signal and Noise
Fig. 7 shows S/N value obtained for each incident spectra of Fig.5. Now S/N value is given by������ ���𝐹𝐹 �𝐸𝐸�𝑑𝑑𝑑𝑑
�𝑆𝑆
��� ��� ������
,������
���
𝑁𝑁𝐹𝐹�� �𝐸𝐸�𝑑𝑑𝑑𝑑
������� ���

where FDD(DT) is incident DD (DT) neutron energy spectrum on the�5�

detector. In Ref. [17], DD neutron and Slowing down componentof DT neutron can be distinguished when S/N > 0.4. But, in thiscase, high accuracy is required for both of Signal and Noise
measurements to decide each spectrum. On the other hands, in Ref.
[5], it has been pointed out that DD neutrons can be confirmed
sufficiency when S/N ≫ 1.0. Thus, for the above reasons, wedefined S/N > 1.0 as 'sufficient condition' and S/N > 0.4 as'detectable condition' in fuel ion ratio diagnostic.

Fig. 7 S/N value dependency on
measurement energy range.

In the case of measuring neutron emission spectrum Fig. 3 (Ti =10 keV, nt/nd = 1.0), S/N > 0.4 can be achieved regardless of thecollimator direction, φ. S/N > 1.0 can be achieved regardless of theanisotropic consideration when φ = 0°. In addition, S/N > 2.0 can beachieved considering anisotropic (φ = 0°) and measuring high energy
region (around 4.0 MeV). Thus, the measurement of the anisotropic
emission spectrum showed that the S/N was improved about twice.
It was shown that ‘sufficient condition’ is achieved in the ITER
class plasma.
3-3 Detectable plasma parameter range
Fig. 8 shows S/N values at various ion temperature and fuel ion

Fig. 8 S/N value at various plasma
parameter. Solid lines indicate S/N =0.8, 1.0, 1.2, 1.6 and 2.0 respectively.
Red diagonal lines indicate the rangeof plasma parameters that satisfy'sufficient condition (S/N > 1.0)'.

ratio for anisotropic emission spectrum measurement (Ec = 3780
keV). The diagonal line in Fig. 8 indicates the region where S/N >1.0. Due to anisotropic consideration, the parameter range of‘sufficient condition’ is nT/nD < 1.2 and Ti < 20 keV. In addition, it
was shown to achieve detectable condition (S/N > 0.4) even whennT/nD ≒ 3.0. This suggests that the fuel ion ratio diagnosis using
neutron diagnostic is valid over the entire ITER operating range.
Fig. 9 shows the detectable plasma parameter range under‘sufficient condition’. The solid line shows S/N value (= 1.0) at
various plasma parameters for anisotropic emission spectrum
measurement. The dashed and dotted lines indicate that obtained
when measuring the isotropic emission spectrum and the Gaussian

230 - 235

Fig. 9 S/N value (= 1.0) obtained bymeasuring each spectrum. Bluediagonal lines indicate the plasma
parameter region that can be measured
by considering the anisotropy of thespectra.

