None

Kamiya, Junichiro; Morohashi, Yuko

Kurin Tekunoroji, 35(1), p.39 - 42, 2025/01

no abstracts in English

JAEA's efforts to demonstrate high temperature gas-cooled reactors for carbon-neutral

Inaba, Yoshitomo; Sato, Hiroyuki; Sumita, Junya; Ohashi, Hirofumi; Nishihara, Tetsuo; Sakaba, Nariaki

Nihon Kikai Gakkai-Shi, 127(1267), p.25 - 28, 2024/06

https://doi.org/10.1299/jsmemag.127.1267_25

Aiming to contribute to net-zero emissions through early social implementation of HTGRs, JAEA promote five projects: HTTR-Heat Application Test, HTGR Domestic Demonstration Reactor, UK HTGR Demonstration Program, UK HTGR Fuel Development Program, and Poland HTGR Research Reactor Basic Design. In addition to these five projects, this article provides an overview of the safety demonstration tests using HTTR.

Structure of basaltic glass at pressures up to 18 GPa

Ohashi, Tomonori*; Sakamaki, Tatsuya*; Funakoshi, Kenichi*; Hattori, Takanori; Hisano, Naoki*; Abe, Jun*; Suzuki, Akio*

American Mineralogist, 107(3), p.325 - 335, 2022/03

https://doi.org/10.2138/am-2021-7742

The basaltic glass structure were investigated to 18 GPa using in situ X-ray and neutron diffraction. The O-O coordination number (CN) starts to rise with maintaining the mean O-O distance (r) above 2-4 GPa, and then CN stops increasing and r begins to shrink along with the increase in the Al-O coordination number (CN) above 9 GPa. This is interpreted by the change in the contraction mechanism from tetrahedral network bending to oxygen packing ratio increase via the CN increase. The oxygen packing fraction exceeds the value for dense random packing, suggesting that the oxygen-packing hypothesis cannot account for the pressure-induced structural transformations of silica and silicate glasses. The CN increase at 2-4 GPa reflects the elastic softening of silicate glass, which may causes anomalous elastic moduli of basaltic glass at 2 GPa.

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.

HTTR-GT/H test plant; System performance evaluation for HTTR gas turbine cogeneration plant

Sato, Hiroyuki; Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Yan, X.; Ohashi, Hirofumi

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.759 - 766, 2016/11

This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H plant) so as to confirm that the design meets the requirements with respect to the demonstration test objective. Start-up and shut down operation sequences as well as operability of load following operation were investigated. In addition, system dynamic and control analyses for the test plant in the events of loss of generator load and upset of H plant were performed. The simulation results presented in the paper show that the test plant is suitable for the test bed to validate control schemes against postulated transients in the commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C). The results also lead us to the conclusion that HTTR-GT/H plant can be used to test operational procedure unique to HTGR direct-cycle gas turbine cogeneration.

HTFP for calculation of amount of additionally released fission products from fuel rods of pin-in-block-type high temperature gas-cooled reactors during accident

Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi

JAEA-Data/Code 2015-008, 39 Pages, 2015/06

JAEA-Data-Code-2015-008.pdf:10.32MB

HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.

Study of origin on tritium release into primary coolant for research and testing reactors; Tritium release rate evaluated from JMTR, JRR-3M and JRR-4 operation data

Ishitsuka, Etsuo; Motohashi, Jun; Hanawa, Yoshio; Komeda, Masao; Watahiki, Shunsuke; Mukanova, A.*; Kenzhina, I. E.*; Chikhray, Y.*

JAEA-Technology 2014-025, 77 Pages, 2014/08

JAEA-Technology-2014-025.pdf:43.46MB

It has been shown that tritium concentration in the primary coolant of the JMTR and JRR-3M increases during its operation. In this report, to clarify the tritium sources, the tritium release rate into the primary coolant in each operation cycle for the JMTR, JRR-3M and JRR-4 was evaluated. As a result, the tritium release rate is 8 Bq/Wd in the JRR-4, which has not the beryllium core components installed, and no increase in the tritium concentration during reactor operation is observed. In contrast, the tritium release rate is about 1095 and 60140 Bq/Wd in the JRR-3M and JMTR respectively, which cores contain beryllium components, and where the tritium content increases while reactor operates. It is also observed that the amount of released tritium is lower in the case of new beryllium components installation, and increases with the reactor operating cycle.

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki

Nuclear Engineering and Design, 271, p.309 - 313, 2014/05

https://doi.org/10.1016/j.nucengdes.2013.11.052

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Conceptual design of small-sized HTGR system, 5; Safety design and preliminary safety analysis

Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2013-017, 71 Pages, 2014/02

JAEA-Technology-2013-017.pdf:3.64MB

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.

Evaluation of integrity of coated fuel particles of practical high temperature gas-cooled reactor system

Aihara, Jun; Ueta, Shohei; Ohashi, Hirofumi; Tachibana, Yukio

JAEA-Technology 2012-044, 9 Pages, 2013/02

JAEA-Technology-2012-044.pdf:1.22MB

Evaluation of integrity of coated fuel particles of GTHTR300 has already been carried out. On the other hand, new knowledge on pressure vessel failure, one of the causes of failure of coated fuel particles under irradiation, was obtained by preliminary irradiation experiment of coated fuel particles for HTTR up to official burnup of 7% FIMA. Then in this report, extrapolative evaluation of volume averaged pressure vessel failure probability was carried out based on method which is already published. Considering evaluated pressure vessel failure probability, in addition to corrosion behavior of SiC layer and fuel kernel migration of GTHTR300 coated fuel particles, already analyzed in past paper, volume averaged failure probability of coated fuel particles of GTHTR300 at refueling is small in view of public dose by accident, if initial failure probabilities of coated fuel particles are same as those of HTTR first loading fuel.