A Simulation study for closed cycle and continuous hydrogen production by a thermo-chemical water-splitting IS process

Kubo, Shinji; Kasahara, Seiji; Sato, Hiroyuki; Imai, Yoshiyuki; Iwatsuki, Jin; Tanaka, Nobuyuki; Miyashita, Reiko*; Tago, Yasuhiro*; Onuki, Kaoru

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/12

A stable hydrogen production via the IS process is relatively difficult because of the unique characteristics of the closed-cycle condition involved. This issue is therefore a high targeted priority when industrializing the process as feasible in a chemical plant. In system of IS process coupled with helium gas heat source, a process control method to maintain mass balance of the process was devised. The method is equipped with measurements of Bunsen reaction composition and allocation of heat for the O and H production sections in strict proportion. Via computer simulation for closed-cycle and fully multi-section driven by high-temperature helium gas, the system worked automatically to maintain stoichiometric production ratio in response to shifts of helium gas conditions.

Nondestructive detection and assay of nuclear material by laser Compton -rays from energy-recovery linacs

Hajima, Ryoichi; Kikuzawa, Nobuhiro; Hayakawa, Takehito; Minehara, Eisuke

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Nondestructive assay of nuclear materials is one of the most urgent research issues for the management of nuclear waste. We apply nuclear resonant fluorescence (NRF), a fingerprint of each isotope, to the nondestructive detection. Detection of NRF signals from objectives irradiated by quasi-monochromatic -rays enables one to make quantitative assay of nuclear materials in a nondestructive manner. In order to obtain high-flux -rays for the above purpose, we propose to utilize an energy-recovery linac (ERL), which produces a high-brightness electron beam with high-average current. A design study of high-flux -ray source based on an ERL shows that a -ray flux of ph/sec/keV is obtained and detection of U-238 of 1 Bq/g in a concrete drum is possible within 1 second. In this paper, we present a design of the assay system and results of Monte Carlo simulations of NRF.

Development of disassembly & pin chopping technology for FBR spent fuels

Kobayashi, Tsuguyuki*; Namba, Takashi*; Kawabe, Yukinari*; Washiya, Tadahiro

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Japan Atomic Power Company (JAPC) and Japan Atomic Energy Agency (JAEA) have been developing fuel disassembly and fuel pin chopping systems for a future Japanese commercial FBR. At first, the wrapper tube is cut by the slit-cut to pull it out, then the fuel pins are cut by the crop-cut at their end-plugs to separate them from the entrance nozzle. The pins are transferred to the magazine of the chopping machine. A series of tests were performed to develop this procedure. As the result of mechanical cutting tests, the CBN wheel was selected. The slit-cut tests were carried out to evaluated the cutting performance of the wheel. Fuel pin handling tests were performed to transfer them from the disassembly machine to the chopping machine. The Saucer tray was selected to receive the disassembled pins. All the pins were transferred and loaded into a magazine of the chopping machine.

Research and development on Water-Cooled Solid Breeder Test Blanket Module in JAEA

Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Hirose, Takanori; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Seki, Yohji; Suzuki, Satoshi; Mori, Kensuke*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Research and development on plasma facing components for fusion reactors in JAEA

Suzuki, Satoshi; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Seki, Yohji; Mori, Kensuke*; Nishi, Hiroshi; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

This paper presents the present status of R&D activities on plasma facing components (PFCs) for fusion reactors, such as ITER and fusion demonstration reactor (DEMO). The PFCs as typified by divertor and first wall components are subjected to high heat flux and particle flux from fusion plasma. It is essential to have sufficient heat removal capability and robust structure against those loadings. JAEA has been carried out to develop the ITER-PFCs which consist of copper alloys and armor materials with high thermal conductivity, such as carbon fiber composites, tungsten and beryllium. The demonstration of the thermomechanical performance has successfully been made under close mutual cooperation between the participant countries of ITER. Currently, the activity on the development of the ITER-PFCs is in a qualification phase prior to the bulk production for construction. In JAEA, the R&Ds on the DEMO-PFCs is being made in parallel with the development activity of the ITER-PFCs.