JAEA-Conf 2020-001

distribution, respectively. Due to anisotropic consideration, parameter range to satisfy the condition is
expanded. In particular, the S/N value is well improved in the high nT/nD region. Focusing on the modificationfrom Gaussian reduces the effect of deuteron density reduction on neutron emission rate. These result showsthat consideration of anisotropy is effective.
4. Conclusion remarks
In this study, we evaluated the improvement of nT/nD density ratio measurement by anisotropic emission
spectra. When the installation direction of the collimator was set to φ = 0 ° (tangential line), it was found that
there was a measurement region where the number of incident neutrons increased by a maximum of about 8times compared to the assumption of isotropic emission. As a result, the S/N value was doubled compared tothe isotropic emission spectrum measurement in the case of Ti = 10 keV, nT/nD = 1.0. And, due to anisotropic
neutron emission spectrum measurement, the parameter range of ‘sufficient condition’ (S/N > 1.0) is nT/nD <1.25 and Ti < 20 keV. In addition, it was shown to achieve ‘detectable condition’ (S/N > 0.4) even when nT/nD≒ 3.0. Therefore, measurement of anisotropic emission spectrum is effective for improving performance offuel ion ratio diagnosis.
In future work, we need to discuss the plasma parameter range that can be measured in more detailconsidering the performance of the detector (response function, detection efficiency, etc.). And, discussionof electron density dependence is also needed. In addition, at high nT/nD, precise calculation of DT slowingdown component by neutron transport calculation is required because of 0.4 < S/N <1.0. Thus, it is also
important to improve the accuracy of nuclear data when considering the feasibility of fuel ion ratio diagnostic.
And, it is also necessary to study the practicality of this method in the nuclear fusion DEMO reactor.
References[1] G. Ericsson, et al., Rev. Sci. Instrum. 81, 10D324 (2010).
[2] V. G. Kiptily, Nucl. Fusion, 55, 023008 (2015).
[3] N. C. Luhmann, et al., Fusion Sci. Technol. 53, 335-96 (2008).
[4] L. Bertalot, et al., JINST, 7, C04012 (2012).
[5] J. Källne, et al., Rev. Sci. Instrum. 62, 2781 (1991).
[6] S. B. Korsholm, et al., Rev. Sci. Instrum. 81, 10D323 (2010).
[7] P. Antozzi, et al., Nucl. Instrum. Methods Phys. Res. Sect. A, 368, 457-472 (1996).
[8] Y. Kwamoto, et al., J. Plasma Fusion Res. 11, 2405078 (2016).
[9] H. S. Bosch and G. M. Hale, Nucl. Fusion, 32, 611-631 (1992).
[10] H. Matsuura, et al., J. Plasma Fusion Res. 9, 48-53 (2010).
[11] M. Drosg and O. Schwerer, Production of monoenergetic neutrons between 0.1 and 23 MeV: neutron energies
and cross-sections, Handbook of Nuclear Activation Data (Vienna: IAEA) STI/DOC/10/273, ISBN 92-0135087-2 (1987).
[12] John D. Gaffey, Jr., J. Plasma Phys. 16, 149-169 (1976).
[13] M. E. Swan, et al., Proceedings of the 22nd IEEE / NPSSS symposium on Fusion Engineering,
Albuquerque, HM. June 17-21, 1 (2007).
[14] K. Okada, et al., 32nd EPS Conference on Plasma Phys. Tarragona, 27 June - 1 July 2005 ECA
Vol.29C, P-4.096 (2005).
[15] Y. Nagaya, et al., JAERI 1348 (2005).
[16] K. Shibata, et al., J. Nucl. Sci. Technol. 48, 1–30 (2011).
[17] J. Källne, et al., Rev. Sci. Instrum. 68, 581 (1997).

231 - 236

国際単位系（SI）
表１．SI 基本単位
SI 基本単位

基本量

名称
記号
長
さメ ートル m
質
量 キログラム kg
時
間
秒
s
電
流ア ンペア A
熱力学温度 ケ ル ビ ン K
物 質 量モル mol
光度 カ ン デ ラ cd

表２．基本単位を用いて表されるSI組立単位の例
SI 組立単位
組立量
名称
記号
面
積 平方メートルm2
体
積 立方メートルm3
速 さ ， 速 度 メートル毎秒m/s
加速度 メートル毎秒毎秒
m/s2
波
数 毎メートルm-1
密 度 ， 質 量 密 度 キログラム毎立方メートル
kg/m3
面
積
密
度 キログラム毎平方メートル
kg/m2
比
体
積 立方メートル毎キログラム
電流
密
度 アンペア毎平方メートル
磁 界 の 強 さ アンペア毎メートル
量 濃 度 (a) ， 濃 度 モル毎立方メートル
質量
濃度 キログラム毎立方メートル輝
度 カンデラ毎平方メートル屈折率 (b) （数字の） １
比 透 磁 率 (b) （数字の） １

乗数
24

10
1021
1018
1015
1012
109
106
103

3

m /kg
A/m2
A/m
mol/m3
kg/m3
cd/m2
1
1

102
101

ヨ

表５．SI 接頭語
記号
乗数
名称
タ
Ｙ
シ
10-1 デ

ゼ
タ
エ ク サ

Ｚ
Ｅ

10-2

ペ
テ

タ
ラ

Ｐ
Ｔ

ギ
メ

ガ
ガ

Ｇ
Ｍ

マイクロ
ノ
10-9 ナコ
10-12 ピ
10-15 フェムト

キ
ロ
ヘ ク ト
デ
カ

ｋ
ｈ
da

10-3
10-6

記号

セ ン チ
ミ
リ

ト
10-18 ア
10-21 ゼ プ ト
10-24 ヨ ク ト

d

c
m
µ
n
p
f
a
z
y

表６．SIに属さないが、SIと併用される単位
名称
記号
SI 単位による値
分
min 1 min=60 s
時
h 1 h =60 min=3600 s（a）量濃度（amount concentration）は臨床化学の分野では物質濃度（substance concentration）ともよばれる。
（b）これらは無次元量あるいは次元１をもつ量であるが、そのことを表す単位記号である数字の１は通常は表記しない。

d
°
’

