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Journal Articles

Reduction of heat input to IS (iodine-sulfur) process by removal of HI-I$$_{2}$$-H$$_{2}$$O mixture purification

Kasahara, Seiji; Kubo, Shinji; Tanaka, Nobuyuki; Yan, X.; Onuki, Kaoru

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 10 Pages, 2013/04

In thermochemical hydrogen production IS process, purification of all of the HI-I$$_{2}$$-H$$_{2}$$O (HIx) mixture right after Bunsen reaction has been supposed to prevent negative influences of impurities on operation of following process components. An experimental investigation on electro-electrodialysis (EED) in JAEA suggests possibility of removal of the purification of the HIx fed to anode side of the EED cell. In this study, reduction of heat input to IS process by removal of HIx mixture purification was investigated by process flow calculation. The net heat input to the HI separation subsection was 143.1 kJ/mol-HI with removal of purification of the HIx fed to EED anode side, which was considerably smaller than that of 278.9 kJ/mol-HI in the case of no purification removal. The main factor for the reduction was elimination of the heat input for the anode side feed purification. The removal of the purification lowered EED voltage and reduced electricity input to the EED. The removal also decreased feed rate to the HI distillation column and decreased heat input to the reboiler of the column. These side effects also reduced the net heat input.

Journal Articles

Hydrogen concentration behavior in the IHTS of Monju

Ito, Kazuhiro; Tanabe, Hiromi; Kaneko, Yoshihisa; Kagota, Eiichi; Takahashi, Yasuo

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 10 Pages, 2013/04

The Monju is equipped with two types of hydrogen-meters to detect water leakage in steam generators. Since they are so highly-sensitive as to detect minor water leak from a steam generator tube, they sometimes detect hydrogen concentration increases at plant operational condition changes such as start-up without any water leak. No water leak was experienced during one year operation of the Startup Test up to 40% in 1995, although hydrogen concentration sometimes increased at plant operational condition changes. The H behavior of Monju IHTS during the previous Startup Test was examined and the following knowledge was obtained: The in-sodium H behaves in parallel with the IHTS sodium temperature. In-cover-gas H behavior is more complicated and sensitive to plant operational condition changes such as plant load changes than the in-sodium one. Both types of H-meters underwent a certain degree of zero level drift during one year operation.

Journal Articles

Ongoing validation of sodium fire analysis code system for SFR safety evaluation

Ohno, Shuji; Hamase, Erina; Kamide, Hideki

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 8 Pages, 2013/04

This paper describes outline and current status of ongoing verification and validation activitiess of sodium fire analysis code system for fast reactor plant. The simulation accuracy of sodium droplet burning model is assessed and shown. The effectiveness of three-dimensional gas thermal-hydraulic analysis is also investigated.

Journal Articles

Benchmark calculations on control rod withdrawal tests performed during Phenix End-of-Life experiments

Pascal, V.*; Prulhi$`e$re, G.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Semenov, M.*; Taiwo, T.*; et al.

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 11 Pages, 2013/04

The control rod withdrawal test was one of the various Phenix End-of-Life tests performed in 2009. The main goal was to determine the impact of a rod insertion and/or extraction on the radial power distribution in the fissile core at nominal power. The framework of the Technical Working Group on Fast Reactors (TWG-FR) activities in IAEA, decided to launch a Coordinated Research Project (CRP), devoted to benchmarking analyses on the test. The CRP was performed by experts coming from CEA, ANL, IGCAR, IPPE, IRSN, JAEA, KIT and PSI. After a short description of the test conducted in the Phenix reactor, this paper presents some results obtained in the course of the CRP with special emphasis on control rod efficiencies and power deformation by subassemblies. The paper also discusses the discrepancies found when comparing calculated results with experimental data as well as some preliminary conclusions on the source of these discrepancies.

Journal Articles

Applicability evaluation of tagging-gas failed fuel detection and location system for sodium-cooled large reactor

Aizawa, Kosuke; Chikazawa, Yoshitaka

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 8 Pages, 2013/04

Failed fuel detection and location system for an advanced loop-type sodium-cooled large fast reactor has been studied. In this study, a Tagging-gas (Tag) FFDL system has been investigated. Main concerns of the Tag-FFDL are; gap conductance decrease in the fuel pins and irradiation changing the isotope ratio of tagging gas. The heat conductivity of Kr and Xe used for the tagging gas are lower than that of He used in gas plenum of fuel pin. Thus, maximum fuel temperature with tagging gas FFDL system was analyzed to evaluate tagging gas effect on the fuel temperature. It is expected that the isotope ratio change due to irradiation becomes larger than conventional ones, since JSFR adopts high burnup fuel. Therefore, tag gas isotope change in the JSFR condition has been evaluated regarding transmutation and fission gas release. From the results of the investigations, applicability of Tag-FFDL for JSFR has been evaluated.

Journal Articles

Evaluation of external hazard on JSFR reactor building

Yamamoto, Tomohiko; Kato, Atsushi; Chikazawa, Yoshitaka; Ito, Kei; Iwasaki, Mikinori*; Akiyama, Yo*; Oya, Takeaki*

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 9 Pages, 2013/04

The Japan sodium-cooled fast reactor (JSFR) is planning to adopt a steel-plate reinforced concrete (SC) structure reactor building and an advanced seismic isolation system for reactor building. In the response of Fukushima Dai-ichi Nuclear Power Plant (Fukushima I NPP) accident, the evaluation and countermeasure study of earthquake, and other external hazards on JSFR has been analyzed based on 2010 JSFR design. This paper describes the detail of evaluation and countermeasure of earthquake, tsunami and other external hazards to JSFR reactor building.

Journal Articles

Evaluation of sodium combustion in the JSFR SCCV

Kato, Atsushi; Chikazawa, Yoshitaka; Yamamoto, Tomohiko; Ohno, Shuji; Kubo, Shigenobu; Sakaba, Hiroshi*; Akiyama, Yo*; Iwasaki, Mikinori*

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 9 Pages, 2013/04

After the accident of TEPCO's Fukushima Dai-ichi Nuclear Power Plant, evaluations of severe events beyond the design basis on a NPP are focused. As one of those activities, wide range of sodium combustion and hydrogen generation potentials have been analyzed to investigate potential consequences on SCCV. Structural and boundary integrity of SCCV have been evaluated from sodium combustion analyses for pressure and temperature loads. Hydrogen generation has also been evaluated as potential loads of SCCV.

Journal Articles

Status of experimental and analytical thermal-hydraulic studies on severe accident events at Fukushima Dai-ichi NPP

Takase, Kazuyuki; Yoshida, Hiroyuki; Liu, W.; Misawa, Takeharu; Nagatake, Taku; Yamashita, Susumu

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 6 Pages, 2013/04

Journal Articles

Proposal of assessment of structural integrity on severe accidents for JSFR

Hirose, Yuichi*; Ando, Masanori; Onizawa, Takashi; Wakai, Takashi; Sato, Kenichiro*

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 6 Pages, 2013/04

The purpose of this study is to develop assessment of structural integrity for JSFR's primary system made from 316FR steel and Mod.9Cr-1Mo steel in severe accidents that sodium temperature exceeds the design basis temperature as 650 $$^{circ}$$C. It is important of sodium boundary to prevent damages in high-temperature environment. From this standpoint, the way of stress calculation, evaluation formula including limiting value, safety factor and cumulative damages are considered. This paper provides example to apply these assessment for JSFR under development in Japan.

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