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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

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.

Journal Articles

Evaluation of heat removal during the failure of the core cooling for new critical assembly

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

In order to investigate the basic neutronics characteristics of the accelerator-driven subcritical system (ADS), JAEA has a plan to construct a new critical assembly in the J-PARC project, Transmutation Physics Experimental Facility (TEF-P). This study aims to evaluate the natural cooling characteristics of TEF-P core which has large decay heat by minor actinide (MA) fuel, and to achieve a design that does not damage the core and the fuels during the failure of the core cooling system. In the evaluation of the TEF-P core temperature, empty rectangular lattice tube outer of the core has a significant effect on the heat transfer characteristics. The experiments by using the mockup device were performed to validate the heat transfer coefficient and experimental results were obtained. By using the obtained experimental results, the three-dimensional heat transfer analysis of TEF-P core were performed, and the maximum core temperature was obtained, 294$$^{circ}$$C. This result shows TEF-P core temperature would be less than 327$$^{circ}$$C that the design criterion of temperature.

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

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

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 verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. 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. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, 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 1$$^{circ}$$C. 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.

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

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

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 1$$^{circ}$$C. 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.

JAEA Reports

Evaluation of heat exchange performance for intermediate heat exchanger in HTTR

Tochio, Daisuke; Nakagawa, Shigeaki

JAERI-Tech 2005-040, 39 Pages, 2005/07

JAERI-Tech-2005-040.pdf:1.88MB

In High Temperature Engineering Test Reactor (HTTR) of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary pressurized water cooler and the intermediate heat exchanger. Heat exchangers in main cooling system of HTTR should satisfy two conditions, achievement of reactor coolant outlet temperature 850 $$^{circ}$$C/950 $$^{circ}$$C and removal of reactor generated heat 30 MW. That is, heat exchange performance should be ensured as that in heat exchanger designing. In this report, heat exchange performance for Intermediate heat exchanger (IHX) in main cooling system is evaluated with rise-to-power-up test and in-service operation data. Moreover, the applicability of IHX thermal-hydraulic design method is discussed with comparison of evaluated data with designed value.

Journal Articles

Heat exchanger performance in main cooling system on high temperature test operation at High Temperature Gas-Cooled Reactor 'HTTR'

Tochio, Daisuke; Nakagawa, Shigeaki; Furusawa, Takayuki*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.147 - 155, 2005/06

High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at JAERI achieved the reactor outlet coolant temperature of 950$$^{circ}$$C for the first time in the world at Apr. 19, 2004. To remove of generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value.

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.

JAEA Reports

Management techniques of the JRR-4 heat exchanger

Horiguchi, Hironori; Oyama, Koji; Ishikuro, Yasuhiro; Hirane, Nobuhiko; Ito, Kazuhiro; Kameyama, Iwao

JAERI-Tech 2005-001, 38 Pages, 2005/02

JAERI-Tech-2005-001.pdf:2.79MB

After JRR-4 heat exchanger was renewed in made of stainless steel from carbon steel, it was examined how to manage the heat exchanger. The main subject is the cleaning technology of the heat exchanger. The recovery of old heat exchanger cooling performance has been by only chemical cleaning. Now we use chemical and dry cleaning as a new technique. It helps prevent of corrosions of secondary pipes and reduce of management costs. This report describes the performance management and cleaning technology of the JRR-4 heat exchanger and the management of the JRR-4 coolant.

Journal Articles

Performance test of HTTR

Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi

Nuclear Engineering and Design, 233(1-3), p.291 - 300, 2004/10

 Times Cited Count:6 Percentile:42.52(Nuclear Science & Technology)

The High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 850$$^{circ}$$C on December 7, 2001 during the "rise-to-power tests". Two kinds of tests were carried out during the "rise-to-power tests". One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, control system. From the test results of the "rise-to-power tests" up to 30MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely.

