JOPSS - Search Results

JAEA Reports

Backfill material characteristics using the bentonite/excavated rock mixture in the Horonobe Underground Research Laboratory Project (Contract research)

Sugita, Yutaka; Kikuchi, Hirohito*; Hoshino, Emiko*

JAEA-Data/Code 2020-017, 39 Pages, 2021/01

JAEA-Data-Code-2020-017.pdf:2.96MB

In Japan, high-level radioactive waste (HLW) will be buried in a purpose built repository in deep underground. In the vertical disposal concept of HLW, nuclear waste canisters will be emplaced in excavated vertical disposal holes, surrounded by bentonite/sand mixture. And the galleries will be backfilled with bentonite/excavated rock mixture, which will be isolated with a concrete plug. Japan Atomic Energy Agency has performed swelling test, permeability test, thermal property measurement, uniaxial compression test, water potential measurement and infiltration tests to identify coupled thermal-hydraulic-mechanical-chemical behavior that will operate in the backfill material using excavated rock in the Horonobe Underground Research Laboratory (URL). The obtained data will be used to support an ongoing full scale, in-situ experiment being conducted in the Horonobe URL.

Journal Articles

Analytical results of coolant flow reduction test in the HTTR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10

Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.

JAEA Reports

Validation of the TAC/BLOOST code (Contract research)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

JAERI-Data/Code 2005-003, 31 Pages, 2005/06

JAERI-Data-Code-2005-003.pdf:4.83MB

Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.

JAEA Reports

Proposal for evaluation methods of reactor outlet coolant temperature in HTGRs

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

JAERI-Tech 2005-030, 21 Pages, 2005/05

JAERI-Tech-2005-030.pdf:1.06MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850 $^{circ}$ C/950 $^{circ}$ C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950 $^{circ}$ C respectively on April 19th. Achievement of the reactor outlet coolant temperature of 950 $^{circ}$ C is the first time in Japan as well as the world. This report describes proposal for evaluation methods of reactor outlet coolant temperature in the HTGRs through the HTTR operation experiments. The equation is derived from relationships among PRM reading values, reactor outlet coolant temperature, reactor thermal power and heat removal by VCS. The deliberation processes in this study will be applicable to the research and developments of HTGRs in the future.

Journal Articles

Rise-to-power test result of core outlet coolant tamperature of 950 $^{circ}$ C in HTTR

Iyoku, Tatsuo; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi

UTNL-R-0446, p.14_1 - 14_9, 2005/03

no abstracts in English

JAEA Reports

Rise-to-power test in high temperature engineering test reactor in the high temperature test operation mode; Test progress and summary of test results up to 30MW of reactor thermal power

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tochio, Daisuke; Shimakawa, Satoshi; Nojiri, Naoki; Goto, Minoru; Shibata, Taiju; Ueta, Shohei; et al.

JAERI-Tech 2004-063, 61 Pages, 2004/10

JAERI-Tech-2004-063.pdf:3.14MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850 $^{circ}$ C/950 $^{circ}$ C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950 $^{circ}$ C respectively on April 19th in the single operation mode using only the primary pressurized water cooler. The parallel loaded operation mode using the intermediate heat exchanger and the primary pressurized water cooler was performed from June 2nd and JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test on June 24th from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests were passed successfully in the high temperature test operation mode. Achievement of the reactor-outlet coolant temperature of 950 $^{circ}$ C is the first time in the world. It is possible to extend highly effective power generation with a high-temperature gas turbine and produce hydrogen from water with a high-temperature. This report describes the results of the high-temperature test operation of the HTTR.

JAEA Reports

Test plans of the high temperature test operation at HTTR

Sakaba, Nariaki; Nakagawa, Shigeaki; Takada, Eiji*; Nojiri, Naoki; Shimakawa, Satoshi; Ueta, Shohei; Sawa, Kazuhiro; Fujimoto, Nozomu; Nakazawa, Toshio; Ashikagaya, Yoshinobu; et al.

JAERI-Tech 2003-043, 59 Pages, 2003/03

JAERI-Tech-2003-043.pdf:2.54MB

HTTR plans a high temperature test operation as the fifth step of the rise-to-power tests to achieve a reactor outlet coolant temperature of 950 degrees centigrade in the 2003 fiscal year. Since HTTR is the first HTGR in Japan which uses coated particle fuel as its fuel and helium gas as its coolant, it is necessary that the plan of the high temperature test operation is based on the previous rise-to-power tests with a thermal power of 30 MW and a reactor outlet coolant temperature at 850 degrees centigrade. During the high temperature test operation, reactor characteristics, reactor performances and reactor operations are confirmed for the safety and stability of operations. This report describes the evaluation result of the safety confirmations of the fuel, the control rods and the intermediate heat exchanger for the high temperature test operation. Also, problems which were identified during the previous operations are shown with their solution methods. Additionally, there is a discussion on the contents of the high temperature test operation. As a result of this study, it is shown that the HTTR can safely achieve a thermal power of 30MW with the reactor outlet coolant temperature at 950 degrees centigrade.

JAEA Reports

Rise-to-power test in High Temperature Engineering Test Reactor; Test progress and summary of test results up to 30MW of reactor thermal power

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi; Nojiri, Naoki; Takeda, Takeshi; Saikusa, Akio; Ueta, Shohei; Kojima, Takao; Takada, Eiji*; Saito, Kenji; et al.

JAERI-Tech 2002-069, 87 Pages, 2002/08

JAERI-Tech-2002-069.pdf:10.12MB

Rise-to-power test in the HTTR has been performed from April 23rd to June 6th in 2000 as phase 1 test up to 10MW, from January 29th to March 1st in 2001 as phase 2 test up to 20MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20MW in the high temperature test operation mode. Phase 4 test to achieve the thermal reactor power of 30MW started from October 23rd in 2001. On December 7th it was confirmed that the thermal reactor power reached to 30MW and the reactor outlet coolant temperature reached to 850 $^{circ}$ C. JAERI obtained the certificate of pre-operation test from MEXT because all the pre-operation tests by MEXT were passed successfully. From the test results of rise-up-power test up to 30MW, the performance of reactor and cooling system were confirmed, and it was confirmed that an operation of reactor facility could be performed safely. Some problems to be solved were found through tests. By means of solving them, the reactor operation with the reactor outlet coolant temperature of 950 $^{circ}$ C will be achievable.

JAEA Reports