Improving the safety of the high temperature gas-cooled reactor "HTTR" based on Japan's new regulatory requirements

Hamamoto, Shimpei; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Sekita, Kenji; Watanabe, Shuji; Furusawa, Takayuki; Iigaki, Kazuhiko; et al.

Nuclear Engineering and Design, 388, p.111642_1 - 111642_11, 2022/03

https://doi.org/10.1016/j.nucengdes.2021.111642

Following the Fukushima Daiichi Nuclear Power Plant accident in 2011, the Japan Atomic Energy Agency adapted High-Temperature engineering Test Reactor (HTTR) to meet the new regulatory requirements that began in December 2013. The safety and seismic classifications of the existing structures, systems, and components were discussed to reflect insights regarding High Temperature Gas-cooled Reactors (HTGRs) that were acquired through various HTTR safety tests. Structures, systems, and components that are subject to protection have been defined, and countermeasures to manage internal and external hazards that affect safety functions have been strengthened. Additionally, measures are in place to control accidents that may cause large amounts of radioactive material to be released, as a beyond design based accident. The Nuclear Regulatory Commission rigorously and appropriately reviewed this approach for compliance with the new regulatory requirements. After nine amendments, the application to modify the HTTR's installation license that was submitted in November 2014 was approved in June 2020. This response shows that facilities can reasonably be designed to meet the enhanced regulatory requirements, if they reflect the characteristics of HTGRs. We believe that we have established a reference for future development of HTGR.

Observation of nuclear-spin Seebeck effect

Kikkawa, Takashi*; Reitz, D.*; Ito, Hiroaki*; Makiuchi, Takahiko*; Sugimoto, Takaaki*; Tsunekawa, Kakeru*; Daimon, Shunsuke*; Oyanagi, Koichi*; Ramos, R.*; Takahashi, Saburo*; et al.

Nature Communications (Internet), 12, p.4356_1 - 4356_7, 2021/07

https://doi.org/10.1038/s41467-021-24623-6

Development of module for ADS nitride fuel performance analysis

Shibata, Hiroki; Saito, Hiroaki; Hayashi, Hirokazu; Takano, Masahide

JAEA-Data/Code 2019-023, 138 Pages, 2020/03

JAEA-Data-Code-2019-023.pdf:6.99MB

Transmutation of minor actinides in the form of nitride fuel by the accelerator driven system has been developed to reduce the radiotoxicity and volume in the radioactive wastes. Nitride fuel behavior under irradiation condition is necessary for its design and development. Nitride fuel performance analysis module based on light water reactor fuel performance code, FEMAXI-7, was developed by introducing fundamental properties of nitride pellet, 9Cr-1Mo ferrite cladding, and Pi-Bi coolant. As a result of test analysis with this module, we have understood that the nitride fuel shows excellent behavior under irradiation due to its high thermal conductivity. We found that, however, it may be a main concern that fuel cladding integrity is maintained during irradiation in which pellet-cladding mechanical interaction is increased by He gas release, low creep rate of nitride pellet at low temperatures, and high creep rate of cladding above 873 K.

Temporal change in radiological environments on land after the Fukushima Daiichi Nuclear Power Plant accident

Saito, Kimiaki; Mikami, Satoshi; Ando, Masaki; Matsuda, Norihiro; Kinase, Sakae; Tsuda, Shuichi; Sato, Tetsuro*; Seki, Akiyuki; Sanada, Yukihisa; Wainwright-Murakami, Haruko*; et al.

Journal of Radiation Protection and Research, 44(4), p.128 - 148, 2019/12

https://doi.org/10.14407/jrpr.2019.44.4.128

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 7; Analysis on effectiveness of spray cooling by the SAMPSON code

Suzuki, Hiroaki*; Morita, Yoshihiro*; Naito, Masanori*; Nemoto, Yoshiyuki; Nagatake, Taku; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

In this paper, modification of the SAMPSON code was carried out to enable the analysis of spray cooling. The SAMPSON analysis of a spray cooling experiment was performed to confirm reproducibility of spray cooling behavior of fuel claddings. The modified SAMPSON code was applied to a hypothetical loss-of-coolant accident analysis of the SFP. Effectiveness of spray cooling on cladding temperature behavior was investigated. The SAMPSON analysis showed that spraying from the top of the SFP was effective for cooling the fuel assemblies exposed to the gas phase.

Neutron scattering study of yttrium iron garnet

Shamoto, Shinichi; Ito, Takashi; Onishi, Hiroaki; Yamauchi, Hiroki; Inamura, Yasuhiro; Matsuura, Masato*; Akatsu, Mitsuhiro*; Kodama, Katsuaki; Nakao, Akiko*; Moyoshi, Taketo*; et al.

