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Motome, Yuiko; Agake, Toshiki; Yanagisawa, Hiroshi
JAEA-Technology 2024-015, 30 Pages, 2025/01
JAEA-Technology-2024-015.pdf:1.36MB
The tables for calibration of control rods were verified, which is used positive period method and improved rod drop method of periodic inspection at Nuclear Safety Research Reactor (NSRR). Those tables are "DOUBLING TIME-REACTIVITY" and "DECAY OF NEUTRON FLUX AFTER INSTANTANEOUS REDUCTION OF REACTIVITY". They are prepared around 1975. Since those tables do not clearly express source of values and records of data used in calculations, the authors verified those tables again. For the verification, the tables were reproduced as follows. For the positive period method, the relationship between the period and reactivity was analytically evaluated by using the inhour equation with NSRR's parameters. For the improved rod drop method, the ratios of neutron flux after the rod drop with parameters of negative reactivities was calculated using the EUREKA- 2 code. As a result, the values described in the tables well agree with those by the present evaluation because it is confirmed that standard deviations of the differences in the value by between the present evaluation and the tables are less than 0.035%. For this reason, it is verified that these tables are valid in the practical use for NSRR operations.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10
The benchmark analyses on the TRIGA reactor of the IEU-COMP-THERM-013 (ICT-013) in the ICSBEP handbook using the MVP version 3 code were carried out with the Japanese, US and European nuclear data libraries, including JENDL-5. The analyses on effective neutron multiplication factors (k
's) were also performed for another TRIGA reactor defined in the ICT-003 as well as the ICT-013. As a result, it is confirmed that the calculated k's vary in the range of 0.8% with the libraries. The results also suggest that there might be unknown bias in the ICT-013 for the calculated k's. For the analyses on the control rod worths of the ICT-013, it is confirmed that differences in the worths among the libraries become smaller than those in k's. Most of the errors involved in k's are considered to be cancelled in the calculation of the worths since the worths are obtained by subtraction of the reciprocals of two sorts of k's. The differences in the worths by between the benchmark and alternative methods defined in the ICT-013 were tried to be examined from the aspect of the differences in horizontal flux distributions by between those methods. It is confirmed that the differences in the worths for two shim control rods, which are fully withdrawn at a delayed critical state, are well understood by that attempt, but on the other hand the differences are not sufficiently explained for a regulating control rod, which is partially inserted to attain delayed criticality.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
JAEA-Technology-2022-030.pdf:2.57MBJAEA-Technology-2022-030(errata).pdf:0.11MB
Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.
Mahmood, M. S.; Nagaya, Yasunobu; Mori, Takamasa
JAERI-Tech 2004-027, 30 Pages, 2004/03
JAERI-Tech-2004-027.pdf:2.26MB
The benchmark experiments of the TRIGA Mark-II reactor in the ICSBEP handbook have been analyzed with the Monte Carlo code MVP using the cross section libraries based on JENDL-3.3, JENDL-3.2 and ENDF/B-VI.8. MCNP calculations have been also performed with the ENDF/B-VI.6 library for comparison between the MVP and MCNP results. For both cores labeled 132 and 133, which have different core configurations, the ratio of the calculated to the experimental results (C/E) for keff obtained by the MVP code is 0.999 for JENDL-3.3, 1.003 for JENDL-3.2, and 0.998 for ENDF/B-VI.8. For the MCNP code, the C/E values are 0.998 for both the Core 132 and 133. All the calculated results agree with the reference values within the experimental uncertainties. The results obtained by MVP with ENDF/B-VI.8 and MCNP with ENDF/B-VI.6 differ only by 0.02% for Core 132, and by 0.01% for Core 133.
Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi; Uetsuka, Hiroshi
Proceedings of 7th International Topical Meeting on Research Reactor Fuel Management (ENS RRFM2003), p.109 - 113, 2003/03
Uranium-zirconium hydride (U-ZrHx) fuel has been widely utilized in the world as TRIGA reactor fuel. In order to obtain the fuel performance data under accident conditions and to enhance accountability of the safety assessment of the reactors using the fuel, irradiation tests under power burst type accident conditions have been conducted in the NSRR. Five pulse irradiation tests have been performed at peak fuel enthalpies ranging from 187 J/g to 483 J/g. Cladding surface temperature increased rapidly at the pulse and DNB occurred in peak fuel enthalpy over 187 J/g in the tests. The DNB occurred at lower fuel enthalpy in the U-ZrH1.6 fuel than in the UO
fuel rods. The rod internal pressure rose up to as high as 1MPa in the transient heating tests, suggesting considerable release of the hydrogen decomposed from the fuel. The peak pressure was lower than equilibrium hydrogen pressure of 1.5MPa at the peak temperature, suggesting the transient effect. Considerable PCMI was observed in the tests, through cladding elongation up to 3.3 mm synchronized to the pellet stack deformation.
Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; ; ;
JAERI-Tech 98-031, 225 Pages, 1998/08
no abstracts in English
W.S.Charlton*; T.A.Parish*; S.Raman*; Shinohara, Nobuo; Ando, Masaki
PHYSOR 96: Int. Conf. on the Physics of Reactors, 3, p.F11 - F20, 1996/00
no abstracts in English
Fuketa, Toyoshi; Ishijima, Kiyomi; Tanzawa, Sadamitsu; Nakamura, Takehiko; Sasajima, Hideo; Kashima, Yoichi; ;
JAERI-Research 95-005, 53 Pages, 1995/01
JAERI-Research-95-005.pdf:1.96MB
no abstracts in English
Kaminaga, Masanori
JAERI-M 94-052, 40 Pages, 1994/03
no abstracts in English
Nakano, Yoshihiro; Ichikawa, Hiroki; Nakajima, Teruo
Proc. of the 16th Int. Meeting on Reduced Enrichment for Research and Test Reactors, 0, p.313 - 320, 1994/03
no abstracts in English
Yanagisawa, Kazuaki; Ugajin, Mitsuhiro;
Kaku Nenryo Kogaku; Genjo To Tembo, p.285 - 304, 1993/11
no abstracts in English
Inabe, Teruo; Ishijima, Kiyomi; Tanzawa, Sadamitsu; Shimazaki, Junya; Nakamura, Takehiko; ; ; ; ; ; et al.
JAERI-M 88-113, 55 Pages, 1988/06
no abstracts in English
;
JAERI-M 5361, 48 Pages, 1973/08
no abstracts in English
