Development of chemical reactors for thermochemical water-splitting IS (Iodine-Sulfer) process

Terada, Atsuhiko; Iwatsuki, Jin; Ota, Hiroyuki; Noguchi, Hiroki; Ishikura, Shuichi*; Hino, Ryutaro; Hirayama, Toshio

Koon Gakkai-Shi, 32(1), p.63 - 68, 2006/01

https://doi.org/10.7791/jhts.32.63

Japan Atomic Energy Research Institute has been conducting study on thermochemical IS process for hydrogen production. A pilot test of IS process is under planning that covers four R&D subjects: (1) construction of a pilot test plant made of industrial materials and completion of hydrogen production test using electrically-heated helium gas as the process heat supplier, (2) development of analytical code system, (3) component tests to assist the hydrogen production test and also to improve the process performance for the commercial plant, (4) design study of HTTR-IS system.

R&D for HTGRs core components

Hanawa, Satoshi; Shibata, Taiju; Sumita, Junya; Kikuchi, Takayuki; Sawa, Kazuhiro; Ishihara, Masahiro; Iyoku, Tatsuo

Koon Gakkai-Shi, 32(1), p.36 - 42, 2006/01

http://ci.nii.ac.jp/naid/10029253598

no abstracts in English

Development of intermediate heat exchanger for high temperature engineering test reactor

Hamamoto, Shimpei; Saikusa, Akio; Shinohara, Masanori; Tachibana, Yukio

Koon Gakkai-Shi, 32(1), p.43 - 49, 2006/01

http://ci.nii.ac.jp/naid/10029253611

The intermediate heat exchanger (IHX) of the High Temperature Engineering Test Reactor (HTTR) is one of the high temperature components of the HTTR and a helium-helium type heat exchanger with the heat capacity of 10 MW. The internal structures such as heat transfer tubes made of Hastelloy XR are used normally at the temperature higher than 900C. They compose the pressure boundary between the primary and secondary helium. Their creep strain and creep damage are evaluated based on the high temperature structural design guideline. The IHX is the first high temperature heat exchanger applied to the reactor. Therefore, Research and development works on the IHX were carried out to confirm the structural integrity of the IHX elements as follows. Experimental and analytical studies were carried out to confirm the structural integrity of the IHX as follows: (1) creep collapse of the tube against external pressure, (2) creep fatigue of the tube against thermal stress, (3) seismic behavior of the tube bundles, (4) thermal hydraulic behavior of the tube bundles and (5) in-service inspection technology of the tube. This report describes the objective and component tests procedure on the IHX and its results.

Research and development of future high temperature gas cooled reactor

Takada, Shoji; Katanishi, Shoji; Yan, X.; Kunitomi, Kazuhiko

Koon Gakkai-Shi, 32(1), p.54 - 62, 2006/01

http://ci.nii.ac.jp/naid/10029253638

Design and development of the high temperature gas cooled reactor with a gas turbine electric generation system, the GTHTR300, was carried out by Japan Atomic Energy Research Institute. The GTHTR300 is expected to be a new electric generation system in the early 2010s. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japanese first high temperature gas cooled reactor HTTR so that many R&Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost. This paper describes the unique design features of the GTHTR300 and its deployment plan.

Development of fabrication technology of ITER shielding blanket

Enoeda, Mikio

Koon Gakkai-Shi, 30(5), p.256 - 262, 2004/09

http://ci.nii.ac.jp/naid/10029252387

Fabrication technologies for ITER in-vessel components, especially the shielding blanket with the separable first wall panel has been developed. Hot Isostatic Pressing (HIP) has been applied to the bonding of Cu-alloy/stainless steel and beryllium/Cu-alloy. First wall mock-ups fabricated by using HIP were tested under high heat fluxes and showed sufficient heat removal and thermal fatigue performance. Water jet and electrical discharge machining have been applied to manufacture slots into the first wall panel and the shield block. With these technologies, a first wall panel prototype and a shielding block 1/2 mock-up were successfully fabricated.

First wall and divertor materials as plasma facing components

Suzuki, Satoshi

Koon Gakkai-Shi, 30(5), p.243 - 247, 2004/09

Selection and the development of plasma facing materials for fusion devices, mainly ITER, are presented. For the divertor, CFC (Carbon fiber reinforced carbon composite) materials are utilized as plasma facing materials in the lower part of vertical targets in ITER. Since the design maximum heat flux to the vertical targets is 20 MW/m, CFC materials, which have higher thermal conductivity than pure copper, are preferable from a heat removal point of view. On the other hand, a plasma facing material of a dome and a liner is tungsten because tungsten has low sputtering yield and has relatively high thermal conductivity among metals. First wall covers 80% of the plasma facing area of ITER. The plasma facing material of the first wall should have good compatibility with plasma. Therefore, beryllium is utilized as a plasma facing material from the low contamination and the minimization of the oxygen impurity to the plasma points of view.

Thermal shock analysis of liquid-mercury spallation target

Ishikura, Shuichi*; Kogawa, Hiroyuki; Futakawa, Masatoshi; Hino, Ryutaro; Date, Hidefumi*

Koon Gakkai-Shi, 28(6), p.329 - 335, 2002/11

http://ci.nii.ac.jp/naid/10010215001

The developments of the neutron scattering facilities are carried out under the high-intensity proton accelerator project promoted by JAERI and KEK. To estimate the structural integrity of the heavy liquid-metal (Hg) target used as a spallation neutron source in a MW-class neutron scattering facility, dynamic stress behavior due to the incident of a 1MW-pulsed proton beam were analyzed by using FEM code. Two-type target containers with semi-cylindrical type and flat-plate type window were used as models for analyses. As a result, it is confirmed that the stress (pressure wave) generated by dynamic thermal shock becomes the largest at the center of window, and the flat-plate type window is more advantageous from the structural viewpoint than the semi-cylindrical type window. It has been understood that the stress generated in the window by the pressure wave can be treated as the secondary stress.

Effects of boron addition on weldability and high temperature strength properties of Hastelloy alloy XR

Watanabe, Katsutoshi; Nakanishi, Tsuneo*; Takatsu, Tamao*; Sahira, Kensho*; Nakajima, Hajime

Koon Gakkai-Shi, 16(6), p.368 - 376, 1990/11

http://ci.nii.ac.jp/naid/40004069342

no abstracts in English

Acceleration test for radiation damage in reactor materials

Shiraishi, K.

Koon Gakkai-Shi, 6(5), p.179 - 186, 1980/00

http://ci.nii.ac.jp/naid/40004068993

no abstracts in English

Pulse method of thermal diffusirity measurements

Koon Gakkai-Shi, 2(2), p.55 - 64, 1976/02

no abstracts in English