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

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

Proceedings of 8th International Conference on New Energy and Future Energy Systems (NEFES 2023) (Internet), p.27 - 34, 2023/00

https://doi.org/10.3233/ATDE231072

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.

Comparative study on annual Cs discharge rates after the Fukushima Dai-ichi Nuclear Power Plant accident from two distinct watershed simulation models

Kitamura, Akihiro; Imaizumi, Yoshitaka*; Yamaguchi, Masaaki; Yui, Mikazu; Suzuki, Noriyuki*; Hayashi, Seiji*

Kankyo Hoshano Josen Gakkai-Shi, 2(3), p.185 - 192, 2014/09

Annual discharge rates of radioactive cesium through selected rivers due to the Fukushima Dai-ichi Nuclear Power Plant accident were simulated by two different watershed models. One is the Soil and Cesium Transport, SACT, model which was developed by Japan Atomic Energy Agency and the other one is the Grid-Catchment Integrated Modeling System, G-CIEMS, which was developed by National Institute of Environmental Studies. We choose the Abukuma, the Ukedo, and the Niida rivers for the present study. Comparative results showed that while components and assumptions adopted in two models differ, both methods predicted the same order of magnitude estimates.

Elaboration of the system based code concept; Activities in JSME and ASME, 2; Development of evaluation tools based on LRFD

Machida, Hideo*; Asayama, Tai; Watanabe, Taigo*; Hojo, Kiminobu*; Hayashi, Masaaki*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

https://doi.org/10.1115/ICONE22-30525

In the System Based Code (SBC), a reliability target is defined according to the importance of components (risk and/or failure probability), and grade of material, design, manufacture and maintenance are chosen to satisfy the reliability target. Therefore, reliability evaluation of components plays the important role of the concept of SBC. Until now, the LRFD methods were developed for burst due to internal pressure, plastic collapse due to membrane and bending stress, fatigue, limit load assessment of flawed pipe, and buckling of thin wall cylinder. This paper describes the action plans of development of the reliability assessment methods and an examination results up to date.

Plasticity collapse strength of stainless piping subjected to combined tension and bending

Matsubara, Masaaki*; Hayashi, Tatsuya*; Suzuki, Masato*; Shiraishi, Taisuke*; Sakamoto, Kenji*; Wakai, Takashi

Nihon Kikai Gakkai M&M 2011 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), 2 Pages, 2011/07

In this study, we investigated the influence of two or more flaws on the collapse strength of austenitic stainless steel piping to improve the safety of a nuclear power plant. The multiple flaws may be more often than single flaw. Therefore, it is important to evaluate the collapse strength of the piping with multiple flaws. Piping with a single flaw and multiple flaws were used in the examination. Collapse strength of the piping is verified under the combination condition of tension and bending. It aims to model from multiple flaws to the single flaw for simplification of the. As a result, (1) It is possible to evaluate the collapse strength by identifying the multiple flaws as the single flaw conservatively. (2) On each load condition, the parallel notches on piping can be accounted to be the single notch.

Overview of national centralized tokamak program; Mission, design and strategy to contribute ITER and DEMO

Ninomiya, Hiromasa; Akiba, Masato; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko; Hosogane, Nobuyuki; Ikeda, Yoshitaka; Inoue, Nobuyuki; et al.

Journal of the Korean Physical Society, 49, p.S428 - S432, 2006/12

To contribute DEMO and ITER, the design to modify the present JT-60U into superconducting coil machine, named National Centralized Tokamak (NCT), is being progressed under nationwide collaborations in Japan. Mission, design and strategy of this NCT program is summarized.

Engineering design and control scenario for steady-state high-beta operation in national centralized tokamak

Tsuchiya, Katsuhiko; Akiba, Masato; Azechi, Hiroshi*; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hashizume, Hidetoshi*; Hayashi, Nobuhiko; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 81(8-14), p.1599 - 1605, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.08.066

no abstracts in English

Study on Mixing Phenomena in T-Pipe Junction; Temperature Measurement Test in Pipe by Liquid Crystal Sheet

Tanaka, Masaaki; Kawashima, Shigeyo*; Igarashi, Minoru; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki

JNC TN9400 2003-117, 65 Pages, 2004/03

JNC-TN9400-2003-117.pdf:3.49MB

Temperature fluctuation due to mixing of hot and cold fluids gives thermal fatigue to the structure (thermal striping phenomena).Investigation of this phenomenon is significant for the safety of a fast breeder reactor, which uses liquid metal as a coolant. In Japan Nuclear Cycle Development Institute, experiments and numerical analyses have been carried out to understand this phenomenon and also to construct the evaluation rule, which can be applied to the design. A water experiment of fluid mixing in T-pipe with long cycle fluctuation (WATLON),which notices thermal striping phenomena in the T-pipe junction, is performed to investigate the key factor of mixing phenomena by reason of long cycle fluctuation observed in a plant. By the former visualization test, it was showed that the flow pattern of branch pipe jet could be classified into (A) impinging jet, (B) deflecting jet (C) re-attachment jet and (D) wall jet according to the inflow condition. It was confirmed that the each jet pattern could be predicted by the momentum ratio of the each piping fluid. In this study, a thermo-chromic liquid crystal sheet was put on the inner wall in the main pipe, and temperature field on the wall surface was visualized. We established a new method to convert the color image data to temperature data. And measurement uncertainty of this method was evaluated + and - about 2.0 [deg-C], using by the typical picture in the temperature calibration test. From the temperature fluctuation visualization test by liquid crystal sheet, the cold spot was formed in just downstream region from the outlet of the branch pipe in the cases of the wall jet and impinging jet. Since this cold spot moved in time, high value of temperature fluctuation intensity was shown around the cold spot. And the validity of this method was shown from the comparison of the thermocouple data installed in a wall surface with the temperature conversion result.

Study on Mixing Phenomena in T-pipe Junction; Clarification of fluid mixing phenomena and temperature fluctuation transfer from fluid to structure

Igarashi, Minoru; Tanaka, Masaaki; Kimura, Nobuyuki; Nakane, Shigeru*; Kawashima, Shigeyo*; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki

JNC TN9400 2003-092, 100 Pages, 2003/11

JNC-TN9400-2003-092.pdf:5.1MB

A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. Measurement of flow velocity using particle image velocimetry and temperature measurement were carried out. Normalized power spectrum density of temperature fluctuation had same profile, when the momentum ratio of the main and branch pips is the same. From the velocity measurement test, when the momentum ratio is the same, flow pattern at mixing region shows the alomost same tendency. Temperature transfer characteristics from fluid to structure can be estimated by a constant heat transfer coefficient in time.

Present status of JT-60SU design

; Ushigusa, Kenkichi; Kikuchi, Mitsuru; Nagashima, Keisuke; Neyatani, Yuzuru; Miya, Naoyuki; ; Takahashi, Yoshikazu; Hayashi, Takumi; Kuriyama, Masaaki; et al.

Proceedings of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 1, p.233 - 236, 1998/00

no abstracts in English

Large-scaled thermohydraulic tests plan for cooling systems in fast reactors; Experimental models of reactor vessel and the primary cooling system

Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.

PNC TN9410 96-279, 51 Pages, 1996/08

PNC-TN9410-96-279.pdf:2.92MB

Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.