Neutron-production double-differential cross sections of Pb and Bi in proton-induced reactions near 100 MeV

Iwamoto, Hiroki; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Ishi, Yoshihiro*; Uesugi, Tomonori*; Yashima, Hiroshi*; Nishio, Katsuhisa; Sugihara, Kenta*; elik, Y.*; et al.

Nuclear Instruments and Methods in Physics Research B, 544, p.165107_1 - 165107_15, 2023/11

https://doi.org/10.1016/j.nimb.2023.165107

The lack of double-differential cross-section (DDX) data for neutron production below the incident proton energy of 200 MeV hinders the validation of spallation models in technical applications, such as research and development of accelerator-driven systems (ADSs). The present study aims to obtain experimental DDX data for ADS spallation target materials in this energy region and identify issues related to the spallation models by comparing them with the analytical predictions. The DDXs for the () reactions of Pb and Bi in the 100-MeV region were measured over an angular range of 30 to 150 using the time-of-flight method. The measurements were conducted at Kyoto University utilizing the FFAG accelerator. The DDXs obtained were compared with calculation results from Monte Carlo-based spallation models and the evaluated nuclear data library, JENDL-5. Comparison between the measured DDX and analytical values based on the spallation models and evaluated nuclear data library indicated that, in general, the CEM03.03 model demonstrated the closest match to the experimental values. Additionally, the comparison highlighted several issues that need to be addressed in order to improve the reproducibility of the proton-induced neutron-production DDX in the 100 MeV region by these spallation models and evaluated nuclear data library.

Neutron capture cross-section measurements with TC-Pn in KUR for some nuclides targeted for decommissioning

Nakamura, Shoji; Endo, Shunsuke; Kimura, Atsushi; Shibahara, Yuji*

KURNS Progress Report 2022, P. 73, 2023/07

The present study is concerned with the neutron capture cross-sections that contribute to the evaluation of the amount of radionuclides possessing problems in decommissioning. In this study, Sc, Cu, Zn, Ag, In and W were selected among the objective nuclides, and their thermal-neutron capture cross-sections were measured using TC-Pn equipment of the KUR of the Institute for Integrated Radiation and Nuclear Science, Kyoto University. High purity metal samples were prepared. A gold-aluminum ally wire, cobalt and molybdenum foils were used to monitor the neutron flux at the irradiation position of TC-Pn. The flux monitors and metal samples were irradiated for 1 hour at 1-MW operation of the KUR. After irradiation, the irradiation capsule was opened, samples and flux monitors were enclosed in a vinyl bag one by one, and then rays emitted from the samples and monitors were measured with a high-purity Ge detector. The thermal-neutron flux component was derived with the reaction rates of flux monitors (Au, Co and Mo) on the basis of Westcott's convention, and found to be (5.920.10)10 n/cm/sec at the irradiation position. The measured reaction rate for each metal sample divided by the evaluated thermal-neutron capture cross-section should give the same value of the thermal-neutron flux component if the cross section is suitable. This time, we found that the cross sections of Sc and Zn were consistent with the evaluated one, but those of other nuclides were inconsistent with their evaluated ones; that is, it turned out that their thermal-neutron capture cross-sections should be modified.

ACE-FRENDY-CBZ; A New neutronics analysis sequence using multi-group neutron transport calculations

Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi

Journal of Nuclear Science and Technology, 60(2), p.132 - 139, 2023/02

https://doi.org/10.1080/00223131.2022.2087783

A new multi-group neutronics analysis sequence ACE-FRENDY-CBZ is proposed. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross section data of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement with the reference Monte Carlo results was obtained within 30 pcm differences in the bare systems and the thorium-reflected system, and approximately 100 pcm differences in the uranium-reflected systems. The use of the current-weighted total cross sections in the multi-group neutron transport calculations had non-negligible impacts over 100 pcm on k-eff, and the calculations with the current-weighted total cross sections systematically underestimated k-eff in the uranium-reflected systems.

