Evaluation of thermal neutron scattering law of nuclear-grade isotropic graphite

Nakayama, Shinsuke; Iwamoto, Osamu; Kimura, Atsushi

EPJ Web of Conferences, 294, p.07001_1 - 07001_6, 2024/04

https://doi.org/10.1051/epjconf/202429407001

Graphite is a candidate of moderator in innovative nuclear reactors such as molten salt reactors. Scattering of thermal neutrons by the moderator material has a significant impact on the reactor core design. To contribute to the development of innovative nuclear reactors, an evaluation method of thermal neutron scattering law for reactor grade graphite was studied. The inelastic scattering component due to lattice vibration was evaluated based on the phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The coherent elastic scattering component due to crystal structure was evaluated based on neutron transmission and scattering experiments recently performed in the J-PARC/MLF facility. In comparison with the neutron transmission experiments, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important. Based on the above methods, thermal neutron scattering law data for reactor-grade graphite at room temperature were evaluated.

Failure probability evaluation for steam generator tubes with wall-thinning

Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03

The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.

The BCC FCC hierarchical martensite transformation under dynamic impact in FeMnAlNiTi alloy

Li, C.*; Fang, W.*; Yu, H. Y.*; Peng, T.*; Yao, Z. T.*; Liu, W. G.*; Zhang, X.*; Xu, P. G.; Yin, F.*

Materials Science & Engineering A, 892, p.146096_1 - 146096_11, 2024/02

https://doi.org/10.1016/j.msea.2024.146096

Technology information on High Temperature Gas-cooled Reactor (HTGR)

HTGR Design Group, HTGR Project Management Office

JAEA-Technology 2023-019, 39 Pages, 2024/01

JAEA-Technology-2023-019.pdf:1.34MB

In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.

Linearization of thermal neutron scattering cross section to optimize the number of energy grid points

Tada, Kenichi

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10

The number of energy grids of the thermal neutron scattering law data has a large impact on the data size of a cross section file of continuous energy Monte Carlo calculation codes. The optimization of the number of energy grids is an effective way to reduce the data size. This study developed the linearization method of the thermal neutron scattering cross section to optimize the number of energy grids and the linearization function was implemented in the nuclear data processing code FRENDY. The linearization process which is used in the resonance reconstruction and the Doppler broadening was adopted. The criticality benchmarks which use ZrH as the moderator were calculated to estimate the impact of the difference of the energy grids on neutronics calculations. The calculation results indicate that the linearization of the thermal neutron scattering cross section improves the prediction accuracy of neutronics calculations.

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger; Project overview

Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.

JSME series in thermal and nuclear power generation Vol.3 (Sodium-cooled fast reactor development; R&Ds on thermal-hydraulics and safety assessment towards social implementation)

Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.

Effect of decay heat on pyrochemical reprocessing of minor actinide transmutation nitride fuels

Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08

https://doi.org/10.3327/taesj.J22.009

The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.

Contribution to the carbon neutrality by advanced reactor coexisting with renewable energy system

Ohno, Shuji; Yamano, Hidemasa

Nihon Genshiryoku Gakkai-Shi ATOMO, 65(7), p.438 - 442, 2023/07

https://doi.org/10.3327/jaesjb.65.7_438

This report mainly introduces international/domestic development trends of advanced reactor systems that can coexist with renewable energy in addition to describing the importance of dispatchable energy for the extension of variable renewable energy.

Extreme phonon anharmonicity underpins superionic diffusion and ultralow thermal conductivity in argyrodite AgSnSe

Ren, Q.*; Gupta, M. K.*; Jin, M.*; Ding, J.*; Wu, J.*; Chen, Z.*; Lin, S.*; Fabelo, O.*; Rodriguez-Velamazan, J. A.*; Kofu, Maiko; et al.

Nature Materials, 22, p.999 - 1006, 2023/05

https://doi.org/10.1038/s41563-023-01560-x