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Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji
JAEA-Data/Code 2022-012, 270 Pages, 2023/03
JAEA-Data-Code-2022-012.pdf:38.25MB
In a light water reactor, which is a commercial nuclear power plant, a severe accident such as loss of cooling function in the reactor pressure vessel (RPV) and exposure of fuel rods due to a drop in the water level in the reactor can occur when a trouble like loss of all AC power occurs. In the event of such a severe accident, the RPV may be damaged due to in-vessel conditions (temperature, molten materials, etc.) and leakage of radioactive materials from the reactor may occur. Verification and estimation of the process of RPV damage, molten fuel debris spillage and expansion, etc. during accident progression will provide important information for decommissioning work. Possible causes of RPV failure include failure due to loads and restraints applied to the RPV substructure (mechanical failure), failure due to the current eutectic state of low-melting metals and high-melting oxides with the RPV bottom members (failure due to inter-material reactions), and failure near the melting point of the structural members at the RPV bottom. Among the failure factors, mechanical failure is verified by numerical analysis (thermal hydraulics and structural analysis). When conducting such a numerical analysis, the heat transfer properties (thermal conductivity, specific heat, density) and material properties (thermal conductivity, Young's modulus, Poisson's ratio, tensile, creep) of the materials (zirconium, boron carbide, stainless steel, nickel-based alloy, low alloy steel, etc.) constituting the RPV and in-core structures to near the melting point are required to evaluate the creep failure of the RPV. In this document, we compiled data on the properties of base materials up to the melting point of each material constituting the RPV and in-core structures, based on published literature. In addition, because welds exist in the RPV and in-core structures, the data on welds are also included in this report, although they are limited.
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
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.
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 1C. 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.
Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro
Nihon Genshiryoku Gakkai-Shi, 47(5), p.342 - 349, 2005/05
For the purpose to extend high temperature nuclear heat application, JAERI constructed the HTTR, High Temperature Engineering Test Reactor, and has carried out research and development of high temperature gas cooled reactor system aiming at high efficiency power generation and hydrogen production. This paper explains the history, main results, present status of research and development of HTTR project, international cooperation of research and development of HTGR and future plan aiming at development of Japanese original future HTGR-Hydrogen production system. This paper includes results from the study, which is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.
Inaba, Yoshitomo; Fumizawa, Motoo; Hishida, Makoto; Ogawa, Masuro; Aritomi, Masanori*; Kozaki, Yasutsugu*; *; *; *; *; et al.
JAERI-Tech 96-019, 122 Pages, 1996/05
no abstracts in English
Inaba, Yoshitomo; Fumizawa, Motoo; Hishida, Makoto; Ogawa, Masuro; ; Aritomi, Masanori*; Kozaki, Yasutsugu*; *; *; *; et al.
JAERI-Review 96-007, 87 Pages, 1996/05
JAERI-Review-96-007.pdf:2.66MB
no abstracts in English
Baba, Osamu
Sci. Technol. Jpn., 15(58), p.7 - 9, 1996/00
no abstracts in English
; Sato, Osamu; ; *
Nucl. Eng. Des., 136, p.211 - 217, 1992/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
; Sato, Osamu; ; *
Proc. for the Seminar on 10th Int. Conf. SMiRT, p.VI.6.1 - VI.6.8, 1989/00
no abstracts in English
Nakano, Hiroko; Takeuchi, Tomoaki; Hirota, Noriaki; Ide, Hiroshi; Hanawa, Masashige*; Tsuchiya, Kunihiko
no journal, ,
Since the accident at the Fukushima Dai-ichi Nuclear Power Plants (NPPs), it has started a research and development which corresponds to the provisions to monitor NPPs situations during a severe accident (SA). Considering that reactor vessel might be exposed to high radiation at high temperature under SA, difficulties may arrear in the measurement of temperature and pressure in the reactor vessel for long time. Thus it is necessary for the measurements of them to develop mineral insulated (MI) cables with radiation- and heat-resistant. In this study, the electric and corrosion properties of the MI cables were evaluated in the conditions that may simulate SA to develop MI cables that can be used even in SA environment. Based on the experiments above, the prospect is obtained that MI cable using nickel alloy as a sheath material can be proposed as a high-temperature MI cable that can be used even in SA environment.
