Measurement of defect-induced electrical resistivity change of tungsten wire at cryogenic temperature using high-energy proton irradiation

Iwamoto, Yosuke; Yoshida, Makoto*; Matsuda, Hiroki; Meigo, Shinichiro; Satoh, Daiki; Yashima, Hiroshi*; Yabuuchi, Atsushi*; Kinomura, Atsushi*; Shima, Tatsushi*

JPS Conference Proceedings (Internet), 28, p.061003_1 - 061003_5, 2020/02

https://doi.org/10.7566/JPSCP.28.061003

To predict the lifetime of target materials in high-energy radiation environments at spallation neutron sources, radiation transport codes such as PHITS are used to calculate the displacements per atom (DPA) value. In this work, to validate calculated DPA values of tungsten, we implemented 0.25-mm-diameter wire sample of tungsten in a proton irradiation device with a Gifford-McMahon cryocooler and measured the defect-induced electrical resistivity changes related to the displacement cross section using 389-MeV protons at 10 K. In comparison with experimental data under 1.1 and 1.9 GeV proton irradiation, we found that damage rate of tungsten increases with proton energy due to increase the number of secondary particle s produced by nuclear reactions.

Development of active neutron NDA system

Toh, Yosuke

JAEA-Conf 2019-001, p.47 - 52, 2019/11

no abstracts in English

Simulation study on the design of nondestructive measurement system using fast neutron direct interrogation method to nuclear materials in fuel debris

Maeda, Makoto; Furutaka, Kazuyoshi; Kureta, Masatoshi; Ozu, Akira; Komeda, Masao; Toh, Yosuke

Journal of Nuclear Science and Technology, 56(7), p.617 - 628, 2019/07

https://doi.org/10.1080/00223131.2019.1611500

Measurement of neutron-production double-differential cross sections of C, Al, Fe, and Pb by 20, 34, 48, 63, and 78 MeV protons in the most-forward direction

Satoh, Daiki; Iwamoto, Yosuke; Ogawa, Tatsuhiko

Nuclear Instruments and Methods in Physics Research A, 920, p.22 - 36, 2019/03

https://doi.org/10.1016/j.nima.2018.12.049

The Particle and Heavy Ion Transport code System (PHITS) is a general purpose particle transport simulation code developed by the Japan Atomic Energy Agency (JAEA). The PHITS is utilized in various areas including a shielding design of accelerator facilities. Unfortunately, it is known that theoretical models and evaluated nuclear data used in the PHITS cannot reproduce the neutron production in most-forward direction for proton incidences. Hence, the present study aimed to obtain the experimental data of neutron-production double-differential cross sections of C, Al, Fe, and Pb by 20, 34, 48, 63, and 78 MeV protons in most-forward direction for improvement of theoretical models and nuclear data. The experiment has been performed at the ion irradiation facility (TIARA) of the National Institutes for Quantum and Radiological Science and Technology. The proton beams provided by the cyclotron were incident to the target sample. The neutrons produced by nuclear reactions were pass through the collimator in the most-forward direction, and measured with an organic scintillator at the experimental room. The kinetic energy of those neutrons was determined by the time-of-flight method. The obtained results were compared with the results of the theoretical model INCL and the nuclear-data library JENDL-4.0/HE used in the PHITS. It was found that the INCL and JENDL-4.0/HE cannot reproduce the peak structures observed for light nuclei, because they do not consider the nuclear transition between discrete states of nucleus. In addition, the JENDL-4.0/HE agreed with the experimental data of energy-integrated cross section within a factor of 2, but the INCL gave approximately 6 times larger values.

Development of differential die-away technique in an integrated active neutron NDA system for nuclear non-proliferation and nuclear security

Ozu, Akira; Maeda, Makoto; Komeda, Masao; Toh, Yosuke; Koizumi, Mitsuo; Seya, Michio

Proceedings of 2017 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC 2017) (Internet), 4 Pages, 2018/11

https://doi.org/10.1109/NSSMIC.2017.8532755

Experimental validation of the brightness distribution on the surfaces of coupled and decoupled moderators composed of 99.8% parahydrogen at the J-PARC pulsed spallation neutron source

Harada, Masahide; Teshigawara, Makoto; Oi, Motoki; Klinkby, E.*; Zanini, L.*; Batkov, K.*; Oikawa, Kenichi; Toh, Yosuke; Kimura, Atsushi; Ikeda, Yujiro

