Development of Behavior Analysis Code for MA Transmutation Nitride Fuel in Accelerator-Driven System

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

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 23(3), p.74 - 80, 2024/08

https://doi.org/10.3327/taesj.J23.018

Partitioning and transmutation of minor actinides techniques have been developed to reduce the radiotoxicity and volume in the high-level radioactive wastes. Minor actinide nitride fuel has been chosen as a candidate for transmutation of long-lived nuclides by accelerator-driven system. Understanding irradiation behavior of nitride fuel is important for its design and development, however, experimental data on irradiation tests of actinide nitrides and these solid solutions are scarce. Recently, in JAEA, nitride fuel performance analysis module based on light water reactor fuel performance code, FEMAXI-7, has been developed to simulate irradiation behavior of the nitride fuel. In this study, performance analysis was carried out focusing on the pellet-cladding mechanical interaction (PCMI), which was pointed out as the most effective factor for the fuel safety during irradiation. Simulation results show that PCMI does not cause the creep rupture of the cladding.

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.

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

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

http://wrfpm2017.org/upload/draft_file/F-226-PD5.pdf

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 99 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

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.

Lattice structure transformation and change in surface hardness of NiNb and NiTa intermetallic compounds induced by energetic ion beam irradiation

Kojima, Hiroshi*; Yoshizaki, Hiroaki*; Kaneno, Yasuyuki*; Semboshi, Satoshi*; Hori, Fuminobu*; Saito, Yuichi; Okamoto, Yoshihiro; Iwase, Akihiro*

Nuclear Instruments and Methods in Physics Research B, 372, p.72 - 77, 2016/04

https://doi.org/10.1016/j.nimb.2016.02.002

NiNb and NiTa intermetallic compounds, which show the complicated lattice structures were irradiated with 16 MeV Au ions at room temperature. The X-ray diffraction measurement revealed that the lattice structure of these intermetallic compounds changed from the ordered structures to the amorphous state by the ion irradiation. The irradiation-induced amorphization caused the increase in Vickers hardness. The result was compared with our previous results for NiAl and NiV, and was discussed in terms of the intrinsic lattice structures of the samples.

Source term estimation for the Fukushima Daiichi Nuclear Power Station accident by combined analysis of environmental monitoring and plant data through atmospheric dispersion simulation

Nagai, Haruyasu; Terada, Hiroaki; Chino, Masamichi; Katata, Genki; Mikami, Satoshi; Saito, Kimiaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4044 - 4052, 2015/08

JAEA has estimated the atmospheric releases of radionuclide during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric transport and deposition model (ATDM). To improve our source term, we are trying to develop more sophisticated estimation method and use new information from severe accident analysis and observation data. As the first step of new trial, we used Cs/Cs ratios of inventories in FNPS1 reactors Unit 1 to 3 and those in surface deposition. By considering temporal change in Cs/Cs ratio of released plume and ATDM simulations, spatial distribution of Cs/Cs ratio in surface deposition was explained. This result can be used to specify from which reactor the dominant release occurred for each time period, and consequently provide useful information to severe accident analysis for the FNPS1 case.

Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Homma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Takada, Shoji; Shinozaki, Masayuki

Nuclear Engineering and Design, 271, p.499 - 504, 2014/05

https://doi.org/10.1016/j.nucengdes.2013.12.022

The long term high temperature operation using HTTR was carried out to establish the technical basis of HTGR in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to demonstrate the stability of plant during long-term operation in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, Stability and reliability of the components and facility was demonstrated by evaluating the heat transfer performance of high temperature components, the performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the operation, the technical basis for the operation and maintenance technology of HTGRs was established.

The Influence on the vulcanized rubber physical properties by radiation grafting

Saito, Hiroaki*; Mizote, Norihito*; Ueki, Yuji; Seko, Noriaki

JAEA-Review 2013-059, JAEA Takasaki Annual Report 2012, P. 57, 2014/03

https://jopss.jaea.go.jp/pdfdata/JAEA-Review-2013-059.pdf

no abstracts in English