日
度
分

表３．固有の名称と記号で表されるSI組立単位
SI 組立単位
組立量
他のSI単位による SI基本単位による
名称
記号
表し方
表し方
（ｂ）
平
面
角 ラジアン(ｂ)
rad
m/m
1
（ｂ）
(ｂ)
(c)
2
2
立
体
角 ステラジアン
sr
1
m /m
周
波
数 ヘルツ（ｄ）
Hz
s-1
力
ニュートン
N
m kg s-2
圧力
応
力 パスカル,Pa
N/m2
m-1 kg s-2
エ ネ ル ギ ー , 仕 事 , 熱 量 ジュールJNm
m2 kg s-2
仕 事 率 ， 工 率 ， 放 射 束 ワット
WJ/s
m2 kg s-3
電
荷
電気量 クーロン,
CsA
電 位 差 （ 電 圧 ） , 起 電 力 ボルト
VW/A
m2 kg s-3 A-1
静電
容量 ファラドF
C/V
m-2 kg-1 s4 A2
電
気
抵
抗 オーム
Ω
V/A
m2 kg s-3 A-2
コ ン ダ ク タ ン ス ジーメンス
S
A/V
m-2 kg-1 s3 A2
磁
束 ウエーバWb
Vs
m2 kg s-2 A-1
磁
束
密
度 テスラ
T
Wb/m2
kg s-2 A-1
イ ン ダ ク タ ン ス ヘンリーH
Wb/A
m2 kg s-2 A-2
セ ル シ ウ ス 温 度 セルシウス度(ｅ)℃K
光
束 ルーメンlm
cd
cd sr(c)照度 ルクスlx
m-2 cdlm/m2
Bqs-1
放 射 性 核 種 の 放 射 能 （ ｆ ） ベクレル（ｄ）
吸収線量, 比エネルギー分与,
グレイ
Gy
J/kg
m2 s-2カーマ
線量当量, 周辺線量当量,Sv
J/kg
m2 s-2シーベルト（ｇ）
方向性線量当量, 個人線量当量酸素
活性 カタールkat
s-1 mol

1 d=24 h=86 400 s
1°=(π/180) rad1’=(1/60)°=(π/10 800) rad”1”=(1/60)’=(π/648 000) radha 1 ha=1 hm 2=104m2
L，l 1 L=1 l=1 dm3=103cm3=10-3m3t
1 t=103 kg

秒
ヘクタール
リットル
トン

表７．SIに属さないが、SIと併用される単位で、SI単位で表される数値が実験的に得られるもの
名称
記号
SI 単位で表される数値
電 子 ボ ル ト
ダ ル ト ン
統一原子質量単位

eV
Da
u

天

ua

文

単

位

1 eV=1.602 176 53(14)×10 -19J

1 Da=1.660 538 86(28)×10-27kg
1 u=1 Da

1 ua=1.495 978 706 91(6)×1011m

表８．SIに属さないが、SIと併用されるその他の単位
名称
記号
SI 単位で表される数値
バー
ル bar １bar=0.1MPa=100 kPa=10 5Pa
水銀柱ミリメートル mmHg １mmHg≈133.322Pa(a)SI接頭語は固有の名称と記号を持つ組立単位と組み合わせても使用できる。しかし接頭語を付した単位はもはや
コヒーレントではない。
(b)ラジアンとステラジアンは数字の１に対する単位の特別な名称で、量についての情報をつたえるために使われる。
実際には、使用する時には記号rad及びsrが用いられるが、習慣として組立単位としての記号である数字の１は明示されない。
(c)測光学ではステラジアンという名称と記号srを単位の表し方の中に、そのまま維持している。
(d)ヘルツは周期現象についてのみ、ベクレルは放射性核種の統計的過程についてのみ使用される。
(e)セルシウス度はケルビンの特別な名称で、セルシウス温度を表すために使用される。セルシウス度とケルビンの単位の大きさは同一である。したがって、温度差や温度間隔を表す数値はどちらの単位で表しても同じである。
(f)放射性核種の放射能（activity referred to a radionuclide）は、しばしば誤った用語で”radioactivity”と記される。
(g)単位シーベルト（PV,2002,70,205）についてはCIPM勧告2（CI-2002）を参照。