Outline of JENDL-5

Iwamoto, Osamu

JAEA-Conf 2022-001, p.21 - 26, 2022/11

https://doi.org/10.11484/jaea-conf-2022-001

Development of nuclear data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

Nuclear data processing is an important interface between an evaluated nuclear data library and nuclear transport calculation codes. JAEA has developed a new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates ACE files which are used for the continuous-energy Monte Carlo codes including PHITS, Serpent, and MCNP; it was released as an open-source software under the 2-clause BSD license in 2019. After FRENDY version 1 was released, many functions are developed: the multi-group neutron cross-section library generation, the statistical uncertainty quantification for the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, etc. We released FRENDY version 2 including these functions. This presentation explains the overview of FRENDY and features of the new functions implemented in FRENDY version 2.

G-HyND: A Hybrid nuclear data estimator with Gaussian processes

Iwamoto, Hiroki; Iwamoto, Osamu; Kunieda, Satoshi

Journal of Nuclear Science and Technology, 59(3), p.334 - 344, 2022/03

https://doi.org/10.1080/00223131.2021.1971120

A hybrid nuclear data estimator (G-HyND) based on a machine learning technique with Gaussian processes (GP) was developed. G-HyND estimates cross-sections from a hybrid training dataset composed of an experimental dataset and an analytical dataset based on a nuclear physics model, and generates the cross-section datasets including the dataset's uncertainty information. It was demonstrated that an experimental dataset and a physics model-based analytical dataset perform a complementary role in nuclear data generation, and that the generated nuclear data from the hybrid training dataset are more reasonable than only those from the experimental dataset. Furthermore, solutions for two inherent GP problems, i.e., overfitting and computational cost, are presented within the G-HyND framework.

Integral experiments of technetium-99 using fast-neutron source reactor "YAYOI"

Nakamura, Shoji; Hatsukawa, Yuichi*; Kimura, Atsushi; Toh, Yosuke; Harada, Hideo

Journal of Nuclear Science and Technology, 58(12), p.1318 - 1329, 2021/12

https://doi.org/10.1080/00223131.2021.1942278

The present study performed fast-neutron capture cross-section measurement of Tc by an activation method using a fast-neutron source reactor "YAYOI" of the University of Tokyo. Technetium-99 samples were irradiated with reactor neutrons using a pneumatic system. Reaction rates of Tc were obtained by measuring decay gamma rays emitted from Tc. The neutron flux at an irradiation position was monitored with gold foils. The fast-neutron capture cross section of Tc at neutron energy of 85 keV was derived as 0.4320.023 barn by using the reaction rates of Tc, evaluated cross-section data and the fast-neutron flux spectrum of the YAYOI reactor. The present study agreed with the evaluated nuclear data library JENDL-4.0.

Multi-group neutron cross section generation capability for FRENDY nuclear data processing code

Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*

Journal of Nuclear Science and Technology, 58(11), p.1165 - 1183, 2021/11

https://doi.org/10.1080/00223131.2021.1921631

The multi-group cross section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross section generations for all nuclides in JENDL-4.0, -4.0u, -54, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issue, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY or the calculation results by MCNP.

Thermal-neutron capture cross-section measurement of Np using graphite thermal column

Nakamura, Shoji; Endo, Shunsuke; Kimura, Atsushi; Shibahara, Yuji*

KURNS Progress Report 2020, P. 94, 2021/08

https://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2020/ProgRep2020.html

The present study selected Np among radioactive nuclides and aimed to converge a contradiction between reported thermal-neutron capture cross sections. Neutron irradiation was carried out using the graphite thermal column equipped with the Kyoto University Research Reactor. A solution equivalent to 950 Bq order of radioactivity was pipetted out of a Np standard solution and dropped onto a fiber filter, which was then dried with an infrared lamp to prepare a Np sample. The Np sample was quantified using 312-keV gamma ray emitted from Pa in a radiation equilibrium with Np. To monitor a thermal-neutron flux component at an irradiation position, the Np sample was irradiated together with several stable nuclides as neutron flux monitors: Sc, Co, Mo, Ta and Au. The reaction rate of Np was obtained from gamma-ray yields given by Np and Pa, and then the thermal-neutron capture cross section of Np was derived.

Estimation of uncertainty in proton-induced spallation neutron multiplicity for Pb, W, Fe, and C targets

Iwamoto, Hiroki; Meigo, Shinichiro

JPS Conference Proceedings (Internet), 33, p.011046_1 - 011046_6, 2021/03

https://doi.org/10.7566/JPSCP.33.011046

no abstracts in English