Radiological assessment of multilateral radioactive effluents in the Pacific Ocean

Nakano, Masanao; Takeishi, Minoru

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

The liquid effluent has been discharged from each nuclear fuel cycle facility in mainly industrially advanced nations with authorization. Since the safety assessment confirms that environmental impact from each facility would be small enough, it was not necessary to assess the effect of the multilateral radioactive effluents. But it will be necessary to quantify the radiological impact between plural countries due to the increase of interest to the global environment and of demand on nuclear energy in the world. In this study, the long-term and multilateral environmental radiological assessment from tritium discharge by effluents from nuclear fuel cycle facilities in the world was carried out. Then, collective and effective doses were evaluated in case of intake of marine products. Additionally, an assessment from the nuclear testing in the atmosphere was carried out. As a result, the concentration and dose level was confirmed not to become an issue for global scale, even if Rokkasho reprocessing plant operates with full scale.

R&D on thermochemical IS process for hydrogen production at JAEA

Onuki, Kaoru; Kubo, Shinji; Terada, Atsuhiko; Iwatsuki, Jin; Takegami, Hiroaki; Okuda, Hiroyuki; Kasahara, Seiji; Tanaka, Nobuyuki; Imai, Yoshiyuki

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production. Closed-cycle operation of the process was demonstrated for one week with hydrogen production rate of about 30NL/h. Components in HSO section were designed featuring a heat exchanger made of silicon carbide. Flowsheet process analysis is now in progress. For higher thermal efficiency, a Bunsen reactor and an EED cell are investigated. Vapor-liquid equilibrium of HI-HO-I for high pressure is measured.

Development of a neutronics design accuracy evaluation solver for next generation reactor physics analysis code system MARBLE

Yokoyama, Kenji; Numata, Kazuyuki*; Hazama, Taira; Ishikawa, Makoto

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

The new solver for cross section adjustment and design accuracy evaluation has been developed for the new reactor physics analysis code system, MARBLE. In this development, object-oriented design was applied for achieving software extendibility. The new solver was successfully designed to easily add a uncertainty prediction method. This extendibility was confirmed by implementing the extended bias method. The new solver reproduces all functions of the conventional code system and can be used as standard solver for cross section adjustment and design accuracy evaluation in MARBLE.

Application of PSA to model facility for MOX fuel fabrication

Tamaki, Hitoshi; Yoshida, Kazuo; Hamaguchi, Yoshikane*

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

A PSA is a comprehensive and structured method for assessing the safety of a nuclear facility. This method also provides risk information that could be applied to effective regulatory activities for nuclear facilities and so on. A PSA procedure for MOX fuel fabrication facility has been developed at JAEA. This procedure consists of two steps. One is called as preliminary PSA using simple methods for likelihood and consequence evaluation through whole processes in the facility. The other step is called as detailed PSA and is carried out to evaluate risk of the significant events using methods corresponding to level 1 PSA and level 2 PSA for nuclear power plants. The procedure was applied to a practical model facility based on process information and handling quantities of materials from the planned MOX fabrication plant to understand risks at whole processes in the model facility. A risk-profile, which consisted of dominant accident sequences, was also obtained through this analysis.

Demonstration of remote fabrication for FBR MOX fuel at the PFPF

Takahashi, Saburo; Kikuno, Hiroshi; Shiromo, Hideo; Kuba, Meiji; Abe, Tomoyuki; Takeda, Seiichiro

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Japan Atomic Energy Agency (JAEA) has been accumulating various experience and knowledge on development of MOX fuel technologies for more than 40 years since 1966. Plutonium Fuel Production Facility (PFPF) has introduced a fully automated and remote operation in 1988 as a pioneer in the world, based on the operational and technical experience obtained in the existing facilities. The PFPF has fabricated MOX fuel assemblies for a fast reactor "JOYO" and a fast breeder reactor "MONJU" so far. Through MOX fuel fabrication for JOYO and MONJU, many operational experiences such as a hold-up material problem have been gained. Based on the experiences, process equipments have been newly developed and a process technology has been improved. As the results, fully automated and remote fabrication technologies including easy contact maintenance of process equipments for FBR MOX fuel have been demonstrated in the PFPF on a large scale.