Inter-codes and nuclear data comparison under collaboration works between IRSN and JAEA

Gunji, Satoshi; Araki, Shohei; Watanabe, Tomoaki; Fernex, F.*; Leclaire, N.*; Bardelay, A.*; Suyama, Kenya

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10

Institut de radioprotection et de sret nuclaire (IRSN) and Japan Atomic Energy Agency (JAEA) have a long-standing partnership in the field of criticality safety. In this collaboration, IRSN and JAEA are planning a joint experiment using the new STACY critical assembly, modified by JAEA. In order to compare the codes (MVP3, MORET6, etc.) and nuclear data (JENDL and JEFF) used by both institutes in the planning of the STACY experiment, benchmark calculations of the Apparatus B and TCA, which are critical assemblies once owned by both institutes, benchmarks from the ICSBEP handbook and the computational model of the new STACY were performed. Including the new STACY calculation model, the calculations include several different neutron moderation conditions and critical water heights. There were slight systematic differences in the calculation results, which may have originated from the processing and/or format of the nuclear data libraries. However, it was found that the calculated results, including the new codes and the new nuclear data, are in good agreement with the experimental values. Therefore, there are no issues to use them for the design of experiments for the new STACY. Furthermore, the impact of the new TSL data included in JENDL-5 on the effective multiplication factor was investigated. Experimental validation for them will be completed by critical experiments of the new STACY by both institutes.

Present Status of Monte Carlo Seminar for Sub-criticality Safety Analysis in Japan

Sakurai, Kiyoshi; Nojiri, Ichiro*

JAERI-Conf 2003-019, p.855 - 857, 2003/10

https://jopss.jaea.go.jp/pdfdata/JAERI-Conf-2003-019-Part1.pdf

This paper provides overview of sub-criticality safety analysis seminar (July 2000-July 2003, JAERI, total 40 engineers from universities, research institutes and enterprises) for nuclear fuel cycle facility with the Monte Carlo method in Japan. MCNP-4C2 system (MS-DOS version) was installed in each note-type personal computer. Fundamental theory of reactor physics and Monte Carlo simulation including MCNP-4C manual was lectured. Effective neutron multiplication factor and neutron spectrum were calculated for JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. In the seminar, methodology of safety management for nuclear fuel cycle facility was discussed in order to prevent criticality accident.

Effects of volume fraction and non-uniform arrangement of water moderator on reactivity

Cao, X.; Suzaki, Takenori; Kugo, Teruhiko; Mori, Takamasa

JAERI-Tech 2003-069, 36 Pages, 2003/08

JAERI-Tech-2003-069.pdf:3.77MB

From the viewpoint of nuclear criticality safety of fuel rod storage and transport, a series of critical experiments concerning effects of water hole size, water gap width, water-to-fuel volume ratio and non-uniform arrangement of water moderator have been performed at the Tank-type Critical Assembly (TCA) of Japan Atomic Energy Research Institute. In the present study, the effects of volume fraction and non-uniform arrangement of water moderator on reactivity are evaluated by the water level worth method and analyzed by the SRAC code. Error sources of experiments and calculations are discussed, especially for an energy group model. The calculation results of diffusion model with 17-group model show good agreement with the experiment results within a few dozen cents.

Analytical study of two-region TCA critical experiments with PWR-type MOX fuel by using Monte Carlo code MVP

Rahman, M.*; Suzaki, Takenori; Mori, Takamasa

JAERI-Research 2003-007, 68 Pages, 2003/03

JAERI-Research-2003-007.pdf:2.48MB

no abstracts in English

Adjustment of total delayed neutron yields of U, U and Pu by using results of in-pile measurements of effective delayed neutron fraction

Sakurai, Takeshi; Okajima, Shigeaki

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 12 Pages, 2002/10

The cross section adjustment method was applied to total delayed neutron yields of U, U and Pu of the JENDL-3.2 file by using experimental results of effective delayed neutron fraction at six cores built in two fast critical facilities of the MASURCA and FCA and a thermal critical facility of the TCA to improve these yields. The adjustment was carried out on the yields given at several incident neutron energy points in the file. Furthermore, to validate these adjusted delayed neutron yields, analyses were performed for the experiments at ZPR fast critical facility. These adjusted yields brought a reduction of uncertainty of calculated and an improvement in agreement of between experiment and calculation.

Adjustment of total delayed neutron yields of U, U and Pu in JENDL-3.2 using benchmark experiments on effective delayed neutron fraction

Sakurai, Takeshi; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 39(1), p.19 - 30, 2002/01

https://doi.org/10.3327/jnst.39.19

The cross section adjustment method was applied to total delayed neutron yields of U, U and Pu of the JENDL-3.2 file by using experimental results of effective delayed neutron fraction at six cores built in two fast critical facilities of the MASURCA and FCA and a thermal critical facility of the TCA to improve these yields. The adjustment was carried out on the yields given at several incident neutron energy points in the file. After the adjustment, the yield of U was almost uniformly decreased by about 3% below 7 MeV. The yield of Pu was increased by 2.6% and that of U was decreased by 0.9% at the thermal energy point, while the change of yield was less than 0.3% at the other energy points for these nuclides. By using these adjusted yields, the uncertainty of calculated was reduced and the agreement of between experiment and calculation was improved.

Study on reduced-moderation water reactor (RMWR) core design; Joint research report, FY1998-1999 (Joint research)

Research Group for Advanced Reactor System; Research Group for Reactor Physics; Research Group for Thermal and Fluid Engineering

JAERI-Research 2000-035, 316 Pages, 2000/09

JAERI-Research-2000-035.pdf:19.81MB

no abstracts in English

Precise determination of for water-moderated U and U-Pu cores by a method using buckling coefficient of reactivity

Suzaki, Takenori; Sakurai, Kiyoshi; Nakajima, Ken; Horiki, Oichiro*

Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 1, p.386 - 394, 1999/00

no abstracts in English

Estimation of subcriticality with the computed values; Analysis using MCNP of experiment on coupled cores

Sakurai, Kiyoshi; Yamamoto, Toshihiro; Arakawa, Takuya*; *

Nihon Genshiryoku Gakkai-Shi, 40(5), p.380 - 386, 1998/00

https://doi.org/10.3327/jaesj.40.380

no abstracts in English

Studies on burnup credit at JAERI

; Nomura, Yasushi; Suyama, Kenya

Proc. of Int. Conf. on the Phys. of Nucl. Sci. and Technol., 1, p.742 - 748, 1998/00

no abstracts in English