Refine your search:     
Report No.
 - 
Search Results: Records 1-7 displayed on this page of 7
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Establishment of reasonable 2-D model to investigate heat transfer and flow characteristics by using scale model of vessel cooling system for HTTR

Takada, Shoji; Ngarayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, Kazuhiro*

Mechanical Engineering Journal (Internet), 7(3), p.19-00536_1 - 19-00536_12, 2020/06

In this study reasonable 2D model was established by using FLUENT for start-up of analysis and evaluation of heat transfer flow characteristics in 1/6 scale model of VCS for HTTR. By setting up pressure vessel temperature around 200$$^{circ}$$C about relatively high ratio of heat transfer via natural convection in total heat removal around 20-30%, which is useful for code to experiment benchmark in the aspect to confirm accuracy to predict temperature distribution of components which is heated up by natural convection flow. The numerical results of upper head of pressure vessel by the $$kappa$$-$$omega$$-SST intermittency transition model, which can adequately reproduce the separation, re-adhesion and transition, reproduced the test results including temperature distribution well in contrast to those by the $$kappa$$-$$varepsilon$$ model in both cases that helium gas is evacuated or filled in the pressure vessel. It was emerged that any local hot spot did not appear on the top of upper head of pressure vessel where natural convection flow of air is separated in both cases. In addition, the plume of high temperature helium gas generated by the heating of heater was well mixed in the upper head and uniformly heated the inner surface of upper head without generating hot spots.

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.

Journal Articles

A Rapid evaluation method of the heat removed by a VCS before rise-to-power tests

Takamatsu, Kuniyoshi

Journal of Thermal Science, 24(3), p.295 - 301, 2015/06

 Times Cited Count:1 Percentile:8.21(Thermodynamics)

Before rise-to-power tests, the actual measured value of heat released from the Reactor Pressure Vessel (RPV) or removed by the Vessel Cooling System (VCS) cannot be obtained. It is difficult for operators to evaluate the reactor outlet coolant temperature supplied from the High Temperature Engineering Test Reactor (HTTR) before rise-to-power tests. Therefore, when the actual measured value of heat released from the RPV or removed by the VCS are changed during rise-to-power tests, operators need to evaluate quickly, within a few minutes, the heat removed by the VCS and the reactor outlet coolant temperature of 30 (MW), at the 100% of the reactor power, before the temperature achieves to 967 ($$^{circ}$$C) which is the maximum temperature limit generating the reactor scram. In this paper, a rapid evaluation method for use by operators is presented.

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

Change in heat exchange performance of VCS cooler and its recovery works

Hamamoto, Shimpei; Watanabe, Shuji; Oyama, Sunao*; Ota, Yukimaru; Tochio, Daisuke; Fujimoto, Nozomu

JAERI-Tech 2005-035, 35 Pages, 2005/07

JAERI-Tech-2005-035.pdf:2.54MB

The Vessel Cooling System (VCS) is one of the safety engineered features of the HTTR. The VCS removes the decay heat of the reactor core when the forced circulation can not be maintained due to pipe rupture accidents, etc. VCS cools the core by water cooling panels surrounding the reactor pressure vessel. The VCS also keeps the temperature of the concrete of the primary shielding under the design limit during the operation. The temperature of cooling water of the VCS has recently tended to rise gradually, though the amount of heat removal of VCS has been constant. Accompanying with the increase of the cooling water temperature, the temperature of the shielding concrete is also possible to rise and exceed the limit. The heat exchange performance of the VCS cooler was evaluated and the deterioration of the cooler was verified. Therefore, the cleaning of heat transfer tubes was carried out to recover the heat exchange performance. The cleaning recovered the performance of the VCS cooler drastically and the cooling water temperature of the VCS could be reduced sufficiently.

JAEA Reports

Data on test results of vessel cooling system of High Temperature Engineering Test Reactor

Saikusa, Akio*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

JAERI-Data/Code 2002-027, 34 Pages, 2003/02

JAERI-Data-Code-2002-027.pdf:1.22MB

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28,1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is a first Reactor Cavity Cooling System applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it is confirmed that the VCS heat removal at 30 MW power operation is higher than 0.3MW. This paper shows outline of the VCS and test results on the VCS performance.

Journal Articles

Advanced vessel cooling system concept for high-temperature gas-cooled reactors

Saikusa, Akio; Kunitomi, Kazuhiko; Shiozawa, Shusaku

Nuclear Technology, 118(2), p.89 - 96, 1997/05

 Times Cited Count:7 Percentile:53.76(Nuclear Science & Technology)

no abstracts in English

7 (Records 1-7 displayed on this page)
  • 1