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Journal Articles

Japanese Evaluated Nuclear Data Library version 5; JENDL-5

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.

Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01

 Times Cited Count:1

Journal Articles

Si-addition contributes to overcoming the strength-ductility trade-off in high-entropy alloys

Wei, D.*; Gong, W.; Tsuru, Tomohito; Lobzenko, I.; Li, X.*; Harjo, S.; Kawasaki, Takuro; Do, H.-S.*; Bae, J. W.*; Wagner, C.*; et al.

International Journal of Plasticity, 159, p.103443_1 - 103443_18, 2022/12

 Times Cited Count:5 Percentile:86.66(Engineering, Mechanical)

Journal Articles

Implementation of resonance upscattering treatment in FRENDY nuclear data processing system

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

Nuclear Science and Engineering, 196(11), p.1267 - 1279, 2022/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The resonance upscattering effect (the thermal agitation effect) is incorporated in the generation capability of multi-group neutron cross sections of the FRENDY nuclear data processing system. The resonance upscattering effect is considered by (1) the variation of self-shielding factors (effective cross sections) due to the change in ultra-fine group spectrum and (2) the variation of group-to-group elastic scattering cross sections. In the verification calculations, impacts on the ultra-fine group spectrum, effective cross sections, and neutronics characteristics (the Doppler effect) are confirmed. The effect of energy group structure and the treatments of resonance upscattering on the Doppler effect through the variation of effective cross sections and the elastic scattering matrix are studied. The results indicate that the FRENDY can provide appropriate multi-group cross sections considering the resonance upscattering effect.

Journal Articles

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, 8 Pages, 2022/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

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.

Journal Articles

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.

Journal Articles

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

 Times Cited Count:5 Percentile:83.31(Nuclear Science & Technology)

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, -5$$alpha$$4, 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.

Journal Articles

Verification of the multi-group generation capability of FRENDY nuclear data processing code for recent nuclear data through comparison of one-group reaction rates

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

Transactions of the American Nuclear Society, 124(1), p.544 - 547, 2021/06

Verification calculations for the capability of multi-group cross section generation in FRENDY (FRENDY/MG) are carried out through the comparison of one-group reaction rates using the multi-group cross sections obtained by FRENDY/MG and NJOY2016. Three different neutron spectra (LWR, FR, and 1/E) are used to calculate one-group reaction rates. The discrepancies of one-group reaction rates are small for most cases, showing the validity of FRENDY/MG. The FRENDY/MG will be released as the part of FRENDY nuclear data processing system in the near future.

Journal Articles

Multi-group cross section library generation by FRENDY for fast reactor neutronics calculations

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

Transactions of the American Nuclear Society, 124(1), p.556 - 558, 2021/06

The FRENDY nuclear data processing code has been used to generate multi-group cross section libraries for the CBZ reactor physics code system. The newly generated libraries have been applied to neutronics calculations of a fast reactor core MET-1000, and several neutronics parameters are calculated. Calculations with other libraries generated by NJOY2016 have been also conducted, and differences in obtained neutronics parameters between the FRENDY-based library and the NJOY-based library have been quantified. Generally reasonable agreement between them has been obtained, so it has been demonstrated that the multi-group libraries for fast reactor neutronics calculations can be generated successfully by FRENDY. Detailed investigation on the impact of the difference in the processing codes on k-effective has been also carried out with a help of the perturbation theory, and the causes of the differences have been identified.

Journal Articles

Development of a reference database for beta-delayed neutron emission

Dimitriou, P.*; Dillmann, I.*; Singh, B.*; Piksaikin, V.*; Rykaczewski, K. P.*; Tain, J. L.*; Algora, A.*; Banerjee, K.*; Borzov, I. N.*; Cano-Ott, D.*; et al.

Nuclear Data Sheets, 173, p.144 - 238, 2021/03

 Times Cited Count:12 Percentile:94.13(Physics, Nuclear)

$$beta$$-delayed neutron emission has been of interest since the discovery of nuclear fission. In nuclear power reactors, delayed-neutron data play a crucial role in reactor kinetics calculations and safe operation. $$beta$$-delayed neutron data also have a significant impact in the field of nuclear structure and astrophysics especially as nuclei farther away from stability are explored at the new generation of radioactive beam facilities. Several compilations of $$beta$$-decay half-lives and delayed-neutron emission probabilities are available, however, complete documentation of measurements and evaluation procedures is often missing for these properties. Efforts to address this gap in nuclear data and create an updated compilation and evaluation of $$beta$$-delayed neutron properties were undertaken under the auspices of the International Atomic Energy Agency (IAEA) which formed a Coordinated Research Project (CRP) on "Development of a Reference Database of Beta-delayed Neutron Emission Data". In this paper we summarize the work that was performed and present the results of the CRP.

