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Nagase, Fumihisa
Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)The fracture threshold of the fuel decreases if the oxidized Zr alloy cladding is strongly constrained by the spacer grid during quenching in a loss-of-coolant accident. Therefore, the estimation of realistic levels of the axial constraint has been a subject of significant interest on fuel safety. In this study, a test assembly consisting of a PWR-type simulated fuel segment and a 3
3 grid piece was heated in steam, cooled, and quenched, and the axial constraint force on the fuel segment was measured. The constraint force of the Zircaloy grid gradually decreased with temperature. Once the Zircaloy grid was heated to 
1060 K, the reduced constraint force had difficulty recovering, and thus the maximum constraint force during cooling and quenching was 
10 N. The constraint force was clearly reduced at 1070 K during the tests with the Inconel grid. However, the reduced constraint force partially recovered during cooling. As a result, the maximum constraint force during cooling and quenching was 20 to 50 N for the Inconel grid. In conclusion, oxidation, ballooning, rupture, or eutectic formation would not generally cause an extremely strong constraint, as predicted by previous studies, at the grid position.
Ozawa, Masaaki*; Amaya, Masaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12
no abstracts in English
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*
Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12
Times Cited Count：15 Percentile：60.5(Physics, Fluids & Plasmas)Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09
Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.
Takamatsu, Kuniyoshi; Hu, R.*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (
C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi
HPR-362, Vol.2, 12 Pages, 2004/05
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.
Department of Hot Laboratories
JAERI-Review 2003-038, 106 Pages, 2003/12
JAERI-Review-2003-038.pdf:9.36MB
no abstracts in English
Iwai, Yasunori; Nakamura, Hirofumi; Konishi, Satoshi; Nishi, Masataka; Willms, R. S.*
Fusion Science and Technology, 41(3), p.668 - 672, 2002/05
no abstracts in English
Iguchi, Tadashi
JAERI-Research 98-054, 216 Pages, 1998/10
JAERI-Research-98-054.pdf:7.94MB
no abstracts in English
Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki; *; *
JAERI-Research 98-009, 44 Pages, 1998/03
JAERI-Research-98-009.pdf:2.49MB
no abstracts in English
Aoto, Kazumi; ; ; ; ; Hirakawa, Yasushi
PNC TN9420 97-007, 786 Pages, 1997/06
PNC-TN9420-97-007.pdf:311.86MB
The results of the research on the damaged thermocouple well which caused the sodium leak accident at the outlet of the C-loop intermidiate heat exchanger (IHX) of the secondary heat transfer system of the prototype fast breeder reactor Monju are described in this report. A lot of tests, inspections, observations and measurements were carried out to confirm that the thermocouple well and its attachments to the pipe including welded part are normal by checking the possibility of weld failure or corrosion at the clearance which may cause the damage of the thermocouple well, and to get information of the dimensions relating the estimation of the leaked sodium volume and the leakage path, etc. These tests, etc., were performed for the thermocouple well except the fractured surface, the thermocouple well, the welded parts between the thermocouple well and the attachment, and between the attachment and the outlet pipe, etc., as written below. (1)Accurate measurement of the dimension. (2)Inspection to check the fixing condition between the thermocouple well and the attachment. (3)Measurement of the residual stress. (4)Non destructive testing at some points. (5)Chemical composition analysis. (6)Microscopic observation of metalogical structure at the welded part. (7)Hardness test. (8)Research on corrosion at the clearance. (9)Structure strength test of the thermocouple well. (10)Bending test of the thermocouple's sheath at high temperature.
Shimoyama, Kazuhito; Usami, Masayuki; Miyake, Osamu; ; ; Tanabe, Hiromi
PNC TN9450 97-007, 81 Pages, 1997/03
None
Iwata, Koji; *; *; *; *; *; *
PNC TN9410 97-042, 8 Pages, 1997/03
A design guide for flow-induced vibration of thermometer wells is proposed to prevent the recurrence of the failure of thermometer wells, which was the direct cause of the 1995 sodium leak incident of the secondary main piping of the prototype fast breeder reactor MONJU. As a supplement to the technical standards in force for MONJU, the design guide specifies the methods of evaluation and the design criteria on structural integrity against flow-induced vibration for thermometer wells, which are inserted into pipes of fast breeder reactors. The design guide is a PNC's (Power Reactor and Nuclear Fuel Development Corporation) internal guide for MONJU, which is to be used, with the permission of outside authorities, to confirm the integrity of the existing equipments as well as to make an improved design of thermometer wells. The proposed design guide was prepared by the Special Working Group on Thermometer Design Guide, organized in PNC during the period from May to November, 1996.
