Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

https://doi.org/10.1299/jsmeicone.2019.27.1343

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Replacement of the glove port equipped with glove box in Nuclear Fuel Reprocessing Facility

Horigome, Kazushi; Taguchi, Shigeo; Nishida, Naoki; Goto, Yuichi; Inada, Satoshi; Kuno, Takehiko

Nihon Hozen Gakkai Dai-14-Kai Gakujutsu Koenkai Yoshishu, p.381 - 384, 2017/08

no abstracts in English

A Study on transmutation of LLFPs using various types of HTGRs

Kora, Kazuki*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi*

Nuclear Engineering and Design, 300, p.330 - 338, 2016/04

https://doi.org/10.1016/j.nucengdes.2016.01.030

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system "GTHTR300", a small modular HTGR plant "HTR50S" and two types of plutonium burner HTGRs "Clean Burn with MA" and "Clean Burn without MA" were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Radionuclide release to stagnant water in the Fukushima-1 Nuclear Power Plant

Nishihara, Kenji; Yamagishi, Isao; Yasuda, Kenichiro; Ishimori, Kenichiro; Tanaka, Kiwamu; Kuno, Takehiko; Inada, Satoshi; Goto, Yuichi

Journal of Nuclear Science and Technology, 52(3), p.301 - 307, 2015/03

https://doi.org/10.1080/00223131.2014.946455

After the severe accident at the Fukushima-1 nuclear power plant, large amounts of contaminated stagnant water have accumulated in turbine buildings and their surroundings. This rapid communication reports calculation of the radionuclide inventory in the core, collection of measured inventory in the stagnant water, and estimation of radionuclide release ratios from the core to the stagnant water. This evaluation is based on data obtained before June 3, 2011. The release ratios of tritium, iodine, and cesium were several tens of percent, whereas those of strontium and barium were smaller by one or two orders of magnitude. The release ratios in the Fukushima accident were equivalent to those in the TMI-2 accident.

Study on methodology to estimate isotope generation and depletion for core design of HTGR

Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Shimakawa, Satoshi

JAEA-Research 2013-035, 84 Pages, 2013/12

JAEA-Research-2013-035.pdf:3.22MB

An investigation on methodology to estimate isotope generation and depletion had been performed in order to improve the accuracy for HTGR core design. Solving the burn-up equations, generating effective cross section and employing nuclide data are the technical problems. Especially for the generating effective cross section, the core burn-up calculation has a technological problem in common with point burn-up calculation. Thus, the investigation had also been performed for the core burn-up calculation to develop new code system in the future. As a result, it was found that the cross section with the extended 108 energy groups structure from the SRAC 107 groups structure to 20 MeV and the cross section collapse using the flux obtained by the deterministic code SRAC is proper for the use. In addition, an investigation on the preparation condition for nuclear data for a safety analysis and a fuel design was also performed. As a result, the needs for the nuclear data ware made clear.

A Study of applicability of JENDL-4.0 to the HTTR criticality analysis

Goto, Minoru; Shimakawa, Satoshi; Tachibana, Yukio

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 6 Pages, 2011/10

In the past, benchmark calculations of criticality approach for the HTTR, which is a Japanese HTGR, were performed by research institutes in several countries, and almost all of the calculations overestimated the excess reactivity. In Japan, the benchmark calculations were performed by JAEA, and were also resulted in overestimation. JAEA improved the calculations by revising the geometric model and replacing the nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem has not been resolved until today. We performed calculations of the HTTR criticality approach with several nuclear data libraries, and found that slight difference in the neutron capture cross section of carbon at thermal energy among the libraries causes significant difference in the values. The cross section value of carbon was not concerned in reactor neutronics calculation because of its small value of the order of 1E-3 burn, and consequently the cross section value had not been revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We thought that the cross section should be revised based on the latest measurement data to improve the accuracy of the HTGR criticality analysis. In May 2010, the latest JENDL (JENDL-4.0) was released by JAEA, and the capture cross section of carbon was revised. JENDL-4.0 yielded 0.4-0.9%k/k smaller values than JENDL-3.3 in the criticality calculations for the HTTR critical approach, and consequently the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved by replacing the nuclear data libraries with JENDL-4.0.

