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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.
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.
Hayashi, Naoki
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The J-PARC is a multi purpose research center using neutron/muon sources for research in material and life science, and using hadron or neutrino beams for nuclear and particle physics. Its proton accelerators have been beam commissioned since 2006. The second accelerator RCS (3GeV Rapid-Cycling Synchrotron) was commissioned in early 2008 and the third accelerator MR (Main Ring; 50GeV proton synchrotron) and experimental facilities will follow. This paper describes the beam commissioning of the accelerators, highlighting the RCS, and discusses the outlook of the RCS power up strategy and future plans for the accelerator complex.
-rays from energy-recovery linacsHajima, 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.
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.
Motooka, Takafumi; Yamamoto, Masahiro
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The corrosion behavior of stainless steel in nitric acid solutions has been usually investigated by short time corrosion tests. The change in corrosion rate with time has been not focused in these tests. In this report, the time-dependency of corrosion rate of stainless steel in nitric acid solution was investigated. Corrosion test of SUS304ULC stainless steel was conducted in boiling nitric acid solution containing Cr(VI) or V(V). The concentration of nitric acid solutions was ranged from 3 to 8 M, and the concentration of Cr(VI) or V(V) was ranged from 0.05 to 1 g/L. The samples of stainless steel were changed every 48 hours without renewal of the nitric acid solution during 240 hours. The obtained results were as follows; (1) The presence of Cr(VI) or V(V) in nitric acid solution accelerated the corrosion rate of stainless steel. The higher corrosion rate was observed in higher nitric acid concentration and greater content of Cr(VI) or V(V). (2) In boiling nitric acid solution with Cr(VI), the corrosion rate rapidly decreased in 8 M nitric acid solution. While in nitric acid solutions with V(V), the corrosion rate was little decreased. (3) The time-dependency of corrosion rates had a good correlation with corrosion potentials of stainless steel. Thus, the corrosion rate of stainless steel in nitric acid solutions with either Cr(VI) or V(V) would be controlled by corrosion potential.
Murakami, Haruko*; Ahn, J.*
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Sakaba, Nariaki; Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Japan Atomic Energy Agency launched the HTTR-IS program which is a nuclear hydrogen production demonstration program using the Japan's first high-temperature gas-cooled reactor HTTR in 2005. It is expected to be the world's first demonstration of nuclear hydrogen. The candidate system of the hydrogen production is a thermochemical water splitting iodine sulphur process (IS process). The thermochemical water splitting process can produce massive quantity of hydrogen without carbon dioxide greenhouse gas emission. This paper focused on the key issues to be developed for the IS process to couple with the HTTR (HTTR-IS system). The key issues to be established are the safety philosophy for non-nuclear grade system as a conventional chemical system and simplification of the plant for an economic competitiveness. The conceptual safety study for non-nuclear system was carried out. The key elements were proposed which can exempt the IS process from "Prevention System 3" and identify abnormal events initiated from the IS process as external events. Also, the conceptual design study for integration of the components such as a Bunsen reactor and a sulphuric acid decomposer was carried out. Reduction of number of components was proposed by coupling with some of equipment. The proposed philosophy and its supporting technologies are expected to contribute economically for the commercialization of nuclear hydrogen.
Seo, Toshihiro; Sasao, Eiji; Notoya, Shin; Shimizu, Kazuhiko
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
JAEA has promoted research and development of HLW disposal technology to contribute to both implementation and safety regulation. This R&D covers the geological environment, repository engineering and performance assessment. At two underground laboratories (Mizunami and Horonobe), surface-based investigations have been completed, and the excavation of shafts and drifts is underway. A complementary study of natural disruptive phenomena, such as volcanism and faulting, has also been conducted. In the ENTRY and QUALITY research facilities, the development of engineering technologies with associated advanced models and databases for quantification of the long-term evolution of the near-field has been carried out. To integrate these R&D activities and manage the huge amount of data that they produce, JAEA has initiated a novel project to develop an advanced knowledge management system which will provide a technical knowledge base for supporting both implementers and regulators.
Shibata, Keiichi; Iwamoto, Osamu; Ichihara, Akira; Iwamoto, Nobuyuki; Kunieda, Satoshi; Otsuka, Naohiko*; Fukahori, Tokio; Nakagawa, Tsuneo; Katakura, Junichi
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Evaluated nuclear data play an important role in nuclear engineering. After we released the first version of Japanese Evaluated Nuclear Data Library JENDL-1 about 30 years ago, JENDL has been revised several times by considering the feedback from users and recent experimental data. JENDL-3.3, which was released in 2002, is being used in various fields, and its reliability has been confirmed. We are developing JENDL-4 for innovative reactors with much emphasis on the improvements of FP and MA data. Nuclear model codes were developed. The actinide data were already released as JENDL Actinoid File 2008. The complete library JENDL-4 will be made available in FY2009.