Physical Review B, 97(5), p.054429_1 - 054429_9, 2018/02

https://doi.org/10.1103/PhysRevB.97.054429

Nuclear and magnetic structure and full magnon dispersions of yttrium iron garnet YFeO have been studied by neutron scattering. The lowest-energy dispersion below 14 meV exhibits a quadratic dispersion as expected from ferromagnetic magnons. The imaginary part of -integrated dynamical spin susceptibility "() exhibits a square-root energy-dependence in the low energies. The magnon density of state is estimated from the "() obtained on an absolute scale. The value is consistent with a single chirality mode for the magnon branch expected theoretically.

The Applicability of SiC-SiC fuel cladding to conventional PWR power plant

Furumoto, Kenichiro*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Teshima, Hideyuki*; Yamashita, Shinichiro; Saito, Hiroaki; Shirasu, Noriko

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/F-225-PD2.pdf

Since 2015, Mitsubishi Nuclear Fuel (MNF) has joined in a Japanese R&D project of ATF founded by the Ministry of Economy, Trade and Industry (METI) as a subcontractor to Japan Atomic Energy Agency (JAEA) which is the prime contractor to METI. In this program, MNF plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). This paper reports the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding. As out-of-pile tests, thermal diffusivity measurement and autoclave corrosion test for SiC composite samples were performed. Test apparatus were developed for evaluation of performance of SiC composite under the condition simulated design basis accident (DBA).

Study on the deterioration mechanism of layered rock-salt electrodes using epitaxial thin films; Li(Ni, Co, Mn)O and their Zr-O surface modified electrodes

Abe, Machiko*; Iba, Hideki*; Suzuki, Kota*; Minamishima, Hiroaki*; Hirayama, Masaaki*; Tamura, Kazuhisa; Mizuki, Junichiro*; Saito, Tomohiro*; Ikuhara, Yuichi*; Kanno, Ryoji*

Journal of Power Sources, 345, p.108 - 119, 2017/03

https://doi.org/10.1016/j.jpowsour.2017.01.103

The surface structure of the Li(Ni, Co, Mn)O electrode was studied during charge/discharge process using electrochemical methods and X-ray/Neutron scattering techniques. It was found that during charge/discharge process the coverage of spinel structure increased. The spinel structure has low electrochemical activity and is not involved in Li insertion/extraction. After the surface modification, it was found that the coverage of the spinel structure did not increase. Further, it was also found out that the Li concentration at the electrode/electrolyte interface increased.

Synthesized research report in the second mid-term research phase, Mizunami Underground Research Laboratory Project, Horonobe Underground Research Laboratory Project and Geo-stability Project (Translated document)

Hama, Katsuhiro; Sasao, Eiji; Iwatsuki, Teruki; Onoe, Hironori; Sato, Toshinori; Fujita, Tomoo; Sasamoto, Hiroshi; Matsuoka, Toshiyuki; Takeda, Masaki; Aoyagi, Kazuhei; et al.

JAEA-Review 2016-014, 274 Pages, 2016/08

JAEA-Review-2016-014.pdf:44.45MB

We synthesized the research results from the Mizunami/Horonobe Underground Research Laboratories (URLs) and geo-stability projects in the second midterm research phase. This report can be used as a technical basis for the Nuclear Waste Management Organization of Japan/Regulator at each decision point from siting to beginning of disposal (Principal Investigation to Detailed Investigation Phase).

Baseline design of a proton linac for BNCT at OIST

Kondo, Yasuhiro; Hasegawa, Kazuo; Higashi, Yasuo*; Sugawara, Hirotaka*; Yoshioka, Masakazu*; Kumada, Hiroaki*; Matsumoto, Hiroshi*; Naito, Fujio*; Kurokawa, Shinichi*

Proceedings of 7th International Particle Accelerator Conference (IPAC '16) (Internet), p.906 - 909, 2016/06

https://doi.org/10.18429/JACoW-IPAC2016-MOPOY026

An accelerator based boron neutron capture therapy (BNCT) facility is being planned at Okinawa institute of science and technology (OIST). The proton accelerator consists of a radio frequency quadrupole (RFQ) linac and a drift tube linac (DTL). The required beam power is 60 kW. The present beam energy and current are 10 MeV and 30 mA, respectively. The pulse length is 3.3 ms and the repetition rate is 60 Hz, therefore, the duty factor is 20%. In this paper, present design of this compact, medium current, high duty proton linac is presented.