High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

https://doi.org/10.1016/C2018-0-05383-9

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Study of the applicability of CFD calculation for HTTR reactor

Tsuji, Nobumasa*; Nakano, Masaaki*; Takada, Eiji*; Tokuhara, Kazumi*; Ohashi, Kazutaka*; Okamoto, Futoshi*; Tazawa, Yujiro; Inaba, Yoshitomo; Tachibana, Yukio

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 9 Pages, 2012/10

Passive heat removal performance of the reactor vessel cavity cooling system (RCCS) is of primary concern for enhanced inherent safety of HTGR. In a loss of forced cooling accident, decay heat must be removed by radiation and natural convection of RCCS. Thus thermal hydraulic analysis of reactor internals and RCCS is powerful means for evaluation of the heat removal performance of RCCS. The thermal hydraulic analyses using CFD computation tools are conducted for normal operation of the High Temperature Engineering Test Reactor (HTTR) and are compared to the temperature distribution of measured data. The calculated temperatures on outer faces of the permanent side reflector (PSR) blocks are in fair agreement with measured data. The transient analysis for decay heat removal mode in HTTR is also conducted.

Settlement of materials and life science experimental facility at J-PARC

Harada, Masahide; Meigo, Shinichiro; Ito, Manabu; Dantsuji, Eiji; Takagiwa, Katsunori; Takada, Hiroshi; Maekawa, Fujio; Futakawa, Masatoshi; Nakamura, Mitsutaka; Miyake, Yasuhiro*; et al.

Nuclear Instruments and Methods in Physics Research A, 600(1), p.87 - 90, 2009/02

https://doi.org/10.1016/j.nima.2008.11.058

In MLF of J-PARC, since weights of a building and shields are considerably heavy as 130,000 tons and 80,000 tons, respectively, large settlement of the MLF building is expected. The 3NBT line with 300 m in length is similar. To achieve a precise alignment under the large settlement, we conducted periodical survey measurements at the MLF building and 3NBT. At the completion of construction of the MLF building in December 2005, the settlement was about 40 mm. By extrapolating this result with weights to be added in the future for neutron beam line shields, we predicted that the MLF building settled in about 68mm at the completion of all neutron instrument construction. We decided that the muon target should be installed 5 mm higher than the neutron target with expecting future uneven settlement of the MLF building. The validation in the end of 2007 indicated that the measured level difference came up to the expected value.

Possible unconventional superconductivity and weak magnetism in NaCoOyHO probed by SR

Higemoto, Wataru; Oishi, Kazuki; Koda, Akihiro*; Kadono, Ryosuke*; Sakurai, Hiroya*; Takada, Kazunori*; Muromachi, Eiji*; Sasaki, Takayoshi*

Physica B; Condensed Matter, 374-375, p.274 - 277, 2006/03

https://doi.org/10.1016/j.physb.2005.11.083

no abstracts in English

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR (Contract research)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*

JAERI-Tech 2005-015, 26 Pages, 2005/03

JAERI-Tech-2005-015.pdf:1.77MB

Safety demonstration tests using the HTTR are in progress since 2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.

Core thermal-hydraulic design

Takada, Eiji*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke

Nuclear Engineering and Design, 233(1-3), p.37 - 43, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.07.009

The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88 % of the total flow is achieved at minimum. The maximum fuel temperature appears during the high temperature test operation, and reaches 1492 C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 C. It is confirmed that the core thermal-hydraulic design gives conservative results.

Results of shielding performance test in rise-to-power test of the HTTR

Ueta, Shohei; Takada, Eiji*; Sumita, Junya; Shimizu, Atsushi; Ashikagaya, Yoshinobu; Umeda, Masayuki; Sawa, Kazuhiro

JAERI-Tech 2004-047, 87 Pages, 2004/06

JAERI-Tech-2004-047.pdf:6.24MB

In the radiation shielding design of the High Temperature Engineering Test Reactor (HTTR), strong attention is needed to avoid especially upward neutron streaming. Shielding performance test have been carried out in the Rise-to-power test up to full power operation of 30MW. The measured dose equivalent rates in unrestricted area were lower than the detection limit for neutron-ray, and background level for -ray. The neutron dose equivalent rate measured in the stand pipes room was about 120Sv/h at full power operation, which was much lower than the shielding design (330 mSv/h) and the prediction (10 mSv/h).

Estimation of heat removal characteristics for air-cooler in HTTR

Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Eiji*; Sakaba, Nariaki; Takamatsu, Kuniyoshi

JAERI-Tech 2003-097, 55 Pages, 2004/01

JAERI-Tech-2003-097.pdf:3.34MB

In high temperature engineering test reactor (HTTR) of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler (ACL) by way of the heat exchangers of the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). Therefore, air temperature (secondary-side condition at ACL) is important factor for the heat removal capability of the reactor. Coping with the air temperature, stable reactor inlet temperature control is achieved by adjusting of ACL coolant temperature with coolant (pressurized water and air) flow rate. ACL heat removal characteristic was based on the previous operation data in rise-to-power test and in-service operation at HTTR. And evaluate heat removal capability at summertime air temperature as the most severe condition was estimated. As the result, it was confirmed that the rated power of 30 MW can be removed at the condition of summertime air-temperature.

Safety demonstration test (S1C-2/S2C-1) plan using the HTTR (Contract research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Tochio, Daisuke; Iyoku, Tatsuo

JAERI-Tech 2003-074, 37 Pages, 2003/08

JAERI-Tech-2003-074.pdf:1.83MB

Safety demonstration tests using HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. The first phase of the safety demonstration tests includes reactivity insertion tests by means of control-rod withdrawal and coolant flow reduction tests by tripping the gas circulators. In the second phase, accident simulation tests will be conducted. This paper describes the plan of coolant flow reduction tests by tripping of gas circulators planned in August 2003 with detailed test method, procedure and results of pre-test analysis. The analysis results of the steady state and transient behaviours of the reactor and the plant of the HTTR show that in the case of a rapid decrease of the coolant flow rate, the negative reactivity feedback effect of the core brings the reactor power safely to certain stable level without a reactor scram, and that the temperature transient of the reactor core is slow.

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.