Suzuki, Mitsutoshi; Namekawa, Takashi; Asano, Takashi; Niita, Koji*
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The advanced MOX fabrication process in FaCT project has been studies to investigate an overall characteristic of nuclear material accounting of Pu in the proposed confinement box. Flow field induced by a forced convection inside the box is numerically simulated to evaluate the MOX particle behavior and a radiation field due to the spontaneous and induced neutrons emitted from Cm and Pu is calculated using PHITS code. The possibility of remote-monitoring techniques using non-destructive assay to apply to a future safegurads measure is invetigated.
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.
Morizono, Koji; Takeuchi, Norihiko; Takayama, Koichi; Deshimaru, Takehide; Mukai, Kazuo
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 5 Pages, 2008/10
Prototype fast breeder reactor Monju is the first power generating FBR in Japan which is a plant for research & development to demonstrate reliability of fast breeder reactor as a power plant and establish sodium handling technology etc. Monju started construction in 1985, achieved initial criticality in 1994 and attained 40% output in 1995, however the sodium leak accident occurred during the test operation at the end of that year. Since then, the plant remained shut down for 12 years. In this subject, our efforts for the restart of this long term shut down plant will be presented.
Rintsu, Yuko*; Serizawa, Shigeru*; Yamazaki, Tetsuo*; Umeyama, Nobuaki*; Moriuchi, Shigeru*; Handa, Hiroyuki*; Onishi, Ryoichi*; Takemura, Morio*; Chino, Masamichi; Nagai, Haruyasu; et al.
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
no abstracts in English
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
Nakagawa, Shigeaki; Saikusa, Akio; Tochio, Daisuke; Takeda, Tetsuaki*
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The intermediate heat exchanger (IHX) is one of key components in the very high temperature reactor (VHTR) system. The IHX is a helium-helium heat exchanger and the secondary hot helium gas heated up to about 900 
C in the IHX is provided to the hydrogen production facility such as IS system which produced the hydrogen by the thermo-chemical water-splitting iodine-sulfur process. The calculation to obtain a precise temperature distribution inside the IHX is required to the reliable design in the VHTR system with the design lifetime of 60 years. The 30 days operation in the HTTR with the reactor outlet coolant temperature of 850 C has been performed and the temperature data for the IHX was obtained. The temperature calculation was performed to simulate the temperature distribution inside the IHX during the rated operation of the HTTR. The calculation result shows a good agreement with the experimental data and this calculation code was validated. It was confirmed that the IHX temperature calculation code was able to simulate precisely the temperature distribution inside the heat exchanger.
Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Senzaki, Masao; Inoue, Naoko; Kuno, Yusuke
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Japan has established the commercial nuclear fuel cycle with LWR as a non-weapon state, and now has been developing FR fuel cycle as the next generation technology. Japanese studies on proliferation resistance have started in 1990's by JNC, predecessor concern of JAEA. The early study was focused on the intrinsic features of Pu and safeguards technologies. Based on TOPS report, JNC has developed a quantitative assessment methodology and modified it by JNC designers. JNC and JAEA intends to play the role as a hub that contributes international collaborative study such as GEN IV PR&PP Experts Group and INPRO based on Japanese experiences of safeguards and nonproliferation efforts, and feedback to domestic experts. This paper will introduce the Japanese Proliferation Resistance studies including, the early studies, international workshops/symposiums, the study in FaCT, international collaboration studies, and the future direction identified for the nuclear fuel cycle systems.
Yamamura, Osamu
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Tokai reprocessing plant (TRP) has been processing over 1,123 tons of spent fuels from the beginning of its active operation in Sept. 1977. For 30 years operation of TRP, many technological problems have been overcome to obtain the stable and reliable operation. The process for establishments of maintenance technology in TRP was evaluated through the analysis of significant plant equipment failures reported to the authorities concerned. Through these troubles and its solution, following knowledge could be obtained.
Nagaoki, Yoshihiro; Kikuchi, Shin; Ichimiya, Masakazu
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
"Fast Reactor Cycle Technology Development (FaCT)" project has been conducted since 2006. In this project, design study and research and development (R&D) on innovative technologies for fast reactor (FR) cycle system are implemented in order to present the conceptual designs of commercial and demonstration facilities by 2015 and start operating demonstration fast reactor in 2025. The R&Ds has been stepped forward into the development stage to establish the realization of innovative technologies which bring excellent performance to fast reactor cycle system. The purpose of R&D by 2010 is to decide weather innovative technologies shall be adopted. So promoting R&D of FR, the project governance was organized. Furthermore, several possible R&D have been effectively carried out within the frameworks of international cooperation, such as GNEP, GIF, and INPRO.