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Journal Articles

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 950 $^{circ}$ C, 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.

JAEA Reports

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.

JAEA Reports

Benchmark analysis of KRITZ-2 critical experiments

Okumura, Keisuke; Kawasaki, Kenji*; Mori, Takamasa

JAERI-Research 2005-018, 64 Pages, 2005/08

JAERI-Research-2005-018.pdf:3.26MB

In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degree C) for three different cores loading slightly enriched UO $_{2}$ or MOX fuels. For nuclear data testing, benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO $_{2}$ cores, while the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the libraries are mainly due to the different fission cross section of U-235 in the energy rage below 1.0 eV.

Journal Articles

Irradiation Assisted Stress Corrosion Cracking (IASCC)

Tsukada, Takashi

Zairyo To Kankyo, 52(2), p.66 - 72, 2003/02

http://joi.jlc.jst.go.jp/JST.Journalarchive/jcorr1991/52.66

Irradiation assisted stress corrosion cracking (IASCC) is a potential failure mode suffered by the core-components of austenitic stainless steels in the aged light-water reactor (LWR), which is the intergranular type cracking caused by synergistic effects of neutron/gamma radiation and chemical environment. Effects of radiation on the materials and high-temperature water are discussed in this paper to understand IASCC phenomenon from a mechanistic viewpoint. It is essential to elucidate the radiation-induced microcompositional and microstructural changes in the alloy for mechanistic and predictive investigations of IASCC. Although grain boundary segregations of alloying and impurity elements are significant factors affecting IASCC, it has been considered that the radiation-induced microstructural and mechanical changes of materials play critical roles in IASCC. For mechanistic understanding of IASCC, further fundamental research works with experimental and theoretical approaches are needed. Efforts directed to the researches at the Japan Atomic Energy Research Institute are also described.

JAEA Reports