JAEA Reports

Report on investigation of cause of crack at instrumentation pipe in JMTR; Results of vibration and stress analysis

Hanawa, Satoshi; Tachibana, Yukio; Iyoku, Tatsuo; Ishihara, Masahiro; Ito, Haruhiko

JAERI-Tech 2003-064, 25 Pages, 2003/07

JAERI-Tech-2003-064.pdf:2.84MB

On the 147cycle operation, the water leakage was found at the pressure instrumentation pipe which is attached to the exit pipe of No.1 charge pump of the purification system of primary cooling system at JMTR in the Oarai establishment, JAERI. Then JMTR was shutted down manually on December 10th. It was predicted that the crack on the pressure instrumentation pipe was initiated and propagated by the cyclic load which was caused by the charge pump. Therefore, vibration and stress analyses of pressure instrumentation pipe were performed. From the vibration analysis, the natural frequency of the pressure instrumentation pipe of No.1 charge pump is between 53$$sim$$58Hz, which is close to the resonance frequency of 50Hz. From the stress analysis results, total stress generated on the pressure instrumentation pipe is 112.2MPa at the natural frequency of 53Hz and 74.2Mpa at 58Hz. It was found that the stress of 112.2MPa is close to the fatigue limit of used materials.

JAEA Reports

Setup of the cryogenic cooling system on BL11XU at SPring-8

Kiriyama, Koji*; Shiwaku, Hideaki; Tozawa, Kazukiyo*

JAERI-Tech 2003-061, 21 Pages, 2003/07

JAERI-Tech-2003-061.pdf:3.25MB

Cryogenically cooled monochromator equipped with circulated-liquid-nitrogen cooling system has been developed on JAERI beam-line, BL11XU, at SPring-8. This cooling system has improved the performance of X-ray reflection. The system has difficulty for the operation because of a lack of the manual. So, it is required to establish a systematic guide of the system. Here we report a manual with fruitful know-how to operate the cryogenic cooling system on BL11XU safely and easily. The new manual is wholly tabulated and shows comments, instructions, and valve conditions in all one-step. The operator, by using this manual, can manipulate the system with no misoperation.

JAEA Reports

Report on investigation of cause of crack at instrumentation pipe in JMTR; Data book on examination of pressure instrumentation pipe at JMTR hot laboratory

Working Group for Investigation of Cause of Crack Initiation

JAERI-Tech 2003-060, 183 Pages, 2003/07

JAERI-Tech-2003-060.pdf:55.37MB

On December 10, 2002, the water leakage was found at the pressure instrumentation pipe attached to the exit pipe of No.1 charging pump of the purification system of a primary cooling system at JMTR, and the cracks were detected on the pressure instrumentation pipe by the visual observation. The Investigation Committee for Water Leakage from Instrumentation Pipe in JMTR was established and organized by specialists from inside and outside JAERI on December 16. In order to investigate the cause of crack initiation at the pressure instrumentation pipe, the Working Group was organized in the Department of JMTR. Visual inspection, fractgraphy test, metallographic observation and hardness test for the pressure instrumentation pipe and its weldment were carried out in the JMTR Hot Laboratory. This report summarized above data obtained by investigation on the cause of the crack initiation.

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:11.39(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Report on the water leakage from instrumentation pipe in JMTR

JMTR Pressure Measurement Pipe Investigation Committee

JAERI-Review 2003-014, 117 Pages, 2003/03

JAERI-Review-2003-014.pdf:27.62MB

On December 10、2002, the leak was found at the pressure measurement pipe attached to the exit pipe of No.1 filing pump of the refining system of a primary cooling system at JMTR in Oarai Research establishment JAERI. Investigation Committee for Water Leakage from Instrumentation Pipe in JMTR was established and organized by specialists from inside and outside JAERI on December 16 and its meeting was held in public 3 times by 6th January, 2003. They investigated the cause and countermeasures of cracks, and also investigated enhancement of safety management. This is the report on the cause and countermeasures of cracks and enhancement of safety management.

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:4 Percentile:30.46(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Data on loss of off-site electric power simulation tests of the High Temperature Engineering Test Reactor

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

JAERI-Data/Code 2002-015, 39 Pages, 2002/07

JAERI-Data-Code-2002-015.pdf:1.53MB

no abstracts in English

JAEA Reports

Evaluation of thermal displacement behavior of high temperature piping system in power-up test of HTTR, 1; Results up to 20MW operation

Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio

JAERI-Tech 2002-024, 46 Pages, 2002/03

JAERI-Tech-2002-024.pdf:3.29MB

no abstracts in English

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