Journal Articles

First nuclear transmutation of $$^{237}$$Np and $$^{241}$$Am by accelerator-driven system at Kyoto University Critical Assembly

Pyeon, C. H.*; Yamanaka, Masao*; Oizumi, Akito; Fukushima, Masahiro; Chiba, Go*; Watanabe, Kenichi*; Endo, Tomohiro*; Van Rooijen, W. G.*; Hashimoto, Kengo*; Sakon, Atsushi*; et al.

Journal of Nuclear Science and Technology, 56(8), p.684 - 689, 2019/08

 Times Cited Count:12 Percentile:86.38(Nuclear Science & Technology)

This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 ($$^{237}$$Np) and americium-241 ($$^{241}$$Am), and capture reactions of $$^{237}$$Np. Subcritical irradiation of $$^{237}$$Np and $$^{241}$$Am foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test ($$^{237}$$Np or $$^{241}$$Am) and reference (uranium-235) foils. The first nuclear transmutation of $$^{237}$$Np and $$^{241}$$Am by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.

Journal Articles

Measurement of MA reaction rates under sub-critical condition with spallation neutron source in A-core of KUCA for ADS

Oizumi, Akito; Fukushima, Masahiro; Tsujimoto, Kazufumi; Chiba, Go*; Yamanaka, Masao*; Sano, Tadafumi*; Pyeon, C. H.*

KURNS Progress Report 2018, P. 38, 2019/08

In the nuclear transmutation system such as ADS, the nuclear data validation of MA is required to reduce the uncertainty caused by the nuclear data of MA. This study aims to measure the fission reaction rate ratios (FRRs) of Neptunium-237 ($$^{237}$$Np) or Americium-241 ($$^{241}$$Am) to Uranium-235 ($$^{235}$$U) by using a back-to-back (BTB) fission chamber in the KUCA built as a sub-critical core (k$$_{rm eff}$$ = 0.998) with the nuclear spallation neutron source. The result showed that the measured FRRs of $$^{237}$$Np/$$^{235}$$U and $$^{241}$$Am/$$^{235}$$U were 0.014 $$pm$$0.002 and 0.023 $$pm$$0.005, respectively. These measured values will be used for verification of evaluated nuclear data by conducting detailed analyses.

Journal Articles

Cutting-edge studies on nuclear data for continuous and emerging need, 8; Evolution of the nuclear data library JENDL

Iwamoto, Osamu; Shibata, Keiichi; Iwamoto, Nobuyuki; Chiba, Go*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(6), p.357 - 361, 2018/06

Nuclear data library consists of the results of related studies on nuclear data. Nuclear data can show worth through utilization of the nuclear data library which is the outcome of the nuclear data study. As the last lecture over 8 serial ones, the Japanese nuclear data library JENDL is explained. Sections of "General purpose file and its history", "recent progress of special purpose file", and "international status of nuclear data library" are introduced and one of "prospect of JENDL and nuclear data study" is shown.

Journal Articles

Research and development roadmap for reactor physics 2017; Future of reactor physics projected by next generations

Yamamoto, Akio*; Chiba, Go*; Kirimura, Kazuki*; Miki, Yosuke*; Yokoyama, Kenji

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.241 - 245, 2018/04

no abstracts in English

Journal Articles

Role of multichance fission in the description of fission-fragment mass distributions at high energies

Hirose, Kentaro; Nishio, Katsuhisa; Tanaka, Shoya*; L$'e$guillon, R.*; Makii, Hiroyuki; Nishinaka, Ichiro*; Orlandi, R.; Tsukada, Kazuaki; Smallcombe, J.*; Vermeulen, M. J.; et al.

Physical Review Letters, 119(22), p.222501_1 - 222501_6, 2017/12

 Times Cited Count:40 Percentile:90.27(Physics, Multidisciplinary)

Fission-fragment mass distributions were measured for $$^{237-240}$$U, $$^{239-242}$$Np and $$^{241-244}$$Pu populated in the excitation-energy range from 10 to 60 MeV by multi-nucleon transfer channels in the reaction $$^{18}$$O + $$^{238}$$U at the JAEA tandem facility. Among them, the data for $$^{240}$$U and $$^{240,241,242}$$Np were observed for the first time. It was found that the mass distributions for all the studied nuclides maintain a double-humped shape up to the highest measured energy in contrast to expectations of predominantly symmetric fission due to the washing out of nuclear shell effects. From a comparison with the dynamical calculation based on the fluctuation-dissipation model, this behavior of the mass distributions was unambiguously attributed to the effect of multi-chance fission.

Journal Articles

Cutting-edge studies on nuclear data for continuous and emerging need, 1; Diversifying nuclear applications and need for the nuclear data

Suyama, Kenya; Kunieda, Satoshi; Fukahori, Tokio; Chiba, Go*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 59(10), p.598 - 602, 2017/10

The nuclear data is the data on the reaction probability between the neutron and the nuclide in a narrow sense. However generally speaking, it is the data describing the physical change of the nuclide and the status of the nuclear ration. Since Japan had started the nuclear energy development, the nuclear data has been one of the most important technical development theme. Now, the nuclear data library of Japan, i.e., JENDL, is well recognized internationally because of the highest-accuracy and fully-furnished types of the included data. This serial lecture describes the significance and the status of the nuclear data development, the international trend, and the direction of the future development.

