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Journal Articles

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JAEA Reports

Report of summer holiday practical training 2018; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design

Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.

JAEA-Technology 2019-008, 12 Pages, 2019/07

JAEA-Technology-2019-008.pdf:2.37MB

As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.

JAEA Reports

Data report of ROSA/LSTF experiment SB-SG-10; Recovery actions from multiple steam generator tube rupture accident

Takeda, Takeshi

JAEA-Data/Code 2018-004, 64 Pages, 2018/03

JAEA-Data-Code-2018-004.pdf:3.33MB

Experiment SB-SG-10 was conducted on November 17, 1992 using LSTF. Experiment simulated recovery actions from multiple steam generator (SG) tube rupture accident in PWR. Primary pressure was kept higher than broken SG secondary-side pressure due to coolant injection from high pressure injection (HPI) system into cold and hot legs even after start of full opening of intact SG relief valve (RV). Full opening of power-operated relief valve (PORV) in pressurizer (PZR) resulted in pressure equalization between primary and broken SG systems as well as PZR liquid level recovery. Broken SG RV opened once after start of intact SG RV full opening. Core was filled with saturated or subcooled liquid through experiment. Significant natural circulation prevailed in intact loop after start of intact SG RV full opening. Significant thermal stratification appeared in hot legs especially during time period of HPI coolant injection into hot legs.

Journal Articles

Loss of core cooling test with one cooling line inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji

Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1$$^{circ}$$C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15$$^{circ}$$C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

JAEA Reports

Data report of ROSA/LSTF experiment TR-LF-07; Loss-of-feedwater transient with primary feed-and-bleed operation

Takeda, Takeshi

JAEA-Data/Code 2016-004, 59 Pages, 2016/07

JAEA-Data-Code-2016-004.pdf:3.34MB

The TR-LF-07 test simulated a loss-of-feedwater transient in a PWR. A SI signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The primary feed-and-bleed operation was effective to core cooling because of no core uncovery.

Journal Articles

Loss of core cooling test without one cooling line in Vessel Cooling System (VCS) of High Temperature engineering Test Reactor (HTTR)

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1 degree C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15 degree C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

Journal Articles

Development of simple method to incorporate out-of-core cooling effect on thorium conversion in multi-pass fueled reactor and investigation on characteristics of the effect

Fukaya, Yuji

Annals of Nuclear Energy, 81, p.301 - 305, 2015/07

 Times Cited Count:1 Percentile:86.4(Nuclear Science & Technology)

Development of a simple method to incorporate the out-of-core cooling effect on the thorium conversion in multi-pass fueled reactors and investigation on characteristics of the effect have been performed. For multi-pass fueled reactors, such as Molten Salt Breeder Reactor (MSBR) and Pebble-Bed Modular Reactor (PBMR), fuel moves in the core and exits from the core. The nuclides decay also out of the core, and it should be also considered if it is important for core characteristics. In the present study, $$^{233}$$Pa is considered to evaluate the thorium conversion accurately. To take the effect into account, in the present study, an effective decay constant is proposed to make equilibrium concentration of $$^{233}$$Pa without out-of-core cooling equal to that of out-of-core cooling. With the effective decay constant, the out-of-core cooling effect can be incorporated even with the code system using macroscopic cross sections generated by cell burn-up calculations without any code modification. In addition, the characteristic of out-of-core cooling effect for the thorium conversion is evaluated for thorium fueled reactors of MSBR and PBMR. It is concluded that the out-of-core cooling effect is suitable for MSBR to enhance thorium conversion because of the fast flow rate of fuel salt. On the other hand, the effect is not important and not realistic to employ for PBMR because the in-core residence time of approximately 100 days is longer than the half-life of $$^{233}$$Pa of 27.0 days, and the effect cannot improve the conversion ratio drastically.

Journal Articles

Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01

 Times Cited Count:10 Percentile:38.8(Nuclear Science & Technology)

Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.

JAEA Reports

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAERI-Research 2005-014, 170 Pages, 2005/06

JAERI-Research-2005-014.pdf:7.64MB

A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.

Journal Articles

Heat removal performance of auxiliary cooling system for the High Temperature Engineering Test Reactor during scrams

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki*

Annals of Nuclear Energy, 30(7), p.811 - 830, 2003/05

 Times Cited Count:1 Percentile:88.19(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Safety shutdown of the High Temperature Engineering Test Reactor during loss of off-site electric power simulation test

Takeda, Takeshi; Nakagawa, Shigeaki; Homma, Fumitaka*; Takada, Eiji*; Fujimoto, Nozomu

Journal of Nuclear Science and Technology, 39(9), p.986 - 995, 2002/09

 Times Cited Count:2 Percentile:82.04(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Experimental result of BWR post-CHF tests; Critical heat flux and post-CHF heat transfer coefficient (Contract research)

Iguchi, Tadashi; Iwaki, Chikako*; Anoda, Yoshinari

JAERI-Research 2001-060, 91 Pages, 2002/02

JAERI-Research-2001-060.pdf:6.34MB

no abstracts in English

JAEA Reports

ROSA/LSTF experiment report for RUN SB-CL-24; Repeated core heatup phenomena during 0.5% cold leg break

Suzuki, Mitsuhiro; Anoda, Yoshinari

JAERI-Tech 2000-016, p.173 - 0, 2000/03

JAERI-Tech-2000-016.pdf:7.25MB

no abstracts in English

Journal Articles

Experimental and analytical results for cooling performance of HTTR core bottom structure using HENDEL-T$$_{2}$$

Ioka, Ikuo; Inagaki, Yoshiyuki; Suzuki, Kunihiro; Kunitomi, Kazuhiko;

Nippon Genshiryoku Gakkai-Shi, 37(3), p.217 - 227, 1995/00

 Times Cited Count:1 Percentile:81.61(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Quasi-static core liquid level depression and long-term core uncovery during a PWR LOCA

Kukita, Yutaka; R.R.Schultz*; Nakamura, Hideo; Katayama, Jiro*

Nucl. Saf., 34(1), p.33 - 48, 1993/01

no abstracts in English

Journal Articles

Journal Articles

A Concept of a passive safety light water reactor system requiring reduced maintenance efforts

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Watanabe, Hironori

Transactions of the American Nuclear Society, 69, p.539 - 540, 1993/00

no abstracts in English

Journal Articles

PWR cold-leg small-break LOCA with total HPI failure; Effect of break area on core dryout and intentional depressurization for prevention of excess core dryout

Kumamaru, Hiroshige; Kukita, Yutaka

Journal of Nuclear Science and Technology, 29(12), p.1162 - 1172, 1992/12

no abstracts in English

Journal Articles

Application of new coolant inventory tracking method to PWR small break LOCA simulation experiments at ROSA-IV/LSTF

Suzuki, Mitsuhiro

Journal of Nuclear Science and Technology, 29(6), p.547 - 558, 1992/06

no abstracts in English

44 (Records 1-20 displayed on this page)