Core design study of small-sized high temperature reactor for electricity generation

Goto, Minoru; Shimakawa, Satoshi; Terada, Atsuhiko; Shibata, Taiju; Tachibana, Yukio; Kunitomi, Kazuhiko

Proceedings of ASME 2011 Small Modular Reactors Symposium (SMR 2011) (CD-ROM), 5 Pages, 2011/09

https://doi.org/10.1115/SMR2011-6576

The present study challenges the core design of a small-sized reactor for long refueling interval by increasing core size, fuel loading and fuel burn up compared with the High Temperature engineering Test Reactor (HTTR). The core burn-up calculation suggested that approximately 6 years of long refueling interval was found to be reasonably achieved with operational reactor power of 120 MWt.

Identified charged hadron production in collisions at = 200 and 62.4 GeV

Adare, A.*; Afanasiev, S.*; Aidala, C.*; Ajitanand, N. N.*; Akiba, Yasuyuki*; Al-Bataineh, H.*; Alexander, J.*; Aoki, Kazuya*; Aphecetche, L.*; Armendariz, R.*; et al.

Physical Review C, 83(6), p.064903_1 - 064903_29, 2011/06

https://doi.org/10.1103/PhysRevC.83.064903

Transverse momentum distributions and yields for , and in collisions at = 200 and 62.4 GeV at midrapidity are measured by the PHENIX experiment at the RHIC. We present the inverse slope parameter, mean transverse momentum, and yield per unit rapidity at each energy, and compare them to other measurements at different collisions. We also present the scaling properties such as and scaling and discuss the mechanism of the particle production in collisions. The measured spectra are compared to next-to-leading order perturbative QCD calculations.

Long-term high-temperature operation in the HTTR, 2; Core physics

Goto, Minoru; Fujimoto, Nozomu; Shimakawa, Satoshi; Tachibana, Yukio; Nishihara, Tetsuo; Iyoku, Tatsuo

Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 8 Pages, 2010/10

In the High Temperature Engineering Test Reactor (HTTR), which is a Japanese block-type HTGR, reactivity is controlled by control rods (CRs) and burnable poisons (BPs). The CRs insertion depth into the core should be retained shallow during burnup period, because the large insertion depth leads to significant disturbance of the power distribution, and consequently fuel temperature rises above the limit. Thus, the controllable reactivity with the CRs during operation is small, and then reactivity control through the burnup period largely depends on the BPs. It has not been confirmed an effectiveness of BPs on reactivity control on block-type HTGRs. The HTTR succeeded in long-term high temperature operation, and its burnup reached about 370EFPD. Thereby it became possible to confirm the effectiveness of BPs on reactivity control on the HTTR using its burnup data. We focused on a burnup change in the CRs insertion depth into the core to confirm whether the BPs functioned as designed. Additionally, we compared the change in the CRs insertion depths between analysis results and the experimental data to confirm validity of a whole core burnup calculation with the SRAC/COREBN. As a result, the experimental data showed that although the CRs insertion depth into the core was increased with burnup, it was retained the allowable depth. Meanwhile, the analysis result of the CRs insertion depth was in good agreement with the experimental data.

Impact of capture cross-section of carbon on nuclear design for HTGRs

Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki; Tachibana, Yukio

Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 6 Pages, 2010/10

Capture cross section of carbon in thermal energy range has been regarded as unimportant in neutronics calculations on general reactor design, because of its infinitesimal value of only 3 mb at 2200 m/s. However, it is not negligible for design works for graphite-rich reactors, such as the High Temperature Gas-cooled Reactors (HTGRs). For the High Temperature Engineering Test Reactor (HTTR) of JAEA, five percent differences in capture cross section of carbon makes 0.24% change in thermal utilization factor of the four factor formula. This impact is for the HTTR with a core configuration of full-loaded core, named the packed core. In this case, change of multiplier factor will be equivalent to a change of thermal utilization factor. The impact of the cross section is dependent on an atomic number ratio of graphite/235-uranimu in reactor core. For more graphite-rich core such as the HTTR with ring core configuration, the five percent change of the cross section value makes a 0.47% on multiplier factor. From our studies in the HTTR analysis, a value of capture cross section at 2200 m/s has been revised to 3.86 mb in evaluated nuclear data library of JENDL-4. Comparing with the value of JENDL4, the values in other libraries are about 10-15% smaller as 3.36 mb in ENDF/B-VII, 3.36 mb in JEFF-3.1 and 3.53 mb in JENDL-3.3. It was observed that discrepancy of a multiplier factor between former calculation and experiment of the HTTR showed disagreement in the evaluation of the critical approach tests. Monte Carlo calculation results using JENDL3.3 are overestimated about 0.4% with packed core configuration and 1.0% with ring core, respectively. In this report, the improvement of excess reactivity calculation for the HTTR with newly JENDL-4 is described.