Current status of high temperature gas-cooled reactor development in Japan

Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki

Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08

https://doi.org/10.1016/j.nucengdes.2024.113338

HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.

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.

Conceptual study of an experimental HTGR upgraded from HTTR

Goto, Minoru; Fukaya, Yuji; Mizuta, Naoki; Inaba, Yoshitomo; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

The HTTR (High Temperature engineering Test Reactor) constructed at JAEA-Oarai R&D center is a block-type experimental HTGR (High Temperature Gas-cooled Reactor) with 30 MW thermal power. It attained the first criticality at 1998 and has yielded very useful data for future HTGR design. Although the HTTR was designed very conservatively because the HTTR is the first HTGR for Japan, future HTGRs can be designed with a reasonable conservativeness based on the HTTR data. Additionally, it is possible to enhance the performance of the reactor core by improving the design and introducing new technologies. This paper describes a concept of an experimental HTGR that is upgraded from the HTTR by the reasonable conservativeness, the design improvement and the new technology introduction.

Determination of reactivity and neutron flux using modified neural network for HTGR

Subekti, M.*; Kudo, Kazuhiko*; Nabeshima, Kunihiko; Takamatsu, Kuniyoshi

Atom Indonesia, 43(2), p.93 - 102, 2017/08

https://doi.org/10.17146/aij.2017.683

Reactor kinetics based on point kinetic model have been generally applied as the standard method for neutronics codes. As the central control rod (C-CR) withdrawal test has demonstrated in a prismatic core of HTTR, the transient calculation of kinetic parameter, such as reactivity and neutron fluxes, requires a new method to shorten calculation-process time. Development of neural network method was applied to point kinetic model as the necessity of real-time calculation that could work in parallel with the digital reactivity meter. The combination of TDNN and Jordan RNN, such as TD-Jordan RNN, was the result of the modeling approach. The application of TD-Jordan RNN with adequate learning, tested offline, determined results accurately even when signal inputs were noisy. Furthermore, the preprocessing for neural network input utilized noise reduction as one of the equations to transform two of twelve time-delayed inputs into power corrected inputs.

Visualization of internal structures of reactor core in the HTTR; Proposal of non-destructive inspection by cosmic-ray muon radiography

Takamatsu, Kuniyoshi

Hokeikyo Nyusu, (56), p.2 - 4, 2015/10

http://www.irm.or.jp/news56.pdf

Licensable Patent Information Database

Patent publication (In Japanese)

In our study, we focused on a nondestructive inspection method by cosmic-ray muons which could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.

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 1C. 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.

Cosmic-ray muon radiography for reactor core observation

Takamatsu, Kuniyoshi; Takegami, Hiroaki; Ito, Chikara; Suzuki, Keiichi*; Onuma, Hiroshi*; Hino, Ryutaro; Okumura, Tadahiko*

Annals of Nuclear Energy, 78, p.166 - 175, 2015/04

https://doi.org/10.1016/j.anucene.2014.12.017

In our study, we focused on a nondestructive inspection method by which cosmic-ray muons could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.

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.

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 850C/950C. 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 950C respectively on April 19th. Achievement of the reactor outlet coolant temperature of 950C 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.

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.