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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.
Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021115_1 - 021115_11, 2020/04
The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics -annular core pulse reactor type. The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions.
Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
The NSRR is a research reactor of TRIGA-ACPR type, located in the Nuclear Science Research Institute. The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident conditions. Under the new regulation standards after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami. As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.
Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2017-007, 18 Pages, 2017/03
JAEA-Technology-2017-007.pdf:2.16MB
Nuclear Safety Research Reactor (NSRR) facility have been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident conditions. Unirradiated test fuels used in fuel irradiation experiments and flesh driver fuel elements for reactor operation are stored in the fuel building of the facility. In response to the 2011 off the Pacific coast of Tohoku Earthquake, the impact of NSRR's nuclear fuel material usage facilities on external events beyond design requirements was evaluated. The subcriticality of the flesh fuel storage was confirmed in consideration of earthquake and tsunami as superimposed event.
Motome, Yuiko; Nakatsuka, Toru; Amaya, Masaki; Yonomoto, Taisuke
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Motome, Yuiko; Tamaki, Hitoshi; Yonomoto, Taisuke; Amaya, Masaki
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Agake, Toshiki; Motome, Yuiko; Yanagisawa, Hiroshi; Kobayashi, Tetsuya
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Kobayashi, Tetsuya; Motome, Yuiko; Akiyama, Yoshiya; Yoshida, Soma; Agake, Toshiki; Aizawa, Kazuki
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Nakatsuka, Toru; Tsumura, Takashi; Motome, Yuiko; Amaya, Masaki; Yonomoto, Taisuke
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Motome, Yuiko; Tamaki, Hitoshi; Yonomoto, Taisuke; Amaya, Masaki
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Motome, Yuiko; Nakatsuka, Toru; Amaya, Masaki; Yonomoto, Taisuke
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Motome, Yuiko; Amaya, Masaki; Yonomoto, Taisuke
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Motome, Yuiko; Tobita, Toru; Yonomoto, Taisuke; Amaya, Masaki
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Motome, Yuiko; Tamaki, Hitoshi; Yonomoto, Taisuke; Amaya, Masaki
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Murao, Hiroyuki; Motome, Yuiko
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