Nuclear Instruments and Methods in Physics Research A, 903, p.38 - 45, 2018/09

https://doi.org/10.1016/j.nima.2018.06.011

Study on neutron beam pulse width dependence in the nuclear fuel measurement by the neutron resonance transmission analysis

Kitatani, Fumito; Tsuchiya, Harufumi; Toh, Yosuke; Hori, Junichi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Nakajima, Ken*

KURRI Progress Report 2017, P. 99, 2018/08

https://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2017/CO2.pdf

Neutron resonance transmission analysis for measurement of nuclear materials in nuclear fuel

Tsuchiya, Harufumi; Kitatani, Fumito; Toh, Yosuke; Paradela, C.*; Heyse, J.*; Kopecky, S.*; Schillebeeckx, P.*

Proceedings of INMM 59th Annual Meeting (Internet), 6 Pages, 2018/07

Production and detection of fission-induced neutrons following fast neutron direct interrogation to various dry materials containing U

Sakoda, Akihiro; Nakatsuka, Yoshiaki; Ishimori, Yuu; Nakashima, Shinichi; Komeda, Masao; Ozu, Akira; Toh, Yosuke

Journal of Nuclear Science and Technology, 55(6), p.605 - 613, 2018/06

https://doi.org/10.1080/00223131.2017.1417175

For better nuclear material accountancy, we had developed a non-destructive assay system dedicated to uranium waste drums (JAWAS-N: JAEA Waste Assay System at Ningyo-toge). The system is based on a fast neutron direct interrogation (FNDI) method. To clarify the characteristics of the FNDI method and the performance of JAWAS-N, experimental and computational mock-up tests were carried out using various dry materials that contained known amounts of natural uranium. As a result, linearity between the die-away time () and the counts of fast neutrons attributed to U fission was confirmed experimentally and analytically. Moreover, the MCNP simulation was performed to discuss the radial and axial dependences of U fission probability, neutron detection efficiency, and sensitivity on uranium distributions in the drum. The simulation results agreed with the empirical results reported in a previous paper, providing valuable information on the practice of FNDI-based uranium determination. Furthermore, the nominal detection limits of natural uranium in JAWAS-N were estimated to be 15, 4, and 2 g for = 0.2, 0.3, and 0.4 msec, respectively. The findings obtained here will contribute to the implementation of the FNDI method to assess the quantities of U in actual uranium waste drums.

Development of neutron resonance transmission analysis as a non-destructive assay technique for nuclear nonproliferation

Tsuchiya, Harufumi; Kitatani, Fumito; Maeda, Makoto; Toh, Yosuke; Kureta, Masatoshi

Plasma and Fusion Research (Internet), 13(Sp.1), p.2406004_1 - 2406004_4, 2018/02

http://www.jspf.or.jp/PFR/PDF2018/pfr2018_13-2406004

Recently, it has become important in the field of nuclear nonproliferation and nuclear security to quantify nuclear materials (NMs) of uranium and plutonium in nuclear fuel using a non-destructive assay (NDA) technique. Currently, there is no reliable NDA system to apply to nuclear fuels such as spent fuel, fuel debris and next generation fuel for nuclear transmutation. Accordingly, development of NDA techniques for quantification of NMs in those fuels is an urgent issue. Neutron resonance transmission analysis (NRTA) is one candidate that is applicable to the quantification of NMs. Utilizing pulsed neutron beams, NRTA analyzes the content of a sample by measuring neutron beams that are transmitted from the sample. It is one of the reliable NDA methods that are based on a neutron time-of-flight technique for accurately evaluating nuclear data such as total cross sections and resonance parameters. A present NRTA system generally requires a large electron linear accelerator to produce intense neutron beams. Therefore this is not so easy to apply to various facilities that are used to measure NMs. Given this situation, a compact NRTA system would be required for practical applications of a method to quantify NMs in various samples. In order to realize a compact NRTA system, we consider two types of system: one uses a D-T neutron generator with pulse width of 10 sec and the other a small electron linac with pulse width of 1 sec Assuming each system is applied to measurements of NMs in spent fuel, numerical calculations were carried out and the results showed that the pulse widths of neutron beam largely affect the NRTA measurements. In this presentation, we will talk about the NRTA technique and give a schematic design of a compact NRTA system. Then, comparing calculation results for a D-T tube with those for a small electron linac, we especially discuss how the pulse widths of neutron beams to be used for NRTA affect the measurement of NMs in nuclear fuel.