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.

Evaluation of tritium release curve in primary coolant of research reactors

Ishitsuka, Etsuo; Kenzhina, I. E.*

Physical Sciences and Technology, 4(1), p.27 - 33, 2018/06

https://doi.org/10.26577/phst-2017-1-120

Increase of tritium concentration in the primary coolant for the research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of the tritium release rate into the primary coolant for the JMTR and the JRR-3M are evaluated. 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. These results show the beryllium components in core strongly affect to the tritium release into the primary coolant. As a result, the tritium release rate is related with produced Li by (n,) reaction from Be, and evaluation results of tritium release curve are shown as the dominant source of tritium release into the primary coolant for the JMTR and the JRR-3M are beryllium components. Scattering of the tritium release rate with irradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 3; Progress of component design

Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi*; Eto, Masao*; Miyagawa, Takayuki*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In the frame work of generation IV international forum, safety design criteria and safety design guideline for the generation IV sodium-cooled fast reactors have been developing. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC. In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX was modified for the primary heat exchanger, which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator, protective wall tube type design is under investigation as an option with less R&D risks.

Applicability study of nuclear graphite material IG-430 to VHTR

Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.

First wall and divertor engineering research for power plant in JAERI

Suzuki, Satoshi; Ezato, Koichiro; Hirose, Takanori; Sato, Kazuyoshi; Yoshida, Hajime; Enoeda, Mikio; Akiba, Masato

Fusion Engineering and Design, 81(1-7), p.93 - 103, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.08.030

This paper presents an R&D activity on the plasma facing components (PFCs), such as first wall and divertor, for the fusion power plant. The PFCs of the power plant will be subjected to heavy neutron irradiation and high heat/particle flux from plasma during the continuous operation. In the present design of the PFCs, the candidate structural material is a reduced activation ferritic-martensitic steel, F82H, from the viewpoints of low activation and high robustness against neutron irradiation, and the candidate armor material is tungsten from the low sputtering yield and low tritium retention points of view. To realize the PFCs using such materials, JAERI has bee extensively conducting R&Ds on; (1) high performance cooling tube, (2) tungsten armor materials, (3) selection of a bonding technique for F82H and tungsten materials and (4) evaluation of structural integrity. Recent achievements on these R&Ds are presented.

Principle design and data of graphite components

Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo*

Nuclear Engineering and Design, 233(1-3), p.251 - 260, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.08.012

The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned.

Overview on materials R&D activities in Japan towards ITER construction and operation

Takatsu, Hideyuki; Sato, Kazuyoshi; Hamada, Kazuya; Nakahira, Masataka; Suzuki, Satoshi; Nakajima, Hideo; Kuroda, Toshimasa*; Nishitani, Takeo; Shikama, Tatsuo*; Shu, Wataru

Journal of Nuclear Materials, 329-333(1), p.178 - 182, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.340

This paper presents an overview on ITER-supporting materials research and development activities and major achievements in Japan during the period from the Co-ordinated Technical Activities to date. In view of the completed engineering design of ITER during the Engineering Design Activities period, research and development efforts since then have been focused: those for reduction of component fabrication cost; those in support of domestic preparations of a structural technical code for construction; those necessary for operation, and been extended to component-level testing rather than pure material testing. They cover materials Research and Development for in-vessel components, vacuum vessel, cryogenic steels of superconducting mgnet and diagnostics components. Major achievements in each research and development area are highlighted and their impact or implication to the design, construction and operation of ITER is presented.

Development of advanced heat removal technology and experimental investigation of the cooling limit; Challenges by porous media, mist cooling, and high-speed swirling flow, JAERI's nuclear research promotion program, H11-037 (Contract research)

Toda, Saburo*; Yuki, Kazuhisa*; Akimoto, Hajime

JAERI-Tech 2004-008, 58 Pages, 2004/03

JAERI-Tech-2004-008.pdf:5.14MB

no abstracts in English

Construction of river transfer models for MOGRA

Hayashi, Hiroko*; Matsuoka, Shungo*; Takahashi, Tomoyuki*; Amano, Hikaru

JAERI-Conf 2003-010, p.122 - 130, 2003/09

https://jopss.jaea.go.jp/pdfdata/JAERI-Conf-2003-010.pdf

Two dynamic compartment models were constructed as basic river models for MOGRA (Migration Of GRound Additions), an environmental-load effect predicting code. One is 1 component river model, in which radionuclides in particulate form and dissolved form are considered to be in equilibrium in the river water. Another one is 2 component river model, in which particulate form and dissolved form are considered to be different component and are separately compartmentalized. In each model the river sediment is set in a compartment, and the sedimentation of particulate form and resuspension of radionuclides in the river sediment are taken into account.To verify the analysis function of the constructed models, calculation conditions were set using data of Cs-137 concentration in the river water derived from Kuji river, Japan, and analysis was carried out. Comparing two models, almost no difference is seen when sedimentation velocity is low, while there is apparent difference when sedimentation velocity is high.

User's manual of SECOM2: A Computer code for seismic system reliability analysis

Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Watanabe, Yuichi*; Tamura, Kazuo*

JAERI-Data/Code 2002-011, 205 Pages, 2002/03

JAERI-Data-Code-2002-011.pdf:8.52MB

This report is a user's manual of seismic system reliability analysis code SECOM2 developed at the JAERI for system reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as calculation of component and system failure probabilities for given seismic motion levels at the site of an NPP based on the response factor method, calculation of accident sequence frequencies and the core damage frequency (CDF), importance analysis using various indicators, uncertainty analysis, and calculation of the CDF taking into account the effect of the correlations of responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about responses and capacities of the components which compose the FT, and seismic hazard curve for the NPP site as input. This report presents calculation method used in the SECOM2 code and how to use those functions in the SECOM2 code.