An Experimental study related to axial constraint of fuel rod under LOCA conditions

Nagase, Fumihisa

Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06

https://doi.org/10.1016/j.anucene.2022.109052

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 33 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.

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*

Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12

https://doi.org/10.1088/0029-5515/55/12/123008

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.

New reactor cavity cooling system using novel shape for HTGRs and VHTRs

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.

LOCA and RIA studies at JAERI

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.

Study on thermal-hydraulics during a PWR reflood phase

Iguchi, Tadashi

JAERI-Research 98-054, 216 Pages, 1998/10

JAERI-Research-98-054.pdf:7.94MB

no abstracts in English

Analyses for passive safety of fusion reactor during Ex-vessel loss of coolant accident

Honda, Takuro*; *; *; *; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki

Journal of Nuclear Science and Technology, 32(4), p.265 - 274, 1995/04

https://doi.org/10.3327/jnst.32.265

no abstracts in English

Analytical study on characteristics of falling water limitation in countercurrent two-phase flow in vertical annular channels

Sudo, Yukio

Nihon Kikai Gakkai Rombunshu, B, 60(576), p.2888 - 2894, 1994/08

https://doi.org/10.1299/kikaib.60.2888

no abstracts in English

Analytical study on mechanism of countercurrent flow limitation in vertical rectangular channels

Sudo, Yukio

Nihon Kikai Gakkai Rombunshu, B, 60(574), p.2176 - 2182, 1994/06

https://doi.org/10.1299/kikaib.60.2176

no abstracts in English

Limitation of countercurrent gas-liquid flow in horizontal flow path connected to inolined flow path; Prediction of gas velocity on initiation of liquid penetration

;

Nihon Kikai Gakkai Rombunshu, B, 53(490), p.1685 - 1690, 1987/00

no abstracts in English

Mechanism of falling water limitation under counter-current flow through single vertical flow path

Sudo, Yukio;

Nihon Kikai Gakkai Rombunshu, B, 49(444), p.1685 - 1694, 1983/00

https://doi.org/10.1299/kikaib.49.1685

no abstracts in English