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Journal Articles

Study on sodium-water reaction jet evaluation model based on engineering approaches with particle method

Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*

Nihon Kikai Gakkai Rombunshu (Internet), 88(905), p.21-00310_1 - 21-00310_9, 2022/01

https://doi.org/10.1299/transjsme.21-00310

If a pressurized water/water-vapor leaks from a heat transfer tube in a steam generator (SG) in a sodium-cooled fast reactor (SFR), sodium-water reaction forms high-velocity, high-temperature, and corrosive jet. It would damage the other tubes and might propagate the tube failure in the SG. Thus, it is important to evaluate the effect of the tube failure propagation for safety assessment of SFR. The computational code LEAP-III can evaluate water leak rate during the tube failure propagation with short calculation time, since it consists of empirical formulae and one-dimensional equations of conservation. One of the empirical models, temperature distribution evaluation model, evaluates the temperature distribution in SG as circular arc isolines determined by experiments and preliminary analyses instead of complicated real distribution. In order to improve this model to get more realistic temperature distribution, we have developed the Lagrangian particle method based on engineering approaches. In this study, we have focused on evaluating gas flow in a tube bundle system, and constructed new models for the gas-particles behavior around a tube to evaluate void fraction distribution near the tube. Through the test analysis simulating one target tube system, we confirmed the capability of the models and next topic to improve the models.

Journal Articles

Concepts and basic designs of various nuclear fuels, 5; Fuels for high temperature gas-cooled reactor and molten salt reactor

Ueta, Shohei; Sasaki, Koei; Arita, Yuji*

Nihon Genshiryoku Gakkai-Shi ATOMO, 63(8), p.615 - 620, 2021/08

https://doi.org/10.3327/jaesjb.63.8_615

no abstracts in English

Journal Articles

Study on sodium-water reaction jet evaluation model based on engineering approaches with particle method

Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

If a pressurized water/water-vapor leaks from a heat transfer tube in a steam generator (SG) in a sodium-cooled fast reactor (SFR), sodium-water reaction forms high-velocity, high-temperature, and corrosive jet. It would damage the other tubes and might propagate the tube failure in the SG. Thus, it is important to evaluate the effect of the tube failure propagation for safety assessment of SFR. The computational code LEAP-III can evaluate water leak rate during the tube failure propagation with short calculation time, since it consists of empirical formulae and one-dimensional equations of conservation. One of the empirical models, temperature distribution evaluation model, evaluates the temperature distribution in SG as circular arc isolines determined by experiments and preliminary analyses instead of complicated real distribution. In order to improve this model to get more realistic temperature distribution, we have developed the Lagrangian particle method based on engineering approaches. In this study, we have focused on evaluating gas flow in a tube bundle system, and constructed new models for the gas-particles behavior around a tube to evaluate void fraction distribution near the tube. Through the test analysis simulating one target tube system, we confirmed the capability of the models and next topic to improve the models.

JAEA Reports

Confirmation tests for Warm Pre-stress (WPS) effect in reactor pressure vessel steel (Contract research)

Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka

JAEA-Research 2017-018, 122 Pages, 2018/03

JAEA-Research-2017-018.pdf:44.03MB

Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.

Journal Articles

Nuclear heat supply fluctuation test by non-nuclear heating using HTTR

Takada, Shoji; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Inaba, Yoshitomo; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, the helium gas temperature was heated up to 120 $^{circ}$ C. A sufficiently high temperature disturbance was imposed on the reactor inlet temperature. It was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure, which was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.

Journal Articles

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1 $^{circ}$ C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

JAEA Reports

Research and development plan for advanced high temperature gas cooled reactor fuels and graphite components (Contract research)

Sawa, Kazuhiro; Ueta, Shohei; Shibata, Taiju; Sumita, Junya; Ohashi, Jumpei; Tochio, Daisuke

JAERI-Tech 2005-024, 34 Pages, 2005/03

JAERI-Tech-2005-024.pdf:2.15MB

The Very-High-Temperature Reactor (VHTR) is one of the strong candidates for the Generation IV Nuclear Energy System. JAERI has developed Zirconium carbide (ZrC)-coated fuel particle and ZrC coating layer is expected to maintain its intactness under higher temperature and burn-up comparing conventional SiC-coating layer. JAERI carries out (1) ZrC-coating process development by large-scale coater, (2) inspection method development and (3) irradiation test and post irradiation experiment of ZrC coated particles. Also, JAERI carries out reactivity insertion tests to clarify the coating failure mechanism and tries to increase allowable temperature limit in case of reactivity insertion accident. Furthermore, JAERI develops non-destructive evaluation methods for mechanical properties of graphite components by ultrasonic testing and micro-indentation technique. This report describes these research and development plan and results of FY 2004 as a MEXT contact research.

Journal Articles