Journal Articles

Simultaneous measurement of neutron-induced fission and capture cross sections for $$^{241}$$Am at neutron energies below fission threshold

Hirose, Kentaro; Nishio, Katsuhisa; Makii, Hiroyuki; Nishinaka, Ichiro*; Ota, Shuya*; Nagayama, Tatsuro*; Tamura, Nobuyuki*; Goto, Shinichi*; Andreyev, A. N.; Vermeulen, M. J.; et al.

Nuclear Instruments and Methods in Physics Research A, 856, p.133 - 138, 2017/06

 Times Cited Count:3 Percentile:32.55(Instruments & Instrumentation)

Journal Articles

Fission fragments mass distributions of nuclei populated by the multinucleon transfer channels of the $$^{18}$$O + $$^{232}$$Th reaction

L$'e$guillon, R.; Nishio, Katsuhisa; Hirose, Kentaro; Makii, Hiroyuki; Nishinaka, Ichiro*; Orlandi, R.; Tsukada, Kazuaki; Smallcombe, J.*; Chiba, Satoshi*; Aritomo, Yoshihiro*; et al.

Physics Letters B, 761, p.125 - 130, 2016/10

 Times Cited Count:35 Percentile:92.49(Astronomy & Astrophysics)

JAEA Reports

Proceedings of the International Conference on Physics of Reactors (PHYSOR 2014); September 28-October 3, 2014, Kyoto, Japan

Suyama, Kenya; Sugawara, Takanori; Tada, Kenichi; Chiba, Go*; Yamamoto, Akio*

JAEA-Conf 2014-003, 76 Pages, 2015/03


Japan Atomic Energy Agency organized an international conference PHYSOR 2014 on the reactor physics which is one of basic researches in the nuclear engineering, in cooperation with Research Reactor Institute of Kyoto University. PHYSOR is the world's largest scale international conference in the reactor physics field. It originates in the conference held in Marseille, France in 1990, which originally had been organized in the United States as a Physics of Reactors Topical Meeting of the reactor physics division of the American Nuclear Society every two years. More than 500 papers had been submitted and finally 472 papers were presented in the conference after the paper review process. This report contains the presented papers, which the PHYSOR organizing committee has decided to publish in an official JAEA report with the permission by authors, except for several selected papers to be published in the Journal of Nuclear Science and Technology of the Atomic Energy Society of Japan.

Journal Articles

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Chiba, Satoshi*

Journal of Nuclear Science and Technology, 51(6), p.818 - 831, 2014/06


 Times Cited Count:15 Percentile:77.42(Nuclear Science & Technology)

A plutonium burner HTGR concept of "Clean Burn" has been investigated based on the experience with the HTTR construction and operation, and the GTHTR300 design. The concept aims at effective and safe consumption of the surplus plutonium from LWR spent fuel once through. The Clean Burn employs IMF that can be fabricated without a need to mix the plutonium with uranium. In addition, the incineration of the plutonium can be performed effectively without plutonium generation from the matrix and the reactivity defect. The IMF is also made into a highly stable form to keep fuel integrity and geological disposal stability. The particular fabrication system and the stable form make the IMF more proliferation resistant than the conventional LWR-MOX fuel. This advantage is attractive to the Japanese nuclear fuel cycle policy. Moreover, the Clean Burn has naturally safe feature because based on HTGR concept. A detailed investigation have shown that the proposed concept is effective and promising.

JAEA Reports

Development of a standard data base for FBR core design, 14; Analyses of extensive FBR core characteristics based on JENDL-4.0

Sugino, Kazuteru; Ishikawa, Makoto; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Nagaya, Yasunobu; Hazama, Taira; Chiba, Go*; Yokoyama, Kenji; Kugo, Teruhiko

JAEA-Research 2012-013, 411 Pages, 2012/07


Aiming at evaluating the core design prediction accuracy of fast reactors, various kinds of fast reactor core experiments/tests have been analyzed with the Japan's latest evaluated nuclear data library JENDL-4.0. Totally 643 characteristics of reactor physics experiments/tests and irradiation tests performed using the critical facilities: ZPPR, FCA, ZEBRA, BFS, MASURCA, ultra-small cores of LANL and power plants: SEFOR, Joyo, Monju were dealt. In analyses, a standard scheme/method for fast reactor cores was applied including detailed or precise calculations for best estimation. In addition, results of analyses were investigated from the viewpoints of uncertainties caused by experiment/test, analytical modeling and cross-section data in order to synthetically evaluate the consistency among different cores and characteristics. Further, by utilizing these evaluations, prediction accuracy of core characteristics were evaluated for fast power reactor cores that are under designing in the fast reactor cycle technology development (FaCT) project.

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