表４．単位の中に固有の名称と記号を含むSI組立単位の例
SI 組立単位
組立量
SI 基本単位による
名称
記号
表し方
-1
粘
度 パスカル秒
Pa s
m kg s-1
力 の モ ー メ ン ト ニュートンメートル
Nm
m2 kg s-2
表面
張力 ニュートン毎メートル
N/m
kg s-2角
速度 ラジアン毎秒
rad/s
m m-1 s-1=s-1
角
加速度 ラジアン毎秒毎秒
rad/s2
m m-1 s-2=s-2
熱 流 密 度 , 放 射 照 度 ワット毎平方メートル
W/m2
kg s-3
熱 容 量 , エ ン ト ロ ピ ー ジュール毎ケルビン
J/K
m2 kg s-2 K-1
比 熱 容 量 ， 比 エ ン ト ロ ピ ー ジュール毎キログラム毎ケルビン J/(kg K)
m2 s-2 K-1
比 エ ネ ルギ ー ジュール毎キログラム
J/kg
m2 s-2
熱伝
導率 ワット毎メートル毎ケルビン W/(m K) m kg s-3 K-1
体 積 エ ネ ル ギ ー ジュール毎立方メートル J/m3
m-1 kg s-2
電界の強さ ボルト毎メートル
V/m
m kg s-3 A-1
電荷密
度 クーロン毎立方メートル C/m3
m-3 s A
表面
電
荷 クーロン毎平方メートル C/m2
m-2 s A電 束 密 度 ， 電 気 変 位 クーロン毎平方メートル C/m2
m-2 s A誘
電率 ファラド毎メートル
F/m
m-3 kg-1 s4 A2
透磁率 ヘンリー毎メートル
H/m
m kg s-2 A-2
モ ル エ ネ ル ギ ー ジュール毎モル
J/mol
m2 kg s-2 mol-1
モルエントロピー, モル熱容量 ジュール毎モル毎ケルビン J/(mol K) m2 kg s-2 K-1 mol-1
照 射 線 量 （ Ｘ 線 及 び γ 線 ） クーロン毎キログラム
C/kg
kg-1 s A
吸収線
量率 グレイ毎秒
Gy/s
m2 s-3放射
強度 ワット毎ステラジアン
W/srm4 m-2 kg s-3=m2 kg s-3
放射輝
度 ワット毎平方メートル毎ステラジアン W/(m2 sr) m2 m-2 kg s-3=kg s-3
酵 素 活 性濃 度 カタール毎立方メートル kat/m3
m-3 s-1 mol

名称

オングストローム
海
里
バ
ー
ン

Å
Ｍ

１Å=0.1nm=100pm=10-10m
１M=1852m

b

ノ
ネ
ベ

ト
パ
ル

kn
Np
Ｂ

１b=100fm2=(10-12cm) 2 =10-28m2
１kn=(1852/3600)m/s

ル

dB

ッ
ー

デ

シ

ベ

SI単位との数値的な関係は、対数量の定義に依存。

エ

表９．固有の名称をもつCGS組立単位
名称
記号
SI 単位で表される数値
ル
グ erg 1 erg=10-7 J

ダ
ポ

イ
ア

ス
ス

ト ー ク
チ
ル

フ
ガ

ォ

ン dyn 1 dyn=10-5N
ズ P 1 P=1 dyn s cm-2=0.1Pa s
ス St 1 St =1cm2 s-1=10-4m2 s-1
ブ sb 1 sb =1cd cm-2=104cd m-2
ト ph 1 ph=1cd sr cm-2 =10 4lx
ル Gal 1 Gal =1cm s-2=10-2ms-2

マ ク ス ウ エ ルガ
ウス
エルステッド（ ａ）

Mx
G
Oe

1 Mx = 1G cm2=10-8Wb
1 G =1Mx cm-2 =10-4T
1 Oe (103/4π)A m-1（a）３元系のCGS単位系とSIでは直接比較できないため、等号「は対応関係を示すものである。

キ
レ
ラ

名称
ュ
リ
ン

レ
ガ

ト

表10．SIに属さないその他の単位の例
記号
SI 単位で表される数値
ー Ci 1 Ci=3.7×1010Bq

ゲ

ン

ン R
ド rad
ム rem
マ γ

フ
ェ
ル
ミ
メートル系カラット

1 R = 2.58×10-4C/kg

1 rad=1cGy=10-2Gy
1 rem=1 cSv=10-2Sv
1 γ=1 nT=10-9T
1 フェルミ=1 fm=10-15m

ト
標

準

気

1 メートル系カラット = 0.2 g = 2×10-4kg
ル Torr 1 Torr = (101 325/760) Pa
圧 atm 1 atm = 101 325 Pa

カ

ロ

リ

ー

cal

ミ

ク

ロ

ン

µ

大

」

1 cal=4.1858J（｢15℃｣カロリー），4.1868J（｢IT｣カロリー），4.184J（｢熱化学｣カロリー）

1 µ =1µm=10-6m（第8版，2006年）