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JAEA Reports

Data report of ROSA/LSTF experiment SB-SL-01; Main steam line break accident

Takeda, Takeshi

JAEA-Data/Code 2020-019, 58 Pages, 2021/01

JAEA-Data-Code-2020-019.pdf:3.85MB

An experiment denoted as SB-SL-01 was conducted on March 27, 1990 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment SB-SL-01 simulated a main steam line break (MSLB) accident in a pressurized water reactor (PWR). The test assumptions were made such as auxiliary feedwater (AFW) injection into secondary-side of both steam generators (SGs) and coolant injection from high pressure injection (HPI) system of emergency core cooling system into cold legs in both loops. The MSLB led to a fast depressurization of broken SG, which caused a decrease in the broken SG secondary-side wide-range liquid level. The broken SG secondary-side wide-range liquid level recovered because of the AFW injection into the broken SG secondary-side. The primary pressure temporarily decreased a little just after the MSLB, and increased up to 16.1 MPa following the closure of the SG main steam isolation valves. Coolant was manually injected from the HPI system into cold legs in both loops a few minutes after the primary pressure reduced to below 10 MPa. The primary pressure raised due to the HPI coolant injection, but was kept at less than 16.2 MPa by fully opening a power-operated relief valve of pressurizer. The core was filled with subcooled liquid through the experiment. Thermal stratification was seen in intact loop cold leg during the HPI coolant injection owing to the flow stagnation. On the other hand, significant natural circulation prevailed in broken loop. When the continuous core cooling was ensured by the successive coolant injection from the HPI system, the experiment was terminated. The experimental data obtained would be useful to consider recovery actions and procedures in the multiple fault accident with the MSLB of PWR. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-SL-01.

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.

JAEA Reports

Effects of temperature history during cooling process on cladding ductility reduction under lost of coolant accident conditions

Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi

JAERI-Research 2005-020, 40 Pages, 2005/09

JAERI-Research-2005-020.pdf:4.63MB

In order to investigate effects of quenching temperature and cooling rate before quench on cladding ductility reduction under LOCA conditions, samples cut from non-irradiated 17$$times$$17-type Zircaloy-4 cladding tubes for PWRs were oxidized in steam at 1373 and 1473 K, cooled at 2 to 7 K/s, and quenched at 1073 to 1373 K. The quenched samples were subjected to ring compression test, microstructure observation, and Vickers hardness test. Quenching temperature decrease obviously increased area fraction of $$alpha$$ phase in the radial cross section of the cladding, and reduced cladding ductility. Slow-cooling rate decrease increased unit size and hardness of precipitated $$alpha$$ phase, while $$alpha$$ phase area fraction and cladding ductility were not significantly changed. $$alpha$$ phase is harder than the surrounding region in the metallic layer and has higher oxygen content, indicating its low ductility. Consequently, increase in the area fraction in the cladding is a main cause of the reduction in cladding ductility with decrease in the quenching temperature.

JAEA Reports

Verification of HTTR hydrogen production system analysis code using experimental data of mock-up model test facility with a full-scale reaction tube; Cooling system of the secondary helium gas using steam generator and radiator (Contract research)

Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Maeda, Yukimasa; Takeda, Tetsuaki; Nishihara, Tetsuo; Inagaki, Yoshiyuki

JAERI-Tech 2005-014, 89 Pages, 2005/03

JAERI-Tech-2005-014.pdf:7.25MB

In a hydrogen production system using HTTR, it is required to control a secondary helium gas temperature within an allowable value at an intermediate heat exchanger (IHX) inlet to prevent a reactor scram. To mitigate thermal disturbance of the secondary helium gas caused by the hydrogen production system, a cooling system of the secondary helium gas using a steam generator(SG) and a radiator will be installed at the downstream of the chemical reactor. In order to verify a numerical analysis code of the cooling system, numerical analysis has been conducted. The pressure controllability in SG is highly affected by the heat transfer characteristics of air which flows outside of the heat exchanger tube of the radiator. In order to verify a numerical analysis code of the cooling system, the heat transfer characteristics of air has been investigated with experimental results of a mock-up model test. It was confirmed that numerical analysis results were agreed well with experimental results, and the analysis code was successfully verified.

Journal Articles

Safety analysis of ITER test blanket module for water cooled blanket with pebble bed breeder

Enoeda, Mikio; Kuroda, Toshimasa*; Moriyama, Koichi*; Ohara, Yoshihiro

Journal of Nuclear Science and Technology, 38(11), p.921 - 929, 2001/11

 Times Cited Count:2 Percentile:20.38(Nuclear Science & Technology)

Test module testing in ITER is one of the most important mile-stone for development of the DEMO blanket. In the design of test modules in ITER, it is very important to show that test modules do not cause additional safety concern to ITER. This work has been performed for the evaluation of the substantial safety of Test Module of Water Cooled Solid Blanket, which is the current candidate blanket for the DEMO blanket in Japan. Major issues of the evaluation were establishment of post accident cooling in TM, hydrogen gas generation by Be-steam reaction, and pressure increase and spilled water amount by Loss of Coolant Accident (LOCA) event. The evaluation was performed to derive the upper bound of consequences in significant events, of which scenario can be assumed by the similarity of the safety analysis of Shielding Blanket.

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

Nucl. Eng. Des., 144, p.257 - 268, 1993/00

 Times Cited Count:1 Percentile:19.22(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions for PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

ANP 92: Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants,Vol. 3, p.24.4-1 - 24.4-7, 1992/00

no abstracts in English

JAEA Reports

Evaluation report on CCTF core-II reflood test C2-5(Run 63); Effect of decay heat level on PWR reflood phenomena

Iguchi, Tadashi; ; Okabe, Kazuharu*; Sugimoto, Jun; ; Okubo, Tsutomu; Murao, Yoshio

JAERI-M 91-174, 98 Pages, 1991/10

JAERI-M-91-174.pdf:2.31MB

no abstracts in English

JAEA Reports

Evaluation report on CCTF core-II reflood test C2-1(Run 55); Effect of pressure on reflood phenomena

Iguchi, Tadashi; Sugimoto, Jun; ; Okubo, Tsutomu; Murao, Yoshio

JAERI-M 91-173, 94 Pages, 1991/10

JAERI-M-91-173.pdf:2.01MB

no abstracts in English

Journal Articles

Post-dryout heat transfer of steam-water two-phase flow in rod bundle under high-pressure and low-flow conditions

Kumamaru, Hiroshige; Kukita, Yutaka

ANS Proc. 1991 National Heat Transfer Conf., Vol. 5, p.22 - 29, 1991/00

no abstracts in English

Journal Articles

Method of estimating liquid velocity in a hot leg during reflux cooling of natural circulation in a pressurized water reactor

Soda, Kunihisa

Journal of Nuclear Science and Technology, 19(10), p.813 - 820, 1982/00

 Times Cited Count:4 Percentile:48.3(Nuclear Science & Technology)

no abstracts in English

12 (Records 1-12 displayed on this page)
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